WorldWideScience

Sample records for csp reactor assessment

  1. Ethanol steam reforming heated up by molten salt CSP : reactor assessment

    NARCIS (Netherlands)

    Falco, de M.; Gallucci, F.

    2010-01-01

    In this paper hydrogen production via reforming of ethanol has been studied in a novel hybrid plant consisting in a ethanol reformer and a concentrating solar power (CSP) plant using molten salt as heat carrier fluid. The heat needed for the reforming of ethanol has been supplied to the system by

  2. Ethanol steam reforming heated up by molten salt CSP: Reactor assessment

    NARCIS (Netherlands)

    De Falco, Marcello; Gallucci, F.

    2010-01-01

    In this paper hydrogen production via reforming of ethanol has been studied in a novel hybrid plant consisting in a ethanol reformer and a concentrating solar power (CSP) plant using molten salt as heat carrier fluid. The heat needed for the reforming of ethanol has been supplied to the system by

  3. Assessing the future of a CSP industry in Morocco

    International Nuclear Information System (INIS)

    Mahia, Ramon; Arce, Rafael de; Medina, Eva

    2014-01-01

    This article presents the results of a survey on the feasibility of, and difficulties in, establishing a locally CSP manufacturing industry in Morocco. First, the survey explores which specific components of the CSP production chain could be manufactured in Morocco today and which would require moderate or significant changes being made in that country over the next decade. This paper contributes to demonstrating the potential for a CSP manufacturing industry in Morocco at the present time, ideal business models and current restrictions. Second, on the one hand this survey provides insight into the entrepreneurial, policy- and market-related barriers hampering the development of this industry and, on the other, the relative advantages offered by Morocco for the development of a CSP sector. Complementing the empirical findings on foreign direct investment determinants, this exercise stresses the key relevance of the economic context not only in terms of size, stability and predictability of the market, but also in regard to the critical importance of institutional and policy-related issues such as stability and public policy commitment. The results show that prior experience of firms in developing areas is a crucial issue in the accurate assessment of the risks and benefits associated with FDI decisions. - Highlights: • A CSP industry in Morocco is viable under certain adjustments in the next decade. • Policy related barriers are more critical than entrepreneurial or market obstacles. • It urges to provide a legislative and administrative support for CSP initiatives. • The volume of installed CSP capacity in the region doesn't reach a critical level. • Some foreign investors might have a negative miss perception of Moroccan reality

  4. PyCSP - controlled concurrency

    DEFF Research Database (Denmark)

    Vinter, Brian; Friborg, Rune Møllegaard; Bjørndalen, John Markus

    2010-01-01

    Producing readable and correct programs while at the same time taking advantage of multi-core architectures is a challenge. PyCSP is an implementation of Communicating Sequential Processes algebra (CSP) for the Python programming language, that take advantage of CSP's formal and verifiable approach...... to controlling concurrency and the readability of Python source code. We describe PyCSP, demonstrate it through examples and demonstrate how PyCSP compares to Pthreads in a master-worker benchmark....

  5. Value generation of future CSP projects in North Africa

    International Nuclear Information System (INIS)

    Kost, Christoph; Engelken, Maximilian; Schlegl, Thomas

    2012-01-01

    This paper discusses the value generation potential for local and international industry in different development scenarios of the concentrating solar power (CSP) market in North Africa until 2030. It analyzes the economic impact resulting from the participation of North African and European companies during construction and operation of CSP plants. The assessment is based on a self-developed solar technologies market development model (STMD) that includes economic and technical requirements and constraints for the creation of a local CSP market. In-depth interviews with industry stakeholders provide specific input, validate the calculations and complement the quantitative model results and conclusions. Long-term potential for locally generated revenues from CSP plant construction are modeled and lead to a share of local revenues of up to 60%. Potential market size of solar power plants in North Africa could reach total revenues of 120 Billion euros and thus demand for components and services contribute to national gross domestic products significantly. Recommendations are given for regional industry cooperation and policy actions for the support of local and international CSP industry in North Africa in order to improve the investment environment and growth of renewable energies in the region. - Highlights: ►New economic model to evaluate value generation of CSP take-off in North Africa. ►CSP components are assessed regarding their potentials to be produced locally. ►Potential for locally generated revenues of CSP plants: 60% of total value. ►Socio-economic impacts of RE projects become more relevant to investment decisions.

  6. CSP for Executable Scientific Workflows

    DEFF Research Database (Denmark)

    Friborg, Rune Møllegaard

    and can usually benefit performance-wise from both multiprocessing, cluster and grid environments. PyCSP is an implementation of Communicating Sequential Processes (CSP) for the Python programming language and takes advantage of CSP's formal and verifiable approach to controlling concurrency...... on multi-processing and cluster computing using PyCSP. Additionally, McStas is demonstrated to utilise grid computing resources using PyCSP. Finally, this thesis presents a new dynamic channel model, which has not yet been implemented for PyCSP. The dynamic channel is able to change the internal...... synchronisation mechanisms on-the-fly, depending on the location and number of channel-ends connected. Thus it may start out as a simple local pipe and evolve into a distributed channel spanning multiple nodes. This channel is a necessary next step for PyCSP to allow for complete freedom in executing CSP...

  7. PyCSP - controlled concurrency

    DEFF Research Database (Denmark)

    Friborg, Rune Møllegaard; Vinter, Brian; Bjørndalen, John Markus

    Producing readable and correct programs while at the same time taking advantage of multi-core architectures is a challenge. PyCSP is an implementation of Communicating Sequential Processes algebra (CSP) for the Python programming language, taking advantage of CSP’s formal and verifiable approach...... to controlling concurrency and the readability of Python source code. We describe PyCSP, demonstrate it through examples and demonstrate how PyCSP compares to Pthreads using a benchmark....

  8. PV integration into a CSP plant

    Science.gov (United States)

    Carvajal, Javier López; Barea, Jose M.; Barragan, Jose; Ortega, Carlos

    2017-06-01

    This paper describes a preliminary techno-economic analysis of the integration of a PV plant into an optimized Parabolic Trough Plant in order to reduce the online consumptions and thus, increase the net electricity injected into the grid. The idea is to assess the feasibility of such project and see what configuration would be the optimal. An extra effort has been made in terms of modelling as the analysis has to be done to the integrated CSP + PV plant instead of analyzing them independently. Two different technologies have been considered for the PV plant, fix and one-axis tracking. Additionally three different scenarios have been considered for the CSP plant auxiliary consumptions as they are essential for determining the optimal PV plant (the higher the auxiliary consumption the higher the optimal PV plant). As could be expected, the results for all cases with PV show an improvement in terms of electricity generation and also in terms of LCOE with respect to the CSP plant. Such improvement is slightly higher with tracking technology for this specific study. Although this exercise has been done to an already designed CSP plant (so only the PV plant had to be optimized), the methodology could be applied for the optimization of an integrated CSP + PV plant during the design phase.

  9. Thermochemical storage for CSP via redox structured reactors/heat exchangers: The RESTRUCTURE project

    Science.gov (United States)

    Karagiannakis, George; Pagkoura, Chrysoula; Konstandopoulos, Athanasios G.; Tescari, Stefania; Singh, Abhishek; Roeb, Martin; Lange, Matthias; Marcher, Johnny; Jové, Aleix; Prieto, Cristina; Rattenbury, Michael; Chasiotis, Andreas

    2017-06-01

    The present work provides an overview of activities performed in the framework of the EU-funded collaborative project RESTRUCTURE, the main goal of which was to develop and validate a compact structured reactor/heat exchanger for thermochemical storage driven by 2-step high temperature redox metal oxide cycles. The starting point of development path included redox materials qualification via both theoretical and lab-scale experimental studies. Most favorable compositions were cobalt oxide/alumina composites. Preparation of small-scale structured bodies included various approaches, ranging from perforated pellets to more sophisticated honeycomb geometries, fabricated by extrusion and coating. Proof-of-concept of the proposed novel reactor/heat exchanger was successfully validated in small-scale structures and the next step included scaling up of redox honeycombs production. Significant challenges were identified for the case of extruded full-size bodies and the final qualified approach related to preparation of cordierite substrates coated with cobalt oxide. The successful experimental evaluation of the pilot reactor/heat exchanger system constructed motivated the preliminary techno-economic evaluation of the proposed novel thermochemical energy storage concept. Taking into account experimental results, available technologies and standard design aspects a model for a 70.5 MWe CSP plant was defined. Estimated LCOE costs were calculated to be in the range of reference values for Combined Cycle Power Plants operated by natural gas. One of main cost contributors was the storage system itself, partially due to relatively high cost of cobalt oxide. This highlighted the need to identify less costly and equally efficient to cobalt oxide redox materials.

  10. Selective C(sp2)-C(sp) bond cleavage: the nitrogenation of alkynes to amides.

    Science.gov (United States)

    Qin, Chong; Feng, Peng; Ou, Yang; Shen, Tao; Wang, Teng; Jiao, Ning

    2013-07-22

    Breakthrough: A novel catalyzed direct highly selective C(sp2)-C(sp) bond functionalization of alkynes to amides has been developed. Nitrogenation is achieved by the highly selective C(sp2)-C(sp) bond cleavage of aryl-substituted alkynes. The oxidant-free and mild conditions and wide substrate scope make this method very practical. Copyright © 2013 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  11. PyCSP - Communicating Sequential Processes for Python

    DEFF Research Database (Denmark)

    Vinter, Brian; Bjørndalen, John Markus; Anshus, Otto Johan

    CSP presently supports the core CSP abstractions. We introduce the PyCSP library, its implementation, a few performance benchmarks, and show example code using PyCSP. An early prototype of PyCSP has been used in this year's Extreme Multiprogramming Class at the CS department, university of Copenhagen......The Python programming language is effective for rapidly specifying programs and experimenting with them. It is increasingly being used in computational sciences, and in teaching computer science. CSP is effective for describing concurrency. It has become especially relevant with the emergence...... of commodity multi-core architectures. We are interested in exploring how a combination of Python and CSP can benefit both the computational sciences and the hands-on teaching of distributed and parallel computing in computer science. To make this possible, we have developed PyCSP, a CSP library for Python. Py...

  12. CSP Design Model and Tool Support

    NARCIS (Netherlands)

    Volkerink, H.J.; Volkerink, H.J.; Hilderink, G.H.; Broenink, Johannes F.; Vervoort, Wiek; Welch, P.H.; Bakkers, André

    The CSP paradigm is known as a powerful concept for designing and analysing the architectural and behavioural parts of concurrent software. Although the theory of CSP is useful for mathematicians, the programming language occam has been derived from CSP that is useful for any engineering practice.

  13. Designing Animation Facilities for gCSP

    NARCIS (Netherlands)

    van der Steen, T.T.J.; Groothuis, M.A.; Broenink, Johannes F.

    To improve feedback on how concurrent CSP-based programs run, the graphical CSP design tool has been extended with animation facilities. The state of processes, constructs, and channel ends are indicated with colours both in the gCSP diagrams and in the composition tree (hierarchical tree showing

  14. gCSP occam Code Generation for RMoX

    NARCIS (Netherlands)

    Groothuis, M.A.; Liet, Geert K.; Broenink, Johannes F.; Roebbers, H.W.; Sunter, J.P.E.; Welch, P.H.; Wood, D.C.

    2005-01-01

    gCSP is a graphical tool for creating and editing CSP diagrams. gCSP is used in our labs to generate the embedded software framework for our control systems. As a further extension to our gCSP tool, an occam code generator has been constructed. Generating occam from CSP diagrams gives opportunities

  15. Hybrid concentrated solar power (CSP)–biomass plants in a semiarid region: A strategy for CSP deployment in Brazil

    International Nuclear Information System (INIS)

    Soria, Rafael; Portugal-Pereira, Joana; Szklo, Alexandre; Milani, Rodrigo; Schaeffer, Roberto

    2015-01-01

    The production of electricity using concentrated solar power (CSP) technology is not yet possible in Brazil due to the technology’s high capital costs and the lack of a local industry. However, this study introduces a low-cost approach to CSP in Brazil by describing and simulating the operation of hybrid CSP plants that use sustainably managed biomass in Brazil’s semiarid northeast. Biomass hybridisation of a CSP plant with a solar multiple (SM) of 1.2 and a biomass fill fraction (BFF) of 30% can generate electricity at 110 USD/MWh. The high direct normal irradiation (DNI) and the availability of local low-cost biomass in Brazil’s semiarid northeast suggest the possibility of developing a CSP industry capable of supplying low-cost components under a national program framework, with the co-benefits of local job and income generation. For example, the deployment of 10 CSP plants of 30 MWe each would generate 760 direct and indirect jobs during the 24 months of plant construction and approximately 2100 annual jobs associated with the operation and maintenance (O&M) of the generating units. These 10 new units would generate additional local income on the order of USD 57 million. - Highlights: • CSP plant with supplementary biomass hybridisation is a strategic option for Brazil. • DNI and biomass availability in Brazil's semiarid can foster local CSP industry. • LCOE of CSP would cost 11 cent USD/kWh becoming competitive at solar auctions. • Co-benefits of local job and income generation due to CSP in Brazil are high.

  16. Selective C(sp3)−H aerobic oxidation enabled by decatungstate photocatalysis in flow

    NARCIS (Netherlands)

    Laudadio, G.; Govaerts, S.; Wang, Y.; Ravelli, D.; Koolman, H.; Fagnoni, M.; Djuric, S.; Noel, T.

    2018-01-01

    A mild and selective C(sp3)−H aerobic oxidation enabled by decatungstate photocatalysis has been developed. The reaction can be significantly improved in a microflow reactor enabling the safe use of oxygen and enhanced irradiation of the reaction mixture. Our method allows for the oxidation of both

  17. Durability of coconut shell powder (CSP) concrete

    Science.gov (United States)

    Leman, A. S.; Shahidan, S.; Senin, M. S.; Shamsuddin, S. M.; Anak Guntor, N. A.; Zuki, S. S. Mohd; Khalid, F. S.; Azhar, A. T. S.; Razak, N. H. S.

    2017-11-01

    The rising cost of construction in developing countries like Malaysia has led concrete experts to explore alternative materials such as coconut shells which are renewable and possess high potential to be used as construction material. Coconut shell powder in varying percentages of1%, 3% and 5% was used as filler material in concrete grade 30 and evaluated after a curing period of 7 days and 28days respectively. Compressive strength, water absorption and carbonation tests were conducted to evaluate the strength and durability of CSP concrete in comparison with normal concrete. The test results revealed that 1%, 3% and 5% of CSP concrete achieved a compressive strength of 47.65 MPa, 45.6 MPa and 40.55% respectively. The rate of water absorption of CSP concrete was recorded as 3.21%, 2.47%, and 2.73% for 1%, 3% and 5% of CSP concrete respectively. Although CSP contained a carbon composition of 47%, the carbonation test showed that CSP no signs of carbon were detected inside the concrete. To conclude, CSP offers great prospects as it demonstrated relatively high durability as a construction material.

  18. The value of dispatchability of CSP plants in the electricity systems of Morocco and Algeria

    International Nuclear Information System (INIS)

    Brand, Bernhard; Boudghene Stambouli, Amine; Zejli, Driss

    2012-01-01

    This paper examines the effects of an increased integration of concentrated solar power (CSP) into the conventional electricity systems of Morocco and Algeria. A cost-minimizing linear optimization tool was used to calculate the best CSP plant configuration for Morocco's coal-dominated power system as well as for Algeria, where flexible gas-fired power plants prevail. The results demonstrate that in both North African countries, storage-based CSP plants offer significant economic advantages over non-storage, low-dispatchable CSP configurations. However, in a generalized renewable integration scenario, where CSP has to compete with other renewable generation technologies, like wind or photovoltaic (PV) power, it was found that the cost advantages of dispatchability only justify CSP investments when a relatively high renewable penetration is targeted in the electricity mix. - Highlights: ► Market model to optimize CSP plant configuration in North African power systems. ► Value of storage-based CSP plants compared to non-dispatchable configurations: 28–55 €/MWh. ► Assessment of Morocco's and Algeria's renewable electricity targets until 2030. ► CSP becomes more competitive with intermittent technologies when high RES-E quota are targeted.

  19. Economic assessment and optimal operation of CSP systems with TES in California electricity markets

    Science.gov (United States)

    Dowling, Alexander W.; Dyreson, Ana; Miller, Franklin; Zavala, Victor M.

    2017-06-01

    The economics and performance of concentrated power (CSP) systems with thermal energy storage (TES) inherently depend on operating policies and the surrounding weather conditions and electricity markets. We present an integrated economic assessment framework to quantify the maximum possible revenues from simultaneous energy and ancillary services sales by CSP systems. The framework includes both discrete start-up/shutdown restrictions and detailed physical models. Analysis of coinci-dental historical market and meteorological data reveals provision of ancillary services increases market revenue 18% to 37% relative to energy-only participation. Surprisingly, only 53% to 62% of these revenues are available through sole participation in the day-ahead market, indicating significant opportunities at faster timescales. Motivated by water-usage concerns and permitting requirements, we also describe a new nighttime radiative-enhanced dry-cooling system with cold-side storage that consumes no water and offers higher effciencies than traditional air-cooled designs. Operation of this new system is complicated by the cold-side storage and inherent coupling between the cooling system and power plant, further motivating integrated economic analysis.

  20. Takaful Operators’ Corporate Social Performance (CSP: An Industry Perspective

    Directory of Open Access Journals (Sweden)

    Muhamat Amirul Afif

    2017-01-01

    Full Text Available Takaful operators which are part of Islamic financial institutions (IFIs derive their fundamental principles from shariah. These religious based institutions are expected to fulfill the two important roles in their business operations: commercially profitable and socially responsible. Nevertheless, their societal role is rarely measured and discussed. Therefore, this study appraised the societal role of takaful operators by assessing the components which have been proposed under the corporate social performance (CSP theme for IFIs. This study has arranged structured interview sessions with the Chief Investment Officers and Heads of Investment of each of the eleven takaful operators in Malaysia. The Delphi-style technique was adopted when developing the interview questions. The questions were developed in the form of a five-point Likert scale, addressing specific issues on CSP of takaful operators. In addition, information on takaful operators’ CSR activities, zakat and tax payment were gathered from the companies’ websites and annual report of takaful operators. The study concludes that takaful operators in Malaysia have achieved their societal role through two channels: CSP programmes financed from companies’ profits and fulfillment of CSP as a result of business-community agenda. This study covers every takaful operator in Malaysia and the results reflect industry opinion.

  1. Assessing the potential role of concentrated solar power (CSP) for the northeast power system of Brazil using a detailed power system model

    International Nuclear Information System (INIS)

    Fichter, Tobias; Soria, Rafael; Szklo, Alexandre; Schaeffer, Roberto; Lucena, Andre F.P.

    2017-01-01

    One of the technologies that stand out as an alternative to provide additional flexibility to power systems with large penetration of variable renewable energy (VRE), especially for regions with high direct normal irradiation (DNI), is concentrated solar power (CSP) plants coupled to thermal energy storage (TES) and back-up (BUS) systems. Brazil can develop this technology domestically, especially in its Northeast region, where most of VRE capacity is being deployed and where lies most of the CSP potential of the country. This work applies the Capacity Expansion Model REMix-CEM, which allows considering dispatch constraints of thermal power plants in long-term capacity expansion optimization. REMix-CEM calculates the optimal CSP plant configuration and its dispatch strategy from a central planning perspective. Results showed that the hybridization of CSP plants with jurema-preta biomass (CSP-BIO) becomes a least-cost option for Brazil by 2040. CSP-BIO contributes to the Northeast power system by regularizing the energy imbalance that results from the large-scale VRE expansion along with conventional inflexible power plants. CSP-BIO plants are able to increase frequency response and operational reserve services and can provide the required additional flexibility that the Northeast power system of Brazil will require into the future. - Highlights: • Concentrating solar power (CSP) plants provide flexibility to power systems. • CSP configuration is optimized endogenously during capacity expansion optimization. • CSP hybridized with biomass supports grid-integration of variable renewable energy. • CSP become the least-cost option for the Northeast power system of Brazil by 2040.

  2. A Method to Assess Flux Hazards at CSP Plants to Reduce Avian Mortality

    Energy Technology Data Exchange (ETDEWEB)

    Ho, Clifford K.; Wendelin, Timothy; Horstman, Luke; Yellowhair, Julius

    2017-06-27

    A method to evaluate avian flux hazards at concentrating solar power plants (CSP) has been developed. A heat-transfer model has been coupled to simulations of the irradiance in the airspace above a CSP plant to determine the feather temperature along prescribed bird flight paths. Probabilistic modeling results show that the irradiance and assumed feather properties (thickness, absorptance, heat capacity) have the most significant impact on the simulated feather temperature, which can increase rapidly (hundreds of degrees Celsius in seconds) depending on the parameter values. The avian flux hazard model is being combined with a plant performance model to identify alternative heliostat standby aiming strategies that minimize both avian flux hazards and negative impacts on plant performance.

  3. General introduction CSP Technologies and grid management

    OpenAIRE

    Hoffschmidt, Bernhard

    2017-01-01

    Der Vortrag gibt einen Überblick über alle relevanten CSP Technologien und beschreibt die besondere Charakteristik der Stromproduktion sowie die aktuelle und mittelfristige Markt- und Kostensituation. Für eine weitere Kostenreduktion wird der Vorteil eines PV-CSP Hybrid Kraftwerks beschrieben.

  4. Concentrated solar power (CSP innovation analysis in South Africa

    Directory of Open Access Journals (Sweden)

    Craig, Toyosi Onalapo

    2017-08-01

    Full Text Available South Africa aims to generate 42 per cent of its electricity from renewable energy technology sources by 2030. Concentrating solar power (CSP is one of the major renewable energy technologies that have been prioritised by South Africa, given the abundant solar resources available in the region. Seven CSP plants have been, or are being, built; three of them are already connected to the national grid. However, the impacts of this technology on South African research, development, and innovation have not been investigated to date. This paper thus analyses the CSP technologies in South Africa in terms of the existing technology adoption models and diffusion strategies, used by government and its agencies, to improve the development and deployment of these technologies. It is found that CSP has been treated generally like other renewable energy technologies through the Renewable Energy Independent Power Producer Procurement Programme (REIPPPP, although a tariff plan for CSP plants of the future has been made. No specific technology diffusion or adoption model for CSP was found; so this paper explores how it can be developed.

  5. Accelerated thermal and mechanical testing of CSP assemblies

    Science.gov (United States)

    Ghaffarian, R.

    2000-01-01

    Chip Scale Packages (CSP) are now widely used for many electronic applications including portable and telecommunication products. A test vehicle (TV-1) with eleven package types and pitches was built and tested by the JPL MicrotypeBGA Consortium during 1997 to 1999. Lessons learned by the team were published as a guidelines document for industry use. The finer pitch CSP packages which recently became available were indluded in the next test vehicle of the JPL CSP Consortium.

  6. Machine-Checkable Timed CSP

    Science.gov (United States)

    Goethel, Thomas; Glesner, Sabine

    2009-01-01

    The correctness of safety-critical embedded software is crucial, whereas non-functional properties like deadlock-freedom and real-time constraints are particularly important. The real-time calculus Timed Communicating Sequential Processes (CSP) is capable of expressing such properties and can therefore be used to verify embedded software. In this paper, we present our formalization of Timed CSP in the Isabelle/HOL theorem prover, which we have formulated as an operational coalgebraic semantics together with bisimulation equivalences and coalgebraic invariants. Furthermore, we apply these techniques in an abstract specification with real-time constraints, which is the basis for current work in which we verify the components of a simple real-time operating system deployed on a satellite.

  7. Middle East and North Africa Region Assessment of the Local Manufacturing Potential for Concentrated Solar Power (CSP) Projects

    Energy Technology Data Exchange (ETDEWEB)

    Gazzo, A.; Gousseland, P.; Verdier, J. [Ernst and Young et Associes, Neuilly-Sur-Seine (France); Kost, C.; Morin, G.; Engelken, M.; Schrof, J.; Nitz, P.; Selt, J.; Platzer, W. [Fraunhofer Institute for Solar Energy Systems ISE, Freiburg (Germany); Ragwitz, M.; Boie, I.; Hauptstock, D.; Eichhammer, W. [Fraunhofer Institute for Systems and Innovation Research ISI, Karlsruhe (Germany)

    2011-01-15

    The MENA CSP (Middle East and North Africa - Concentrated Solar Power) plan is an ambitious scheme with an appeal to anyone concerned about climate change and convinced by the need for clean, renewable power. But what does it really mean for the average citizen of say Morocco or Tunisia? The World Bank sees potential for significant job and wealth creation in solar energy producing countries. If the CSP market grows rapidly over the next few years, equipment manufacturing will be essential to supply this new sector. This study proposes roadmaps and an action plan to help develop the potential of locally manufactured CSP components in the existing industry and for new market entrants.

  8. Subject-based feature extraction by using fisher WPD-CSP in brain-computer interfaces.

    Science.gov (United States)

    Yang, Banghua; Li, Huarong; Wang, Qian; Zhang, Yunyuan

    2016-06-01

    Feature extraction of electroencephalogram (EEG) plays a vital role in brain-computer interfaces (BCIs). In recent years, common spatial pattern (CSP) has been proven to be an effective feature extraction method. However, the traditional CSP has disadvantages of requiring a lot of input channels and the lack of frequency information. In order to remedy the defects of CSP, wavelet packet decomposition (WPD) and CSP are combined to extract effective features. But WPD-CSP method considers less about extracting specific features that are fitted for the specific subject. So a subject-based feature extraction method using fisher WPD-CSP is proposed in this paper. The idea of proposed method is to adapt fisher WPD-CSP to each subject separately. It mainly includes the following six steps: (1) original EEG signals from all channels are decomposed into a series of sub-bands using WPD; (2) average power values of obtained sub-bands are computed; (3) the specified sub-bands with larger values of fisher distance according to average power are selected for that particular subject; (4) each selected sub-band is reconstructed to be regarded as a new EEG channel; (5) all new EEG channels are used as input of the CSP and a six-dimensional feature vector is obtained by the CSP. The subject-based feature extraction model is so formed; (6) the probabilistic neural network (PNN) is used as the classifier and the classification accuracy is obtained. Data from six subjects are processed by the subject-based fisher WPD-CSP, the non-subject-based fisher WPD-CSP and WPD-CSP, respectively. Compared with non-subject-based fisher WPD-CSP and WPD-CSP, the results show that the proposed method yields better performance (sensitivity: 88.7±0.9%, and specificity: 91±1%) and the classification accuracy from subject-based fisher WPD-CSP is increased by 6-12% and 14%, respectively. The proposed subject-based fisher WPD-CSP method can not only remedy disadvantages of CSP by WPD but also discriminate

  9. Properties of concrete containing coconut shell powder (CSP) as a filler

    Science.gov (United States)

    Leman, A. S.; Shahidan, S.; Nasir, A. J.; Senin, M. S.; Zuki, S. S. Mohd; Ibrahim, M. H. Wan; Deraman, R.; Khalid, F. S.; Azhar, A. T. S.

    2017-11-01

    Coconut shellsare a type of agricultural waste which can be converted into useful material. Therefore,this study was conducted to investigate the properties of concrete which uses coconut shell powder (CSP) filler material and to define the optimum percentage of CSP which can be used asfiller material in concrete. Comparisons have been made between normal concrete mixes andconcrete containing CSP. In this study, CSP was added into concrete mixes invaryingpercentages (0%, 2%, 4%, 6%, 8% and 10%). The coconut shell was grounded into afine powder before use. Experimental tests which have been conducted in this study include theslump test, compressive test and splitting tensile strength test. CSP have the potential to be used as a concrete filler and thus the findings of this study may be applied to the construction industry. The use of CSP as a filler in concrete can help make the earth a more sustainable and greener place to live in.

  10. Selective C(sp3 )-H Aerobic Oxidation Enabled by Decatungstate Photocatalysis in Flow.

    Science.gov (United States)

    Laudadio, Gabriele; Govaerts, Sebastian; Wang, Ying; Ravelli, Davide; Koolman, Hannes F; Fagnoni, Maurizio; Djuric, Stevan W; Noël, Timothy

    2018-04-03

    A mild and selective C(sp 3 )-H aerobic oxidation enabled by decatungstate photocatalysis has been developed. The reaction can be significantly improved in a microflow reactor enabling the safe use of oxygen and enhanced irradiation of the reaction mixture. Our method allows for the oxidation of both activated and unactivated C-H bonds (30 examples). The ability to selectively oxidize natural scaffolds, such as (-)-ambroxide, pregnenolone acetate, (+)-sclareolide, and artemisinin, exemplifies the utility of this new method. © 2018 The Authors. Published by Wiley-VCH Verlag GmbH & Co. KGaA.

  11. Embodied energy and emergy analyses of a concentrating solar power (CSP) system

    International Nuclear Information System (INIS)

    Zhang Meimei; Wang Zhifeng; Xu Chao; Jiang Hui

    2012-01-01

    Although concentrating solar power (CSP) technology has been projected as one of the most promising candidates to replace conventional power plants burning fossil fuels, the potential advantages and disadvantages of the CSP technology have not been thoroughly evaluated. To better understand the performance of the CSP technology, this paper presents an ecological accounting framework based on embodied energy and emergy analyses methods. The analyses are performed for the 1.5 MW Dahan solar tower power plant in Beijing, China and different evaluation indices used in the embodied energy and emergy analyses are employed to evaluate the plant performance. Our analysis of the CSP plant are compared with six Italian power plants with different energy sources and an American PV plant, which demonstrates the CSP is the superior technology. - Highlights: ► Embodied energy and emergy analyses are employed to evaluate the first solar tower power plant in China. ► Different evaluation indices are quantitatively analyzed to show the advantages of CSP technology. ► This analysis provides insights for making energy policy and investment decisions about CSP technology.

  12. The cost of integration of parabolic trough CSP plants in isolated Mediterranean power systems

    International Nuclear Information System (INIS)

    Poullikkas, Andreas; Hadjipaschalis, Ioannis; Kourtis, George

    2010-01-01

    In this work, a technical and economic analysis concerning the integration of parabolic trough concentrated solar power (CSP) technologies, with or without thermal storage capability, in an existing typical small isolated Mediterranean power generation system, in the absence of a feed-in tariff scheme, is carried out. In addition to the business as usual (BAU) scenario, five more scenarios are examined in the analysis in order to assess the electricity unit cost with the penetration of parabolic trough CSP plants of 50 MWe or 100 MWe, with or without thermal storage capability. Based on the input data and assumptions made, the simulations indicated that the scenario with the utilization of a single parabolic trough CSP plant (either 50 MWe or 100 MWe and with or without thermal storage capability) in combination with BAU will effect an insignificant change in the electricity unit cost of the generation system compared to the BAU scenario. In addition, a sensitivity analysis on natural gas price, showed that increasing fuel prices and the existence of thermal storage capability in the CSP plant make this scenario marginally more economically attractive compared to the BAU scenario. (author)

  13. Classifying regularized sensor covariance matrices: An alternative to CSP

    NARCIS (Netherlands)

    Roijendijk, L.M.M.; Gielen, C.C.A.M.; Farquhar, J.D.R.

    2016-01-01

    Common spatial patterns ( CSP) is a commonly used technique for classifying imagined movement type brain-computer interface ( BCI) datasets. It has been very successful with many extensions and improvements on the basic technique. However, a drawback of CSP is that the signal processing pipeline

  14. Classifying regularised sensor covariance matrices: An alternative to CSP

    NARCIS (Netherlands)

    Roijendijk, L.M.M.; Gielen, C.C.A.M.; Farquhar, J.D.R.

    2016-01-01

    Common spatial patterns (CSP) is a commonly used technique for classifying imagined movement type brain computer interface (BCI) datasets. It has been very successful with many extensions and improvements on the basic technique. However, a drawback of CSP is that the signal processing pipeline

  15. Analyses of the use of natural gas in solar power plants (CSP) hybridization in the Sao Francisco Basin (BA); Analise do uso de gas natural na hibridizacao de plantas termosolares (CSP) na Bacia do Sao Francisco (BA)

    Energy Technology Data Exchange (ETDEWEB)

    Malagueta, Diego Cunha; Penafiel, Rafael Andres Soria; Szklo, Alexandre Salem; Dutra, Ricardo M.; Schaeffer, Roberto [Coordenacao dos Programas de Pos-Graduacao de Engenharia (COPPE/UFRJ), RJ (Brazil)

    2012-07-01

    This study assessed the feasibility of Concentrated Solar Power plants (CSP) in Northeast, Brazil. It focused on parabolic trough solar power plants, which is the most mature CSP technology; and evaluated plants rated at 100 MWe, dry cooling systems (due to the low water availability in Northeast), and with and without hybridization based on natural gas (degree of hybridization varying from 25 to 75%). Hence, the capacity factor of the simulated plants hovered between 23 and 98%, according to the degree of hybridization and the choice of the thermodynamic cycle of the natural gas fueled thermal system: Rankine or combined cycle. The CSP plants were simulated at Bom Jesus da Lapa, in the semi-arid region of Bahia. Given the prospects for natural gas resources in the Sao Francisco Basin, different scenarios for the gas prices were tested. Moreover, two scenarios were tested for the cost of the CSP plants, one based on the current financial environment and the other based on incentive policies, such as fiscal incentives and loans. Findings show that while simple plants levelized costs (LCOE) hovered around 520 R$/MWh, for hybrid plants LCOE may reach 140 to 190 R$/MWh. Therefore, this study proposed incentive policies to promote the increasing investment in hybrid CSP plants. (author)

  16. A Precision Photometric Comparison between SDSS-II and CSP Type Ia Supernova Data

    DEFF Research Database (Denmark)

    Mosher, J.; Sako, M.; Corlies, L.

    2012-01-01

    Consistency between Carnegie Supernova Project (CSP) and SDSS-II Supernova Survey ugri measurements has been evaluated by comparing Sloan Digital Sky Survey (SDSS) and CSP photometry for nine spectroscopically confirmed Type Ia supernova observed contemporaneously by both programs. The CSP data...

  17. Initial review and analysis of the direct environmental impacts of CSP in the northern Cape, South Africa

    Science.gov (United States)

    Rudman, Justine; Gauché, Paul; Esler, Karen J.

    2016-05-01

    The Integrated Resource Plan (IRP) of 2010 and the IRP Update provide the most recent guidance to the electricity generation future of South Africa (SA) and both plans include an increased proportion of renewable energy generation capacity. Given that SA has abundant renewable energy resource potential, this inclusion is welcome. Only 600 MW of the capacity allocated to concentrating solar power (CSP) has been committed to projects in the Northern Cape and represents roughly a fifth of the capacity that has been included in the IRP. Although CSP is particularly new in the electricity generation system of the country, the abundant solar resources of the region with annual DNI values of above 2900 kWh/m2 across the arid Savannah and Nama-Karoo biomes offer a promising future for the development of CSP in South Africa. These areas have largely been left untouched by technological development activities and thus renewable energy projects present a variety of possible direct and indirect environmental, social and economic impacts. Environmental Impact Assessments do focus on local impacts, but given that ecological processes often extend to regional- and landscape scales, understanding this scaled context is important to the alignment of development- and conservation priorities. Given the capacities allocated to CSP for the future of SA's electricity generation system, impacts on land, air, water and biodiversity which are associated with CSP are expected to increase in distribution and the understanding thereof deems valuable already from this early point in CSP's future in SA. We provide a review of direct impacts of CSP on the natural environment and an overview of the anticipated specific significance thereof in the Northern Cape.

  18. Constraint satisfaction problems CSP formalisms and techniques

    CERN Document Server

    Ghedira, Khaled

    2013-01-01

    A Constraint Satisfaction Problem (CSP) consists of a set of variables, a domain of values for each variable and a set of constraints. The objective is to assign a value for each variable such that all constraints are satisfied. CSPs continue to receive increased attention because of both their high complexity and their omnipresence in academic, industrial and even real-life problems. This is why they are the subject of intense research in both artificial intelligence and operations research. This book introduces the classic CSP and details several extensions/improvements of both formalisms a

  19. Value as a parameter to consider in operational strategies for CSP plants

    Science.gov (United States)

    de Meyer, Oelof; Dinter, Frank; Govender, Saneshan

    2017-06-01

    This paper introduced a value parameter to consider when analyzing operational strategies for CSP plants. The electric system in South Africa, used as case study, is severely constrained with an influx of renewables in the early phase of deployment. The energy demand curve for the system is analyzed showing the total wind and solar photovoltaic contributions for winter and summer. Due to the intermittent nature and meteorological operating conditions of wind and solar photovoltaic plants, the value of CSP plants within the electric system is introduced. Analyzing CSP plants based on the value parameter alone will remain only a philosophical view. Currently there is no quantifiable measure to translate the philosophical view or subjective value and it solely remains the position of the stakeholder. By introducing three other parameters, Cost, Plant and System to a holistic representation of the Operating Strategies of generation plants, the Value parameter can be translated into a quantifiable measure. Utilizing the country's current procurement program as case study, CSP operating under the various PPA within the Bid Windows are analyzed. The Value Cost Plant System diagram developed is used to quantify the value parameter. This paper concluded that no value is obtained from CSP plants operating under the Bid Window 1 & 2 Power Purchase Agreement. However, by recognizing the dispatchability potential of CSP plants in Bid Window 3 & 3.5, the value of CSP in the electric system can be quantified utilizing Value Added Relationship VCPS-diagram. Similarly ancillary services to the system were analyzed. One of the relationships that have not yet been explored within the industry is an interdependent relationship. It was emphasized that the cost and value structure is shared between the plant and system. Although this relationship is functional when the plant and system belongs to the same entity, additional value is achieved by marginalizing the cost structure. A

  20. Feasibility Study on HYSOL CSP

    DEFF Research Database (Denmark)

    Nielsen, Lars Henrik; Skytte, Klaus; Pérez, Cristian Hernán Cabrera

    2016-01-01

    Concentrating Solar Power (CSP) plants utilize thermal conversion of direct solar irradiation. A trough or tower configuration focuses solar radiation and heats up oil or molten salt that subsequently in high temperature heat exchangers generate steam for power generation. High temperature molten...... salt can be stored and the stored heat can thus increase the load factor and the usability for a CSP plant, e.g. to cover evening peak demand. In the HYSOL concept (HYbrid SOLar) such configuration is extended further to include a gas turbine fuelled by upgraded biogas or natural gas. The optimised...... integrated HYSOL concept, therefore, becomes a fully dispatchable (offering firm power) and fully renewable energy source (RES) based power supply alternative, offering CO2-free electricity in regions with sufficient solar resources. The economic feasibility of HYSOL configurations is addressed in this paper...

  1. A CSP-Based Agent Modeling Framework for the Cougaar Agent-Based Architecture

    Science.gov (United States)

    Gracanin, Denis; Singh, H. Lally; Eltoweissy, Mohamed; Hinchey, Michael G.; Bohner, Shawn A.

    2005-01-01

    Cognitive Agent Architecture (Cougaar) is a Java-based architecture for large-scale distributed agent-based applications. A Cougaar agent is an autonomous software entity with behaviors that represent a real-world entity (e.g., a business process). A Cougaar-based Model Driven Architecture approach, currently under development, uses a description of system's functionality (requirements) to automatically implement the system in Cougaar. The Communicating Sequential Processes (CSP) formalism is used for the formal validation of the generated system. Two main agent components, a blackboard and a plugin, are modeled as CSP processes. A set of channels represents communications between the blackboard and individual plugins. The blackboard is represented as a CSP process that communicates with every agent in the collection. The developed CSP-based Cougaar modeling framework provides a starting point for a more complete formal verification of the automatically generated Cougaar code. Currently it is used to verify the behavior of an individual agent in terms of CSP properties and to analyze the corresponding Cougaar society.

  2. Sandia capabilities for the measurement, characterization, and analysis of heliostats for CSP.

    Energy Technology Data Exchange (ETDEWEB)

    Andraka, Charles E.; Christian, Joshua Mark; Ghanbari, Cheryl M.; Gill, David Dennis; Ho, Clifford Kuofei; Kolb, William J.; Moss, Timothy A.; Smith, Edward J.; Yellowhair, Julius

    2013-07-01

    The Concentrating Solar Technologies Organization at Sandia National Laboratories has a long history of performing important research, development, and testing that has enabled the Concentrating Solar Power Industry to deploy full-scale power plants. Sandia continues to pursue innovative CSP concepts with the goal of reducing the cost of CSP while improving efficiency and performance. In this pursuit, Sandia has developed many tools for the analysis of CSP performance. The following capabilities document highlights Sandias extensive experience in the design, construction, and utilization of large-scale testing facilities for CSP and the tools that Sandia has created for the full characterization of heliostats. Sandia has extensive experience in using these tools to evaluate the performance of novel heliostat designs.

  3. South African CSP projects under the REIPPP programme - Requirements, challenges and opportunities

    Science.gov (United States)

    Relancio, Javier; Cuellar, Alberto; Walker, Gregg; Ettmayr, Chris

    2016-05-01

    Thus far seven Concentrated Solar Power (CSP) projects have been awarded under the Renewable Energy Independent Power Producer Procurement Programme (REIPPPP), totalling 600MW: one project is in operation, four under construction and two on their way to financial close. This provides an excellent opportunity for analysis of key features of the projects that have contributed to or detracted from the programme's success. The paper draws from Mott MacDonald's involvement as Technical Advisor on the seven CSP projects that have been successful under the REIPPPP to date as well as other global CSP developments. It presents how various programme requirements have affected the implementation of projects, such as the technical requirements, time of day tariff structure, economic development requirements and the renewable energy grid code. The increasingly competitive tariffs offered have encouraged developers to investigate efficiency maximising project configurations and cost saving mechanisms, as well as featuring state of the art technology in their proposals. The paper assesses the role of the project participants (developers, lenders and government) with regards to these innovative technologies and solutions. In our paper we discuss the status of projects and the SA market, analysing the main challenges and opportunities that in turn have influenced various aspects such as technology choice, operational regimes and supply chain arrangements.

  4. The techno-economic optimization of a 100MWe CSP-desalination plant in Arandis, Namibia

    Science.gov (United States)

    Dall, Ernest P.; Hoffmann, Jaap E.

    2017-06-01

    Energy is a key factor responsible for a country's economic growth and prosperity. It is closely related to the main global challenges namely: poverty mitigation, global environmental change and food and water security [1.]. Concentrating solar power (CSP) is steadily gaining more market acceptance as the cost of electricity from CSP power plants progressively declines. The cogeneration of electricity and water is an attractive prospect for future CSP developments as the simultaneous production of power and potable water can have positive economic implications towards increasing the feasibility of CSP plant developments [2.]. The highest concentrations of direct normal irradiation are located relatively close to Western coastal and Middle-Eastern North-African regions. It is for this reason worthwhile investigating the possibility of CSP-desalination (CSP+D) plants as a future sustainable method for providing both electricity and water with significantly reduced carbon emissions and potential cost reductions. This study investigates the techno-economic feasibility of integrating a low-temperature thermal desalination plant to serve as the condenser as opposed to a conventional dry-cooled CSP plant in Arandis, Namibia. It outlines the possible benefits of the integration CSP+D in terms of overall cost of water and electricity. The high capital costs of thermal desalination heat exchangers as well as the pumping of seawater far inland is the most significant barrier in making this approach competitive against more conventional desalination methods such as reverse osmosis. The compromise between the lowest levelized cost of electricity and water depends on the sizing and the top brine temperature of the desalination plant.

  5. Low cost anti-soiling coatings for CSP collector mirrors and heliostats

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Barton Barton [ORNL; Polyzos, Georgios [ORNL; Schaeffer, Daniel A [ORNL; Lee, Dominic F [ORNL; Datskos, Panos G [ORNL

    2014-01-01

    Most concentrating solar power (CSP) facilities in the USA are located in the desert southwest of the country where land and sunshine are abundant. But one of the significant maintenance problems and cost associated with operating CSP facilities in this region is the accumulation of dust, sand and other pollutants on the collector mirrors and heliostats. In this paper we describe the development of low cost, easy to apply anti-soiling coatings based on superhydrophobic (SH) functionalized nano silica materials and polymer binders that posses the key requirements necessary to inhibit particulate deposition on and sticking to CSP mirror surfaces, and thereby significantly reducing mirror cleaning costs and facility downtime.

  6. Rhodium(III)-Catalyzed Amidation of Unactivated C(sp(3) )-H Bonds.

    Science.gov (United States)

    Wang, He; Tang, Guodong; Li, Xingwei

    2015-10-26

    Nitrogenation by direct functionalization of C-H bonds represents an important strategy for constructing C-N bonds. Rhodium(III)-catalyzed direct amidation of unactivated C(sp(3) )-H bonds is rare, especially under mild reaction conditions. Herein, a broad scope of C(sp(3) )-H bonds are amidated under rhodium catalysis in high efficiency using 3-substituted 1,4,2-dioxazol-5-ones as the amide source. The protocol broadens the scope of rhodium(III)-catalyzed C(sp(3) )-H activation chemistry, and is applicable to the late-stage functionalization of natural products. © 2015 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  7. Comparing carbon capture and storage (CCS) with concentrating solar power (CSP): Potentials, costs, risks, and barriers

    International Nuclear Information System (INIS)

    Lilliestam, Johan; Bielicki, Jeffrey M.; Patt, Anthony G.

    2012-01-01

    Coal power coupled with Carbon [Dioxide] Capture and Storage (CCS), and Concentrating Solar Power (CSP) technologies are often included in the portfolio of climate change mitigation options intended to decarbonize electricity systems. Both of these technologies can provide baseload electricity, are in early stages of maturity, and have benefits, costs, and obstacles. We compare and contrast CCS applied to coal-fired power plants with CSP. At present, both technologies are more expensive than existing electricity-generating options, but costs should decrease with large-scale deployment, especially in the case of CSP. For CCS, technological challenges still remain, storage risks must be clarified, and regulatory and legal uncertainties remain. For CSP, current challenges include electricity transmission and business models for a rapid and extensive expansion of high-voltage transmission lines. The need for international cooperation may impede CSP expansion in Europe. Highlights: ► Both technologies could provide low-carbon base load power. ► Both technologies require new networks, for either CO 2 or power transmission. ► CSP is closer to being a viable technology ready for pervasive diffusion. ► The costs associated with market saturation would be lower for CSP. ► The regulatory changes required for CSP diffusion are somewhat greater than for CCS.

  8. Control oriented concentrating solar power (CSP) plant model and its applications

    Science.gov (United States)

    Luo, Qi

    Solar receivers in concentrating solar thermal power plants (CSP) undergo over 10,000 start-ups and shutdowns, and over 25,000 rapid rate of change in temperature on receivers due to cloud transients resulting in performance degradation and material fatigue in their expected lifetime of over 30 years. The research proposes to develop a three-level controller that uses multi-input-multi-output (MIMO) control technology to minimize the effect of these disturbances, improve plant performance, and extend plant life. The controller can be readily installed on any vendor supplied state-of-the-art control hardware. We propose a three-level controller architecture using multi-input-multi-output (MIMO) control for CSP plants that can be implemented on existing plants to improve performance, reliability, and extend the life of the plant. This architecture optimizes the performance on multiple time scalesreactive level (regulation to temperature set points), tactical level (adaptation of temperature set points), and strategic level (trading off fatigue life due to thermal cycling and current production). This controller unique to CSP plants operating at temperatures greater than 550 °C, will make CSPs competitive with conventional power plants and contribute significantly towards the Sunshot goal of 0.06/kWh(e), while responding with agility to both market dynamics and changes in solar irradiance such as due to passing clouds. Moreover, our development of control software with performance guarantees will avoid early stage failures and permit smooth grid integration of the CSP power plants. The proposed controller can be implemented with existing control hardware infrastructure with little or no additional equipment. In the thesis, we demonstrate a dynamics model of CSP, of which different components are modelled with different time scales. We also show a real time control strategy of CSP control oriented model in steady state. Furthermore, we shown different controllers

  9. CSP: A Multifaceted Hybrid Architecture for Space Computing

    Science.gov (United States)

    Rudolph, Dylan; Wilson, Christopher; Stewart, Jacob; Gauvin, Patrick; George, Alan; Lam, Herman; Crum, Gary Alex; Wirthlin, Mike; Wilson, Alex; Stoddard, Aaron

    2014-01-01

    Research on the CHREC Space Processor (CSP) takes a multifaceted hybrid approach to embedded space computing. Working closely with the NASA Goddard SpaceCube team, researchers at the National Science Foundation (NSF) Center for High-Performance Reconfigurable Computing (CHREC) at the University of Florida and Brigham Young University are developing hybrid space computers that feature an innovative combination of three technologies: commercial-off-the-shelf (COTS) devices, radiation-hardened (RadHard) devices, and fault-tolerant computing. Modern COTS processors provide the utmost in performance and energy-efficiency but are susceptible to ionizing radiation in space, whereas RadHard processors are virtually immune to this radiation but are more expensive, larger, less energy-efficient, and generations behind in speed and functionality. By featuring COTS devices to perform the critical data processing, supported by simpler RadHard devices that monitor and manage the COTS devices, and augmented with novel uses of fault-tolerant hardware, software, information, and networking within and between COTS devices, the resulting system can maximize performance and reliability while minimizing energy consumption and cost. NASA Goddard has adopted the CSP concept and technology with plans underway to feature flight-ready CSP boards on two upcoming space missions.

  10. CSPBuilder - CSP based Scientific Workflow Modelling

    DEFF Research Database (Denmark)

    Friborg, Rune Møllegaard; Vinter, Brian

    2008-01-01

    This paper introduces a framework for building CSP based applications, targeted for clusters and next generation CPU designs. CPUs are produced with several cores today and every future CPU generation will feature increasingly more cores, resulting in a requirement for concurrency that has not pr...

  11. A General Catalyst for Site-Selective C(sp(3))-H Bond Amination of Activated Secondary over Tertiary Alkyl C(sp(3))-H Bonds.

    Science.gov (United States)

    Scamp, Ryan J; Jirak, James G; Dolan, Nicholas S; Guzei, Ilia A; Schomaker, Jennifer M

    2016-06-17

    The discovery of transition metal complexes capable of promoting general, catalyst-controlled and selective carbon-hydrogen (C-H) bond amination of activated secondary C-H bonds over tertiary alkyl C(sp(3))-H bonds is challenging, as substrate control often dominates when reactive nitrene intermediates are involved. In this letter, we report the design of a new silver complex, [(Py5Me2)AgOTf]2, that displays general and good-to-excellent selectivity for nitrene insertion into propargylic, benzylic, and allylic C-H bonds over tertiary alkyl C(sp(3))-H bonds.

  12. CSP electricity cost evolution and grid parities based on the IEA roadmaps

    International Nuclear Information System (INIS)

    Hernández-Moro, J.; Martínez-Duart, J.M.

    2012-01-01

    The main object of this paper consists in the development of a mathematical closed-form expression for the evaluation, in the period 2010–2050, of the levelized costs of energy (LCOE) of concentrating solar power (CSP) electricity. For this purpose, the LCOE is calculated using a life-cycle cost method, based on the net present value, the discounted cash flow technique and the technology learning curve approach. By this procedure, the LCOE corresponding to CSP electricity is calculated as a function of ten independent variables. Among these parameters, special attention has been put on the evaluation of the available solar resource, the analysis of the IEA predicted values for the cumulative installed capacity, the initial (2010) cost of the system, the discount and learning rates, etc. One significant contribution of our work is that the predicted evolution of the LCOEs strongly depend, not only on the particular values of the cumulative installed capacity function in the targeted years, but mainly on the specific curved time-paths which are followed by this function. The results obtained in this work are shown both graphically and numerically. Finally, the implications that the results could have in energy planning policies and grid parity calculations are discussed. - Highlights: ► A mathematical closed expression has been developed for calculating the evolution of CSP electricity costs. ► Our technique for the prediction of CSP electricity costs and grid parities is based on IEA Roadmaps. ► The time-table (2010–2050) of cumulative installed CSP capacity is key to electricity cost predictions. ► CSP grid parities can occur within next decade for sites with proper solar resources.

  13. On purpose simulation model for molten salt CSP parabolic trough

    Science.gov (United States)

    Caranese, Carlo; Matino, Francesca; Maccari, Augusto

    2017-06-01

    The utilization of computer codes and simulation software is one of the fundamental aspects for the development of any kind of technology and, in particular, in CSP sector for researchers, energy institutions, EPC and others stakeholders. In that extent, several models for the simulation of CSP plant have been developed with different main objectives (dynamic simulation, productivity analysis, techno economic optimization, etc.), each of which has shown its own validity and suitability. Some of those models have been designed to study several plant configurations taking into account different CSP plant technologies (Parabolic trough, Linear Fresnel, Solar Tower or Dish) and different settings for the heat transfer fluid, the thermal storage systems and for the overall plant operating logic. Due to a lack of direct experience of Molten Salt Parabolic Trough (MSPT) commercial plant operation, most of the simulation tools do not foresee a suitable management of the thermal energy storage logic and of the solar field freeze protection system, but follow standard schemes. ASSALT, Ase Software for SALT csp plants, has been developed to improve MSPT plant's simulations, by exploiting the most correct operational strategies in order to provide more accurate technical and economical results. In particular, ASSALT applies MSPT specific control logics for the electric energy production and delivery strategy as well as the operation modes of the Solar Field in off-normal sunshine condition. With this approach, the estimated plant efficiency is increased and the electricity consumptions required for the plant operation and management is drastically reduced. Here we present a first comparative study on a real case 55 MWe Molten Salt Parabolic Trough CSP plant placed in the Tibetan highlands, using ASSALT and SAM (System Advisor Model), which is a commercially available simulation tool.

  14. Energy loss function for biological material: poly(CSP)

    International Nuclear Information System (INIS)

    Fung, A.Y.C.; Zaider, M.

    1994-01-01

    In this paper calculated cross sections are presented for the interaction of electrons with poly(CSP), a single-stranded chain that contains one cytosine sugar phosphate unit in the elementary cell. To model a single strand of helical DNA (e.g. the base stacking), the Watson-Crick model for the geometry of poly(CSP) has been used. The use, for computational simplicity, of a single, rather than a double stranded polynucleotide may be justified on the basis of previous calculations indicating that -to a good approximation - the electronic structure (other than excitation states) of complementary base pairs may be described as a superposition of the corresponding structures of the individual components. (Author)

  15. Photoredox Generated Radicals in Csp2-Csp3 Bond Construction

    Science.gov (United States)

    Primer, David Neal

    The routine application of Csp3-hybridized nucleophiles in cross-coupling has been an ongoing pursuit in the agrochemical, pharmaceutical, and materials science industries for over 40 years. Unfortunately, despite numerous attempts to circumvent the problems associated with alkyl nucleophiles, application of these reagents in transition metal-catalyzed C-C bond-forming reactions has remained largely restricted. In recent years, many chemists have noted the lack of reliable, turnkey reactions that exist for the installation of Csp3-hybridized centers--reactions that would be useful for delivering molecules with enhanced three-dimensional topology and altered chemical properties. As such, a general method for alkyl nucleophile activation in cross-coupling would offer access to a host of compounds inaccessible by other means. From a mechanistic standpoint, the continued failure of alkylmetallics is inherent to the high energy intermediates associated with a traditional transmetalation. To overcome this problem, we have pioneered an alternate, single-electron pathway involving 1) initial oxidation of an alkylmetallic reagent, 2) oxidative alkyl radical capture at a metal center, and 3) subsequent reduction of the metal center to return its initial oxidation state. This series of steps constitutes a formal transmetalation that avoids the energy-demanding steps that plague a traditional anionic approach. Under this enabling paradigm, a host of alkyl precursors (alkyl-trifluoroborates and -silicates) have been generally used in cross-coupling for the first time. In summary, the synergistic use of an Ir photoredox catalyst and a Ni cross-coupling catalyst to mediate the cross-coupling of (hetero)aryl bromides with diverse alkyl radical precursors will be discussed. Methods for coupling various trifluoroborate classes (alpha-alkoxy, alpha-trifluoromethyl, secondary and tertiary alkyl) will be covered, focusing on their complementarity to traditional protocols. Finally, a

  16. Reactor Safety Assessment System

    International Nuclear Information System (INIS)

    Sebo, D.E.; Bray, M.A.; King, M.A.

    1987-01-01

    The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSAS is designed for use at the USNRC Operations Center in the event of a serious incident at a licensed nuclear power plant. RSAS is a situation assessment expert system which uses plant parametric data to generate conclusions for use by the NRC Reactor Safety Team. RSAS uses multiple rule bases and plant specific setpoint files to be applicable to all licensed nuclear power plants in the United States. RSAS currently covers several generic reactor categories and multiple plants within each category

  17. Reactor safety assessment system

    International Nuclear Information System (INIS)

    Sebo, D.E.; Bray, M.A.; King, M.A.

    1987-01-01

    The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSA is designed for use at the USNRC Operations Center in the event of a serious incident at a licensed nuclear power plant. RSAS is a situation assessment expert system which uses plant parametric data to generate conclusions for use by the NRC Reactor Safety Team. RSAS uses multiple rule bases and plant specific setpoint files to be applicable to all licensed nuclear power plants in the United States. RSAS currently covers several generic reactor categories and multiple plants within each category

  18. A VAR2CSA:CSP conjugate capable of inducing dual specificity antibody responses

    DEFF Research Database (Denmark)

    Matondo, Sungwa; Thrane, Susan; Janitzek, Christoph Mikkel

    2017-01-01

    Catcher peptide. The covalent interaction between SpyTag/SpyCatcher enables the formation of DBL1x-DBL2x-ID2a:CSP conjugate vaccine. Immunogenicity and quality of antibody responses induced by the conjugate vaccine, as well as a control CSP-SpyCatcher vaccine, was tested in BALB/c mice.  Results: Serum samples...... obtained from mice immunized with the conjugate vaccine were able to recognize both untagged DBL1x-DBL2x-ID2a as well as CSP antigen. Moreover, the geometric mean anti-CSP antibody titer was 1.9-fold higher in serum (at day 35 and 55 post-first immunization) from mice immunized with the conjugate vaccine......, as compared to mice receiving the control vaccine.  Conclusion: The data obtained in this study serves as proof-of-concept for the simultaneous induction of antibodies directed against individual antigen components in a dual stage anti-malaria vaccine....

  19. BF3·Et2O-promoted cleavage of the Csp-Csp2 bond of 2-propynolphenols/anilines: route to C2-alkenylated benzoxazoles and benzimidazoles.

    Science.gov (United States)

    Song, Xian-Rong; Qiu, Yi-Feng; Song, Bo; Hao, Xin-Hua; Han, Ya-Ping; Gao, Pin; Liu, Xue-Yuan; Liang, Yong-Min

    2015-02-20

    A novel BF3·Et2O-promoted tandem reaction of easily prepared 2-propynolphenols/anilines and trimethylsilyl azide is developed to give C2-alkenylated benzoxazoles and benzimidazoles in moderate to good yields. Most reactions could be accomplished in 30 min at room temperature. This tandem process involves a Csp-Csp2 bond cleavage and a C-N bond formation. Moreover, both tertiary and secondary propargylic alcohols with diverse functional groups were tolerated under the mild conditions.

  20. Biotype Characterization, Developmental Profiling, Insecticide Response and Binding Property of Bemisia tabaci Chemosensory Proteins: Role of CSP in Insect Defense.

    Directory of Open Access Journals (Sweden)

    Guoxia Liu

    Full Text Available Chemosensory proteins (CSPs are believed to play a key role in the chemosensory process in insects. Sequencing genomic DNA and RNA encoding CSP1, CSP2 and CSP3 in the sweet potato whitefly Bemisia tabaci showed strong variation between B and Q biotypes. Analyzing CSP-RNA levels showed not only biotype, but also age and developmental stage-specific expression. Interestingly, applying neonicotinoid thiamethoxam insecticide using twenty-five different dose/time treatments in B and Q young adults showed that Bemisia CSP1, CSP2 and CSP3 were also differentially regulated over insecticide exposure. In our study one of the adult-specific gene (CSP1 was shown to be significantly up-regulated by the insecticide in Q, the most highly resistant form of B. tabaci. Correlatively, competitive binding assays using tryptophan fluorescence spectroscopy and molecular docking demonstrated that CSP1 protein preferentially bound to linoleic acid, while CSP2 and CSP3 proteins rather associated to another completely different type of chemical, i.e. α-pentyl-cinnamaldehyde (jasminaldehyde. This might indicate that some CSPs in whiteflies are crucial to facilitate the transport of fatty acids thus regulating some metabolic pathways of the insect immune response, while some others are tuned to much more volatile chemicals known not only for their pleasant odor scent, but also for their potent toxic insecticide activity.

  1. Villacidro solar demo plant: Integration of small-scale CSP and biogas power plants in an industrial microgrid

    Science.gov (United States)

    Camerada, M.; Cau, G.; Cocco, D.; Damiano, A.; Demontis, V.; Melis, T.; Musio, M.

    2016-05-01

    The integration of small scale concentrating solar power (CSP) in an industrial district, in order to develop a microgrid fully supplied by renewable energy sources, is presented in this paper. The plant aims to assess in real operating conditions, the performance, the effectiveness and the reliability of small-scale concentrating solar power technologies in the field of distributed generation. In particular, the potentiality of small scale CSP with thermal storage to supply dispatchable electricity to an industrial microgrid will be investigated. The microgrid will be realized in the municipal waste treatment plant of the Industrial Consortium of Villacidro, in southern Sardinia (Italy), which already includes a biogas power plant. In order to achieve the microgrid instantaneous energy balance, the analysis of the time evolution of the waste treatment plant demand and of the generation in the existing power systems has been carried out. This has allowed the design of a suitable CSP plant with thermal storage and an electrochemical storage system for supporting the proposed microgrid. At the aim of obtaining the expected energy autonomy, a specific Energy Management Strategy, which takes into account the different dynamic performances and characteristics of the demand and the generation, has been designed. In this paper, the configuration of the proposed small scale concentrating solar power (CSP) and of its thermal energy storage, based on thermocline principle, is initially described. Finally, a simulation study of the entire power system, imposing scheduled profiles based on weather forecasts, is presented.

  2. Synthesis and characterization in monkey of [{sup 11}C]SP203 as a radioligand for imaging brain metabotropic glutamate 5 receptors

    Energy Technology Data Exchange (ETDEWEB)

    Simeon, Fabrice G.; Liow, Jeih-San; Zhang, Yi; Hong, Jinsoo; Gladding, Robert L.; Zoghbi, Sami S.; Innis, Robert B.; Pike, Victor W. [National Institutes of Health, Molecular Imaging Branch, National Institute of Mental Health, Bethesda, MD (United States)

    2012-12-15

    [{sup 18}F]SP203 (3-fluoro-5-(2-(2-([{sup 18}F]fluoromethyl)-thiazol-4-yl)ethynyl)benzonitrile) is an effective high-affinity and selective radioligand for imaging metabotropic 5 receptors (mGluR5) in human brain with PET. To provide a radioligand that may be used for more than one scanning session in the same subject in a single day, we set out to label SP203 with shorter-lived {sup 11}C (t{sub 1/2} = 20.4 min) and to characterize its behavior as a radioligand with PET in the monkey. Iodo and bromo precursors were obtained by cross-coupling 2-fluoromethyl-4-((trimethylsilyl)ethynyl)-1,3-thiazole with 3,5-diiodofluorobenzene and 3,5-dibromofluorobenzene, respectively. Treatment of either precursor with [{sup 11}C]cyanide ion rapidly gave [{sup 11}C]SP203, which was purified with high-performance liquid chromatography. PET was used to measure the uptake of radioactivity in brain regions after injecting [{sup 11}C]SP203 intravenously into rhesus monkeys at baseline and under conditions in which mGluR5 were blocked with 3-[(2-methyl-1,3-thiazol-4-yl)ethynyl]pyridine (MTEP). The emergence of radiometabolites in monkey blood in vitro and in vivo was assessed with radio-HPLC. The stability of [{sup 11}C]SP203 in human blood in vitro was also measured. The iodo precursor gave [{sup 11}C]SP203 in higher radiochemical yield (>98 %) than the bromo precursor (20-52 %). After intravenous administration of [{sup 11}C]SP203 into three rhesus monkeys, radioactivity peaked early in brain (average 12.5 min) with a regional distribution in rank order of expected mGluR5 density. Peak uptake was followed by a steady decline. No radioactivity accumulated in the skull. In monkeys pretreated with MTEP before [{sup 11}C]SP203 administration, radioactivity uptake in brain was again high but then declined more rapidly than in the baseline scan to a common low level. [{sup 11}C]SP203 was unstable in monkey blood in vitro and in vivo, and gave predominantly less lipophilic radiometabolites

  3. Sublethal doses of neonicotinoid imidacloprid can interact with honey bee chemosensory protein 1 (CSP1) and inhibit its function

    International Nuclear Information System (INIS)

    Li, Hongliang; Tan, Jing; Song, Xinmi; Wu, Fan; Tang, Mingzhu; Hua, Qiyun; Zheng, Huoqing; Hu, Fuliang

    2017-01-01

    As a frequently used neonicotinoid insecticide, imidacloprid can impair the chemoreceptive behavior of honey bees even at sublethal doses, while the physiochemical mechanism has not been further revealed. Here, multiple fluorescence spectra, thermodynamic method, and molecular docking were used to study the interaction and the functional inhibition of imidacloprid to the recombinant CSP1 protein in Asian honey bee, Apis cerana. The results showed that the fluorescence intensity (λ em  = 332 nm) of CSP1 could be significantly quenched by imidacloprid in a dynamic mode. During the quenching process, ΔH > 0, ΔS > 0, indicating that the acting forces of imidacloprid with CSP1 are mainly hydrophobic interactions. Synchronous fluorescence showed that the fluorescence of CSP1 was mainly derived from tryptophan, and the hydrophobicity of tryptophan decreased with the increase of imidacloprid concentration. Molecular docking predicted the optimal pose and the amino acid composition of the binding process. Circular dichroism (CD) spectra showed that imidacloprid reduced the α-helix of CSP1 and caused the extension of the CSP1 peptide chain. In addition, the binding of CSP1 to floral scent β-ionone was inhibited by nearly 50% of the apparent association constant (K A ) in the presence of 0.28–2.53 ng/bee of imidacloprid, and the inhibition rate of nearly 95% at 3.75 ng/bee of imidacloprid at sublethal dose level. This study initially revealed the molecular physiochemical mechanism that sublethal doses of neonicotinoid still interact and inhibit the physiological function of the honey bees' chemoreceptive system. - Highlights: • Sublethal doses of imidacloprid can directly interact with CSP1 in Apis cerana. • Sublethal imidacloprid can inhibit the function of CSP1 binding to semiochemicals. • The fluorescence intensity of CSP1 quenched by imidacloprid in a dynamic mode. • The binding between CSP1 and imidacloprid are driven by hydrophobic interactions.

  4. BdorCSP2 is important for antifeed and oviposition-deterring activities induced by Rhodojaponin-III against Bactrocera dorsalis.

    Directory of Open Access Journals (Sweden)

    Xin Yi

    Full Text Available Rhodojaponin-III is a nonvolatile botanical grayanoid diterpene compound, which has antifeedant and oviposition deterrence effects against many kinds of insects. However, the molecular mechanism of the chemoreception process remains unknown. In this study, the important role of BdorCSP2 in the recognition of Rhodojaponin-III was identified. The full length cDNA encoding BdorCSP2 was cloned from legs of Bactrocera dorsalis. The results of expression pattern revealed that BdorCSP2 was abundantly expressed in the legs of adult B. dorsalis. Moreover, the expression of BdorCSP2 could be up-regulated by Rhodojaponin-III. In order to gain comprehensive understanding of the recognition process, the binding affinity between BdorCSP2 and Rhodojaponin-III was measured by fluorescence binding assay. Silencing the expression of BdorCSP2 through the ingestion of dsRNA could weaken the effect of oviposition deterrence and antifeedant of Rhodojaponin-III. These results suggested that BdorCSP2 of B. dorsalis could be involved in chemoreception of Rhodojaponin-III and played a critical role in antifeedant and oviposition behaviors induced by Rhodojaponin-III.

  5. Tårs 10000 m2 CSP + Flat Plate Solar Collector Plant - Cost-Performance Optimization of the Design

    DEFF Research Database (Denmark)

    Perers, Bengt; Furbo, Simon; Tian, Zhiyong

    2016-01-01

    , was established. The optimization showed that there was a synergy in combining CSP and FP collectors. Even though the present cost per m² of the CSP collectors is high, the total energy cost is minimized by installing a combination of collectors in such solar heating plant. It was also found that the CSP......A novel solar heating plant with Concentrating Solar Power (CSP) collectors and Flat Plate (FP) collectors has been put into operation in Tårs since July 2015. To investigate economic performance of the plant, a TRNSYS-Genopt model, including a solar collector field and thermal storage tank...

  6. The role of CSP in the electricity system of South Africa - technical operation, grid constraints, market structure and economics

    Science.gov (United States)

    Kost, Christoph; Friebertshäuser, Chris; Hartmann, Niklas; Fluri, Thomas; Nitz, Peter

    2017-06-01

    This paper analyses the role of solar technologies (CSP and PV) and their interaction in the South African electricity system by using a fundamental electricity system modelling (ENTIGRIS-SouthAfrica). The model is used to analyse the South African long-term electricity generation portfolio mix, optimized site selection and required transmission capacities until the year 2050. Hereby especially the location and grid integration of solar technology (PV and CSP) and wind power plants is analysed. This analysis is carried out by using detailed resource assessment of both technologies. A cluster approach is presented to reduce complexity by integrating the data in an optimization model.

  7. A New Method to Extract CSP Gather of Topography for Scattered Wave Imaging

    Directory of Open Access Journals (Sweden)

    Zhao Pan

    2017-01-01

    Full Text Available The seismic method is one of the major geophysical tools to study the structure of the earth. The extraction of the common scatter point (CSP gather is a critical step to accomplish the seismic imaging with a scattered wave. Conventionally, the CSP gather is obtained with the assumption that the earth surface is horizontal. However, errors are introduced to the final imaging result if the seismic traces obtained at the rugged surface are processed using the conventional method. Hence, we propose the method of the extraction of the CSP gather for the seismic data collected at the rugged surface. The proposed method is validated by two numerical examples and expected to reduce the effect of the topography on the scattered wave imaging.

  8. Neuroprotective and Anti-Apoptotic Effects of CSP-1103 in Primary Cortical Neurons Exposed to Oxygen and Glucose Deprivation.

    Science.gov (United States)

    Porrini, Vanessa; Sarnico, Ilenia; Benarese, Marina; Branca, Caterina; Mota, Mariana; Lanzillotta, Annamaria; Bellucci, Arianna; Parrella, Edoardo; Faggi, Lara; Spano, Pierfranco; Imbimbo, Bruno Pietro; Pizzi, Marina

    2017-01-18

    CSP-1103 (formerly CHF5074) has been shown to reverse memory impairment and reduce amyloid plaque as well as inflammatory microglia activation in preclinical models of Alzheimer's disease. Moreover, it was found to improve cognition and reduce brain inflammation in patients with mild cognitive impairment. Recent evidence suggests that CSP-1103 acts through a single molecular target, the amyloid precursor protein intracellular domain (AICD), a transcriptional regulator implicated in inflammation and apoptosis. We here tested the possible anti-apoptotic and neuroprotective activity of CSP-1103 in a cell-based model of post-ischemic injury, wherein the primary mouse cortical neurons were exposed to oxygen-glucose deprivation (OGD). When added after OGD, CSP-1103 prevented the apoptosis cascade by reducing cytochrome c release and caspase-3 activation and the secondary necrosis. Additionally, CSP-1103 limited earlier activation of p38 and nuclear factor κB (NF-κB) pathways. These results demonstrate that CSP-1103 is neuroprotective in a model of post-ischemic brain injury and provide further mechanistic insights as regards its ability to reduce apoptosis and potential production of pro-inflammatory cytokines. In conclusion, these findings suggest a potential use of CSP-1103 for the treatment of brain ischemia.

  9. RSAS: a Reactor Safety Assessment System

    International Nuclear Information System (INIS)

    Sebo, D.E.; Dixon, B.W.; Bray, M.A.

    1985-01-01

    The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (NRC). RSAS is being developed for use at the NRC's Operations Center in the event of a serious incident at a licensed nuclear power plant. The system generates situation assessments for the NRC Reactor Safety Team based on a limited number of plant parameters, known operator actions, and plant status data. The RSAS rule base currently covers one reactor type. The extension of the rule base to other reactor types is also discussed

  10. Safety assessment of Department of Energy nuclear reactors

    International Nuclear Information System (INIS)

    1981-03-01

    One of the first tasks of the NFPQT Committee was to determine which DOE reactors would be assessed. The Committee determined that in view of the limited time available to conduct the assessment, 13 DOE reactors were of such size (physical, power or fission product inventory) to warrant review. This determination was approved by the Under Secretary. A decision was also made in the cases of three weapons material production reactors, C, K and P, to concentrate on the K reactor only, since all three are of the same basic design, have the same operating features, are all at the same site, and are all operated by the same contractor. The assessment was accomplished in the following ways: reviewing the results of assessments conducted by the DOE organizations with reactor safety responsibilities, which were undertaken in compliance with the request of the various program directors; reviewing selected documents that were requested by the Committee and assembled at DOE Headquarters; interviewing DOE Headquarters and Field Office personnel; and conducting on-site reviews of four reactors located at four different sites. The four reactors for on-site reviews were: Advanced Test Reactor (ATR); K Production Reactor; High Flux Beam Reactor (HFBR); and High Flux Isotope Reactor (HFIR). Specific findings and recommendations from the assessment are presented

  11. The Climate Services Partnership (CSP): Working Together to Improve Climate Services Worldwide

    Science.gov (United States)

    Zebiak, S.; Brasseur, G.; Members of the CSP Coordinating Group

    2012-04-01

    -searchable database that allows users to see what climate services activities are underway in what locations, to gather and analyze information. As part of the knowledge capture system, more than 10 CSP members are currently developing case studies to describe specific climate services activities; in a few cases, this involves in-depth evaluations of the service in question. Finally, the Economics Working Group of the Climate Services Partnership is analyzing previous methods to economically value climate services in hopes of generating knew knowledge regarding the methods are best suited to assessing the benefits associated with various climate services. Other groups are working to develop guidance materials for the development and use of climate information to support decision and policy-making. The Climate Services Partnership is an open, informal network that builds on activities that are already underway and works to create synergies to improve the provision and development for climate services. Its members currently number more than 50 organizations; it seeks new participants and new initiatives.

  12. Addressing forecast uncertainty impact on CSP annual performance

    Science.gov (United States)

    Ferretti, Fabio; Hogendijk, Christopher; Aga, Vipluv; Ehrsam, Andreas

    2017-06-01

    This work analyzes the impact of weather forecast uncertainty on the annual performance of a Concentrated Solar Power (CSP) plant. Forecast time series has been produced by a commercial forecast provider using the technique of hindcasting for the full year 2011 in hourly resolution for Ouarzazate, Morocco. Impact of forecast uncertainty has been measured on three case studies, representing typical tariff schemes observed in recent CSP projects plus a spot market price scenario. The analysis has been carried out using an annual performance model and a standard dispatch optimization algorithm based on dynamic programming. The dispatch optimizer has been demonstrated to be a key requisite to maximize the annual revenues depending on the price scenario, harvesting the maximum potential out of the CSP plant. Forecasting uncertainty affects the revenue enhancement outcome of a dispatch optimizer depending on the error level and the price function. Results show that forecasting accuracy of direct solar irradiance (DNI) is important to make best use of an optimized dispatch but also that a higher number of calculation updates can partially compensate this uncertainty. Improvement in revenues can be significant depending on the price profile and the optimal operation strategy. Pathways to achieve better performance are presented by having more updates both by repeatedly generating new optimized trajectories but also more often updating weather forecasts. This study shows the importance of working on DNI weather forecasting for revenue enhancement as well as selecting weather services that can provide multiple updates a day and probabilistic forecast information.

  13. Economic opportunities resulting from a global deployment of concentrated solar power (CSP) technologies-The example of German technology providers

    International Nuclear Information System (INIS)

    Vallentin, Daniel; Viebahn, Peter

    2010-01-01

    Several energy scenario studies consider concentrated solar power (CSP) plants as an important technology option to reduce the world's CO 2 emissions to a level required for not letting the global average temperature exceed a threshold of 2-2.4 o C. A global ramp up of CSP technologies offers great economic opportunities for technology providers as CSP technologies include highly specialised components. This paper analyses possible value creation effects resulting from a global deployment of CSP until 2050 as projected in scenarios of the International Energy Agency (IEA) and Greenpeace International. The analysis focuses on the economic opportunities of German technology providers since companies such as Schott Solar, Flabeg or Solar Millennium are among the leading suppliers of CSP technologies on the global market.

  14. Three Unique Implementations of Processes for PyCSP

    DEFF Research Database (Denmark)

    Friborg, Rune Møllegaard; Bjørndalen, John Markus; Vinter, Brian

    2009-01-01

    In this work we motivate and describe three unique implementations of processes for PyCSP: process, thread and greenlet based. The overall purpose is to demonstrate the feasibility of Communicating Sequential Processes as a framework for different application types and target platforms. The result...

  15. A PRECISION PHOTOMETRIC COMPARISON BETWEEN SDSS-II AND CSP TYPE Ia SUPERNOVA DATA

    International Nuclear Information System (INIS)

    Mosher, J.; Sako, M.; Corlies, L.; Folatelli, G.; Frieman, J.; Kessler, R.; Holtzman, J.; Jha, S. W.; Marriner, J.; Phillips, M. M.; Morrell, N.; Stritzinger, M.; Schneider, D. P.

    2012-01-01

    Consistency between Carnegie Supernova Project (CSP) and SDSS-II Supernova Survey ugri measurements has been evaluated by comparing Sloan Digital Sky Survey (SDSS) and CSP photometry for nine spectroscopically confirmed Type Ia supernova observed contemporaneously by both programs. The CSP data were transformed into the SDSS photometric system. Sources of systematic uncertainty have been identified, quantified, and shown to be at or below the 0.023 mag level in all bands. When all photometry for a given band is combined, we find average magnitude differences of equal to or less than 0.011 mag in ugri, with rms scatter ranging from 0.043 to 0.077 mag. The u-band agreement is promising, with the caveat that only four of the nine supernovae are well observed in u and these four exhibit an 0.038 mag supernova-to-supernova scatter in this filter.

  16. Ligand-Promoted C(sp(3) )-H Olefination en Route to Multi-functionalized Pyrazoles.

    Science.gov (United States)

    Yang, Weibo; Ye, Shengqing; Schmidt, Yvonne; Stamos, Dean; Yu, Jin-Quan

    2016-05-17

    A Pd-catalyzed/N-heterocycle-directed C(sp(3) )-H olefination has been developed. The monoprotected amino acid ligand (MPAA) is found to significantly promote Pd-catalyzed C(sp(3) )-H olefination for the first time. Cu(OAc)2 instead of Ag(+) salts are used as the terminal oxidant. This reaction provides a useful method for the synthesis of alkylated pyrazoles. © 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  17. Hardware support for CSP on a Java chip multiprocessor

    DEFF Research Database (Denmark)

    Gruian, Flavius; Schoeberl, Martin

    2013-01-01

    Due to memory bandwidth limitations, chip multiprocessors (CMPs) adopting the convenient shared memory model for their main memory architecture scale poorly. On-chip core-to-core communication is a solution to this problem, that can lead to further performance increase for a number of multithreaded...... applications. Programmatically, the Communicating Sequential Processes (CSPs) paradigm provides a sound computational model for such an architecture with message based communication. In this paper we explore hardware support for CSP in the context of an embedded Java CMP. The hardware support for CSP are on......-chip communication channels, implemented by a ring-based network-on-chip (NoC), to reduce the memory bandwidth pressure on the shared memory.The presented solution is scalable and also specific for our limited resources and real-time predictability requirements. CMP architectures of three to eight processors were...

  18. Measurements of mirror soiling at a candidate CSP site

    CSIR Research Space (South Africa)

    Griffith, DJ

    2013-09-01

    Full Text Available Loss of mirror reflectivity due to soiling at Concentrated Solar Power (CSP) plants is a significant consideration for design and operation of the plant. Increasingly, a bankable case for establishment of a new plant will include an evaluation...

  19. Modelling of a Coil Steam Generator for CSP applications

    DEFF Research Database (Denmark)

    Pelagotti, Leonardo; Sørensen, Kim; Condra, Thomas Joseph

    2014-01-01

    The project investigates a new design for a CSP plant steam generation system, the Coil Steam Generator (CSG). This system allows faster start-ups and therefore higher daily energy production from the Sun. An analytical thermodynamic simulation model of the evaporator and a mechanical analysis...

  20. Purification of cold-shock-like proteins from Stigmatella aurantiaca - molecular cloning and characterization of the cspA gene.

    Science.gov (United States)

    Stamm, I; Leclerque, A; Plaga, W

    1999-09-01

    Prominent low-molecular-weight proteins were isolated from vegetative cells of the myxobacterium Stigmatella aurantiaca and were found to be members of the cold-shock protein family. A first gene of this family (cspA) was cloned and sequenced. It encodes a protein of 68 amino acid residues that displays up to 71% sequence identity with other bacterial cold-shock(-like) proteins. A cysteine residue within the RNP-2 motif is a peculiarity of Stigmatella CspA. A cspA::(Deltatrp-lacZ) fusion gene construct was introduced into Stigmatella by electroporation, a method that has not been used previously for this strain. Analysis of the resultant transformants revealed that cspA transcription occurs at high levels during vegetative growth at 20 and 32 degrees C, and during fruiting body formation.

  1. Test reactor risk assessment methodology

    International Nuclear Information System (INIS)

    Jennings, R.H.; Rawlins, J.K.; Stewart, M.E.

    1976-04-01

    A methodology has been developed for the identification of accident initiating events and the fault modeling of systems, including common mode identification, as these methods are applied in overall test reactor risk assessment. The methods are exemplified by a determination of risks to a loss of primary coolant flow in the Engineering Test Reactor

  2. TGGs for Transforming UML to CSP

    DEFF Research Database (Denmark)

    Greenyer, Joel; Kindler, Ekkart; Rieke, Jan

    Contest. The second transformation problem, a transformation from UML activity diagrams to CSP processes, i.e. a transformation between two models, is a typical application for Triple Graph Grammars (TGGs). We present our contributed solution, presenting the TGG rules and the implementation of our TGG...... interpreter. Moreover, we point out the advantages of our soulution as well as some restrictions of the current implementation. This paper will only briefly state the transformation problem and focus on our TGG approach and the discussion of the rules....

  3. An interplay among FIS, H-NS and guanosine tetraphosphate modulates transcription of the Escherichia coli cspA gene under physiological growth conditions

    Directory of Open Access Journals (Sweden)

    Anna eBrandi

    2016-05-01

    Full Text Available CspA, the most characterized member of the csp gene family of Escherichia coli, is highly expressed not only in response to cold stress, but also during the early phase of growth at 37°C. Here, we investigate at molecular level the antagonistic role played by the nucleoid proteins FIS and H-NS in the regulation of cspA expression under non-stress conditions. By means of both probing experiments and immunological detection, we demonstrate in vitro the existence of binding sites for these proteins on the cspA regulatory region, in which FIS and H-NS bind simultaneously to form composite DNA-protein complexes. While the in vitro promoter activity of cspA is stimulated by FIS and repressed by H-NS, a compensatory effect is observed when both proteins are added in the transcription assay. Consistently with these findings, inactivation of fis and hns genes reversely affect the in vivo amount of cspA mRNA. In addition, by means of strains expressing a high level of the alarmone guanosine tetraphosphate ((pppGpp and in vitro transcription assays, we show that the cspA promoter is sensitive to (pppGpp inhibition. The (pppGpp-mediated expression of fis and hns genes is also analyzed, thus clarifying some aspects of the regulatory loop governing cspA transcription.

  4. Concentrating Solar Power: Best Practices Handbook for the Collection and Use of Solar Resource Data (CSP)

    Energy Technology Data Exchange (ETDEWEB)

    Stoffel, T.; Renne, D.; Myers, D.; Wilcox, S.; Sengupta, M.; George, R.; Turchi, C.

    2010-09-01

    As the world looks for low-carbon sources of energy, solar power stands out as the most abundant energy resource. Harnessing this energy is the challenge for this century. Photovoltaics and concentrating solar power (CSP) are two primary forms of electricity generation using sunlight. These use different technologies, collect different fractions of the solar resource, and have different siting and production capabilities. Although PV systems are most often deployed as distributed generation sources, CSP systems favor large, centrally located systems. Accordingly, large CSP systems require a substantial investment, sometimes exceeding $1 billion in construction costs. Before such a project is undertaken, the best possible information about the quality and reliability of the fuel source must be made available. That is, project developers need to have reliable data about the solar resource available at specific locations to predict the daily and annual performance of a proposed CSP plant. Without these data, no financial analysis is possible. This handbook presents detailed information about solar resource data and the resulting data products needed for each stage of the project.

  5. A CSP plant combined with biomass CHP using ORC-technology in Bronderslev Denmark

    DEFF Research Database (Denmark)

    Perers, Bengt; Furbo, Simon; Yuan, Guofeng

    2017-01-01

    A new CSP plant combined with biomass CHP, using ORC technology, will be built and taken into operation in Bronderslev, Denmark during spring 2017. The price for Biomass is expected to increase with more and more use of this very limited energy source and then CSP will be cost effective in the long...... run, also in the Danish climate. Oil is used as heat transfer fluid instead of steam giving several advantages in this application for district heating at high latitudes. Total efficiencies and costs, competitive to PV plants. are expected....

  6. Chromobacterium Csp_P reduces malaria and dengue infection in vector mosquitoes and has entomopathogenic and in vitro anti-pathogen activities.

    Science.gov (United States)

    Ramirez, Jose Luis; Short, Sarah M; Bahia, Ana C; Saraiva, Raul G; Dong, Yuemei; Kang, Seokyoung; Tripathi, Abhai; Mlambo, Godfree; Dimopoulos, George

    2014-10-01

    Plasmodium and dengue virus, the causative agents of the two most devastating vector-borne diseases, malaria and dengue, are transmitted by the two most important mosquito vectors, Anopheles gambiae and Aedes aegypti, respectively. Insect-bacteria associations have been shown to influence vector competence for human pathogens through multi-faceted actions that include the elicitation of the insect immune system, pathogen sequestration by microbes, and bacteria-produced anti-pathogenic factors. These influences make the mosquito microbiota highly interesting from a disease control perspective. Here we present a bacterium of the genus Chromobacterium (Csp_P), which was isolated from the midgut of field-caught Aedes aegypti. Csp_P can effectively colonize the mosquito midgut when introduced through an artificial nectar meal, and it also inhibits the growth of other members of the midgut microbiota. Csp_P colonization of the midgut tissue activates mosquito immune responses, and Csp_P exposure dramatically reduces the survival of both the larval and adult stages. Ingestion of Csp_P by the mosquito significantly reduces its susceptibility to Plasmodium falciparum and dengue virus infection, thereby compromising the mosquito's vector competence. This bacterium also exerts in vitro anti-Plasmodium and anti-dengue activities, which appear to be mediated through Csp_P -produced stable bioactive factors with transmission-blocking and therapeutic potential. The anti-pathogen and entomopathogenic properties of Csp_P render it a potential candidate for the development of malaria and dengue control strategies.

  7. Intramolecular apical metal-H-Csp3 interaction in molybdenum and silver complexes.

    Science.gov (United States)

    Ciclosi, Marco; Lloret, Julio; Estevan, Francisco; Sanaú, Mercedes; Pérez-Prieto, Julia

    2009-07-14

    The reaction of HTIMP3 (HTIMP3=tris[1-diphenylphosphino)-3-methyl-1H-indol-2-yl]methane) with AgBF4 and Mo(CO)3(NCCH3)3 leads to Ag(HTIMP3)BF4 and Mo(CO)3(HTIMP3), respectively. The metal centre is coordinated to the three phosphorus atoms of the HTIMP3 ligand, which adopts a facial coordination mode, placing a H-Csp3 hydrogen atom at the apical position close to the metal centre. The solid-state structure of Mo(CO)3(HTIMP3) has been determined by X-ray crystallography, and the data have been used as input parameters for obtaining the optimised geometry of the complex using the B3PW91 functional. The silver structure has been modelled from the X-ray parameters of the molybdenum structure. In addition, theoretical calculations on the H-Csp3 downfield shift upon metal coordination has also been performed. They reproduce the experimental H-Csp3 chemical shifts well and supports that proton deshielding is mainly due to the presence of the metal, since the hydrogen is already located in the cone created by the aromatic-phosphino arms in the free ligand.

  8. Analysis of Methodologies for Identifying Exclusion Zones for Concentrating Solar Power (CSP); Analisis de Metodologias de Identificacion de Zonas de Exclusion para Estudios de Potencial de Energia Electrica Termosolar (CSP)

    Energy Technology Data Exchange (ETDEWEB)

    Dominguez, P.; Ramirez, L.; Navarro, A. A.; Polo, J.; Zarza, E.

    2013-07-01

    The aim of this study is the proposal of a valid and unique methodology to any territory of the potential for solar power generation, reducing subjectivity and enabling comparison of results from the examination of several existing methodologies for CSP, particularly those developed by the Institute for diversification and saving of Energy (IDAE), Greenpeace, National renewable energy laboratory (NREL) and the German Aerospace Center (DLR). Subsequently, we apply and compare the results obtained with those already installed CSP plants, giving an idea of the suitability of each methodology to locate plants in areas considered suitable. (Author)

  9. Primary fibroblasts from CSP? mutation carriers recapitulate hallmarks of the adult onset neuronal ceroid lipofuscinosis

    OpenAIRE

    Benitez, Bruno A.; Sands, Mark S.

    2017-01-01

    Mutations in the co- chaperone protein, CSP?, cause an autosomal dominant, adult-neuronal ceroid lipofuscinosis (AD-ANCL). The current understanding of CSP? function exclusively at the synapse fails to explain the autophagy-lysosome pathway (ALP) dysfunction in cells from AD-ANCL patients. Here, we demonstrate unexpectedly that primary dermal fibroblasts from pre-symptomatic mutation carriers recapitulate in vitro features found in the brains of AD-ANCL patients including auto-fluorescent sto...

  10. Sublethal doses of neonicotinoid imidacloprid can interact with honey bee chemosensory protein 1 (CSP1) and inhibit its function.

    Science.gov (United States)

    Li, Hongliang; Tan, Jing; Song, Xinmi; Wu, Fan; Tang, Mingzhu; Hua, Qiyun; Zheng, Huoqing; Hu, Fuliang

    2017-04-29

    As a frequently used neonicotinoid insecticide, imidacloprid can impair the chemoreceptive behavior of honey bees even at sublethal doses, while the physiochemical mechanism has not been further revealed. Here, multiple fluorescence spectra, thermodynamic method, and molecular docking were used to study the interaction and the functional inhibition of imidacloprid to the recombinant CSP1 protein in Asian honey bee, Apis cerana. The results showed that the fluorescence intensity (λ em  = 332 nm) of CSP1 could be significantly quenched by imidacloprid in a dynamic mode. During the quenching process, ΔH > 0, ΔS > 0, indicating that the acting forces of imidacloprid with CSP1 are mainly hydrophobic interactions. Synchronous fluorescence showed that the fluorescence of CSP1 was mainly derived from tryptophan, and the hydrophobicity of tryptophan decreased with the increase of imidacloprid concentration. Molecular docking predicted the optimal pose and the amino acid composition of the binding process. Circular dichroism (CD) spectra showed that imidacloprid reduced the α-helix of CSP1 and caused the extension of the CSP1 peptide chain. In addition, the binding of CSP1 to floral scent β-ionone was inhibited by nearly 50% of the apparent association constant (K A ) in the presence of 0.28-2.53 ng/bee of imidacloprid, and the inhibition rate of nearly 95% at 3.75 ng/bee of imidacloprid at sublethal dose level. This study initially revealed the molecular physiochemical mechanism that sublethal doses of neonicotinoid still interact and inhibit the physiological function of the honey bees' chemoreceptive system. Copyright © 2017 Elsevier Inc. All rights reserved.

  11. Safety re-assessment of AECL test and research reactors

    International Nuclear Information System (INIS)

    Winfield, D.J.

    1990-01-01

    Atomic Energy of Canada Limited currently has four operating engineering test/research reactors of various sizes and ages; a new isotope-production reactor Maple-X10, under construction at Chalk River Nuclear Laboratories (CRNL), and a heating demonstration reactor, SDR, undergoing high-power commissioning at Whiteshell Nuclear Research Establishment (WNRE). The company is also performing design studies of small reactors for hot water and electricity production. The older reactors are ZED-2, PTR, NRX, and NRU; these range in age from 42 years (NRX) to 29 years (ZED-2). Since 1984, limited-scope safety re-assessments have been underway on three of these reactors (ZED-2, NRX AND NRU). ZED-2 and PTR are operated by the Reactor Physics Branch; all other reactors are operated by the respective site Reactor Operations Branches. For the older reactors the original safety reports produced were entirely deterministic in nature and based on the design-basis accident concept. The limited scope safety re-assessments for these older reactors, carried out over the past 5 years, have comprised both quantitative probabilistic safety-assessment techniques, such as event tree and fault analysis, and/or qualitative techniques, such as failure mode and effect analysis. The technique used for an individual assessment was dependent upon the specific scope required. This paper discusses the types of analyses carried out, specific insights/recommendations resulting from the analysis, and the plan for future analysis. In addition, during the last four years safety assessments have been carried out on the new isotope-, heat-, and electricity-producing reactors, as part of the safety design review, commissioning and licensing activities

  12. Probabilistic safety assessment for research reactors

    International Nuclear Information System (INIS)

    1986-12-01

    Increasing interest in using Probabilistic Safety Assessment (PSA) methods for research reactor safety is being observed in many countries throughout the world. This is mainly because of the great ability of this approach in achieving safe and reliable operation of research reactors. There is also a need to assist developing countries to apply Probabilistic Safety Assessment to existing nuclear facilities which are simpler and therefore less complicated to analyse than a large Nuclear Power Plant. It may be important, therefore, to develop PSA for research reactors. This might also help to better understand the safety characteristics of the reactor and to base any backfitting on a cost-benefit analysis which would ensure that only necessary changes are made. This document touches on all the key aspects of PSA but placed greater emphasis on so-called systems analysis aspects rather than the in-plant or ex-plant consequences

  13. Risk-assessment methodology for fast breeder reactors

    International Nuclear Information System (INIS)

    Ott, K.O.

    1976-04-01

    The methods applied or proposed for risk assessment of nuclear reactors are reviewed, particularly with respect to their applicability for risk assessment of future commercial fast breeder reactors. All methods are based on the calculation of accident consequences for relatively few accident scenarios. The role and general impact of uncertainties in fast-reactor accident analysis are discussed. The discussion shows the need for improvement of the methodology. A generalized and improved risk-assessment methodology is outlined and proposed (accident-spectra-progression approach). The generalization consists primarily of an explicit treatment of uncertainties throughout the accident progression. The results of this method are obtained in form of consequence distributions. The width and shape of the distributions depend in part on the superposition of the uncertainties. The first moment of the consequence distribution gives an improved prediction of the ''average'' consequence. The higher-consequence moments can be used for consideration of risk aversion. The assessment of the risk of one or a certain number of nuclear reactors can only provide an ''isolated'' risk assessment. The general problem of safety risk assessment and its relation to public acceptance of certain modes of power production is a much broader problem area, which is also discussed

  14. Bounded Delay Timing Analysis of a Class of CSP Programs

    DEFF Research Database (Denmark)

    Hulgaard, Henrik; Burns, Steven M.

    1997-01-01

    We describe an algebraic technique for performing timing analysis of a class of asynchronous circuits described as CSP programs (including Martin's probe operator) with the restrictions that there is no OR-causality and that guard selection is either completely free or mutually exclusive...

  15. Phenomenological Studies on Sodium for CSP Applications: A Safety Review

    Energy Technology Data Exchange (ETDEWEB)

    Armijo, Kenneth Miguel [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Concentrating Solar Technologies Dept.; Andraka, Charles E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Concentrating Solar Technologies Dept.

    2016-09-01

    Sodium as a heat transfer fluid (HTF) can achieve temperatures above 700°C to improve power cycle performance for reducing large infrastructure costs of high-temperature systems. Current concentrating solar power (CSP) sensible HTF’s (e.g. air, salts) have poor thermal conductivity, and thus low heat transfer capabilities, requiring a large receiver. The high thermal conductivity of sodium has demonstrated high heat transfer rates on dish and towers systems, which allow a reduction in receiver area by a factor of two to four, reducing re-radiation and convection losses and cost by a similar factor. Sodium produces saturated vapor at pressures suitable for transport starting at 600°C and reaches one atmosphere at 870°C, providing a wide range of suitable latent operating conditions that match proposed high temperature, isothermal input power cycles. This advantage could increase the receiver and system efficiency while lowering the cost of CSP tower systems. Although there are a number of desirable thermal performance advantages associated with sodium, its propensity to rapidly oxidize presents safety challenges. This investigation presents a literature review that captures historical operations/handling lessons for advanced sodium systems, and the current state-of-knowledge related to sodium combustion behavior. Technical and operational solutions addressing sodium safety and applications in CSP will be discussed, including unique safety hazards and advantages using latent sodium. Operation and maintenance experience from the nuclear industry with sensible and latent systems will also be discussed in the context of safety challenges and risk mitigation solutions.

  16. The Moroccan solar plan. A comparative analysis of CSP and PV utilization until 2020

    International Nuclear Information System (INIS)

    Richts, Christoph

    2012-01-01

    The present master thesis conducts technical and economic simulations of large-scale Photovoltaic (PV) and Concentrated Solar Power (CSP) plants for the Moroccan Solar Plan. It provides a database of performance indicators such as energy yields, capacity factors, typical efficiencies and losses of technical components, LCOE, and difference costs (DC: LCOE minus avoided costs of the conventional power system) for fixed tilted, 1-axis horizontal, 1-axis vertical and 2-axis tracking PV and CSP with no, 6, 12 and 18 full load hours of thermal storage. HelioClim irradiation data of 2005 for the sites in Ouarzazate, Ain Ben Mathar, Boujdour, Laayoune and Tarfaya is used ranging between 1,927 - 2,428 kWh/m 2 /y (DNI) and 1,968 - 2,154 kWh/m 2 /y (GHI). In the base scenario minimum LCOE are 9.6 - 5.4 EURct/kWh for PV (2012 - 2020) varying between 0.90 - 1.55 EURct/kWh among sites and technologies. CSP reaches 12.8 - 9.2 EURct/kWh and a bandwidth of 2.3 - 1.6 EURct/kWh. Average DC are lowest for horizontal 1-axis tracking (0.4 and -7.7 EURct/kWh for plants built in 2012 and 2020 respectively) and CSP with 6 hours of storage (1.3 and -3.5 EURct/kWh). PV is cheaper for all sites and technologies due to higher learning curves and less initial investment, but cannot contribute to coverage of the daily evening peak in Morocco. Four different MSP-scenarios with 2000 MW of solar energy require total investments of 3.7 - 7.5 billion EUR and yield 7.9% - 12.8% of the electricity demand in 2020 (given a growth 7%/y) depending on the ratio of PV and CSP utilization. The average LCOE are 8.3 - 11.7 EURct/kWh and the total discounted DC (10%/y) are -254 - 391 million EUR. Thus, solar energy is partly less expensive than a business-as-usual scenario. An extensive sensitivity analysis for WACC and price escalation of conventional energy shows that for only PV and only CSP scenarios in 55 and 22 out of 72 cases the DC are negative - although no environmental costs for conventional

  17. Antenna-predominant and male-biased CSP19 of Sesamia inferens is able to bind the female sex pheromones and host plant volatiles.

    Science.gov (United States)

    Zhang, Ya-Nan; Ye, Zhan-Feng; Yang, Ke; Dong, Shuang-Lin

    2014-02-25

    Insect chemosensory proteins (CSPs) are proposed to capture and transport hydrophobic chemicals across the sensillum lymph to olfactory receptors (ORs), but this has not been clarified in moths. In this study, we built on our previously reported segment sequence work and cloned the full length CSP19 gene (SinfCSP19) from the antennae of Sesamia inferens by using rapid amplification of cDNA ends. Quantitative real time-PCR (qPCR) assays indicated that the gene was expressed in a unique profile, i.e. predominant in antennae and significantly higher in male than in female. To explore the function, recombinant SinfCSP19 was expressed in Escherichia coli cells and purified by Ni-ion affinity chromatography. Binding affinities of the recombinant SinfCSP19 with 39 plant volatiles, 3 sex pheromone components and 10 pheromone analogs were measured using fluorescent competitive binding assays. The results showed that 6 plant volatiles displayed high binding affinities to SinfCSP19 (Ki = 2.12-8.75 μM), and more interesting, the 3 sex pheromone components and analogs showed even higher binding to SinfCSP19 (Ki = 0.49-1.78 μM). Those results suggest that SinfCSP19 plays a role in reception of female sex pheromones of S. inferens and host plant volatiles. Copyright © 2013 Elsevier B.V. All rights reserved.

  18. Next Generation Solar Collectors for CSP

    Energy Technology Data Exchange (ETDEWEB)

    Molnar, Attila [3M Company, St. Paul, MN (United States); Charles, Ruth [3M Company, St. Paul, MN (United States)

    2014-07-31

    The intent of “Next Generation Solar Collectors for CSP” program was to develop key technology elements for collectors in Phase 1 (Budget Period 1), design these elements in Phase 2 (Budget Period 2) and to deploy and test the final collector in Phase 3 (Budget Period 3). 3M and DOE mutually agreed to terminate the program at the end of Budget Period 1, primarily due to timeline issues. However, significant advancements were achieved in developing a next generation reflective material and panel that has the potential to significantly improve the efficiency of CSP systems.

  19. Criteria for the assessment of reactor potential

    International Nuclear Information System (INIS)

    Carruthers, R.

    1982-01-01

    This article outlines some of the more general criteria to be used in assessing reactor potential. The interdependence of plasma and engineering parameters is considered. This demonstrated how it is the first wall power loading which is the critical parameter in assessing economic prospects. Taking some of the current conceptual designs of fusion reactors and raising the wall loading to the value needed to approach a competitive cost leads to a very challenging set of parameters. Although developed in terms of a tokamak they are figures which are applicable more generally to fusion reactors which are toroidal in form. It is not at all obvious that the tokamak could ever satisfy this criterion of economic viability, so we should not be using the parameters of existing tokamak reactor designs as the basis for assessing alternative approaches. We need to see whether there is an alternative sufficiently different as to offer a better chance of reaching these more onerous parameters. Unfortunately, so many of the alternatives differ only in magnetic geometry and their physical geometry leads to the same problems as faced by the tokamak. The traditional approach -- devising intriguing ''boxes'' for studying the confinement of plasma and then speculating on their reactor potential -- should give way to new initiatives. What we need in the fusion program is more ''reactor relevance pull'' and less ''plasma physics push'' when planning future activities

  20. Reduced host cell invasiveness and oxidative stress tolerance in double and triple csp gene family deletion mutants of Listeria monocytogenes.

    Science.gov (United States)

    Loepfe, Chantal; Raimann, Eveline; Stephan, Roger; Tasara, Taurai

    2010-07-01

    The cold shock protein (Csp) family comprises small, highly conserved proteins that bind nucleic acids to modulate various bacterial gene expressions. In addition to cold adaptation functions, this group of proteins is thought to facilitate various cellular processes to promote normal growth and stress adaptation responses. Three proteins making up the Listeria monocytogenes Csp family (CspA, CspB, and CspD) promote both cold and osmotic stress adaptation functions in this bacterium. The contribution of these three Csps in the host cell invasion processes of L. monocytogenes was investigated based on human Caco-2 and murine macrophage in vitro cell infection models. The DeltacspB, DeltacspD, DeltacspAB, DeltacspAD, DeltacspBD, and DeltacspABD strains were all significantly impaired in Caco-2 cell invasion compared with the wild-type strain, whereas in the murine macrophage infection assay only, the double (DeltacspBD) and triple (DeltacspABD) csp mutants were also significantly impaired in cell invasion compared with the wild-type strain. The DeltacspBD and DeltacspABD mutants displayed the most severely impaired invasion phenotypes. The invasion ability of these two mutant strains was also further analyzed using cold-stress-exposed organisms. In both cell infection models a significant reduction in invasiveness was observed after cold stress exposure of Listeria organisms. The negative impact of cold stress on subsequent cell invasion ability was, however, more severe in cold-sensitive csp mutants (DeltacspBD and DeltacspABD) compared with the wild type. The impaired macrophage invasion and intracellular growth of DeltacspBD and DeltacspABD also led us to examine oxidative stress resistance capacity in these two mutant strains. Both strains also displayed higher oxidative stress sensitivity relative to the wild-type strain. Our data indicate that besides cold and osmotic stress adaptation roles, Csp family proteins also promote efficient host cell invasion and

  1. Independent assessment for new nuclear reactor safety

    Directory of Open Access Journals (Sweden)

    D'Auria Francesco

    2017-01-01

    Full Text Available A rigorous framework for safety assessment is established in all countries where nuclear technology is used for the production of electricity. On the one side, industry, i.e. reactor designers, vendors and utilities perform safety analysis and demonstrate consistency between results of safety analyses and requirements. On the other side, regulatory authorities perform independent assessment of safety and confirm the acceptability of safety of individual reactor units. The process of comparing results from analyses by reactor utilities and regulators is very complex. The process is also highly dependent upon mandatory approaches pursued for the analysis and from very many details which required the knowledge of sensitive proprietary data (e.g. spacer designs. Furthermore, all data available for the design, construction and operation of reactors produced by the nuclear industry are available to regulators. Two areas for improving the process of safety assessment for individual Nuclear Power Plant Units are identified: New details introduced by industry are not always and systematically requested by regulators for the independent assessment; New analytical techniques and capabilities are not necessarily used in the analyses by regulators (and by the industry. The established concept of independent assessment constitutes the way for improving the process of safety assessment. This is possible, or is largely facilitated, by the recent availability of the so-called Best Estimate Plus Uncertainty approach.

  2. Independent assessment for new nuclear reactor safety

    International Nuclear Information System (INIS)

    D'Auria, F.; Glaeser, H.; Debrecin, N.

    2017-01-01

    A rigorous framework for safety assessment is established in all countries where nuclear technology is used for the production of electricity. On one side, industry, i.e. reactor designers, vendors and utilities perform safety analysis and demonstrate consistency between results of safety analyses and requirements. On the other side, regulatory authorities perform independent assessment of safety and confirm the acceptability of safety of individual reactor units. The process of comparing results from analyses by reactor utilities and regulators is very complex. The process is also highly dependent upon mandatory approaches pursued for the analysis and from very many details which required the knowledge of sensitive proprietary data (e.g. spacer designs). Furthermore, all data available for the design, construction and operation of reactors produced by the nuclear industry are available to regulators. Two areas for improving the process of safety assessment for individual Nuclear Power Plant Units are identified: New details introduced by industry are not always and systematically requested by regulators for the independent assessment; New analytical techniques and capabilities are not necessarily used in the analyses by regulators (and by the industry). The established concept of independent assessment constitutes the way for improving the process of safety assessment. This is possible, or is largely facilitated, by the recent availability of the so-called Best Estimate Plus Uncertainty (BEPU) approach. (authors)

  3. AnalysisThe Availability of Using Concentrated Solar Power (CSP as Electricity Source in Al-Hilla City

    Directory of Open Access Journals (Sweden)

    Wisam Shamkhi Jaber

    2017-03-01

    Full Text Available The needing of using clean energy increases every year because of the negative impact of emissions from electricity power plant and to reduce the costs of generating power by using natural energies like solar, wind, and other sources. The availability of using solar energy as source of producing electricity in Al-Hilla city by using Concentrating Solar Power (CSP was investigated in this research. The major parameters in this study were the city position, and the annually amount of solar received, also, number of charts related to solar parameters for the management of CSP were derived and showed in this research. The using of CSP as electricity power can be important solution to force the problem of high cost of electricity power fuel needed and the lack of power produced because of increasing of power consumed specially in summer season.

  4. Savannah River Site K-Reactor Probabilistic Safety Assessment

    International Nuclear Information System (INIS)

    Brandyberry, M.D.; Bailey, R.T.; Baker, W.H.; Kearnaghan, D.P.; O'Kula, K.R.; Wittman, R.S.; Woody, N.D.; Amos, C.N.; Weingardt, J.J.

    1992-12-01

    This report gives the results of a Savannah River Site (SRS) K-Reactor Probabilistic Safety Assessment (PSA). Measures of adverse consequences to health and safety resulting from representations of severe accidents in SRS reactors are presented. In addition, the report gives a summary of the methods employed to represent these accidents and to assess the resultant consequences. The report is issued to provide useful information to the U. S. Department of Energy (DOE) on the risk of operation of SRS reactors, for insights into severe accident phenomena that contribute to this risk, and in support of improved bases for other DOE programs in Heavy Water Reactor safety

  5. L-Reactor operation, Savannah River Plant: environmental assessment

    International Nuclear Information System (INIS)

    1982-08-01

    The purpose of this document is to assess the significance of the effects on the human environment of the proposed resumption of L-reactor operation at the Savannah River Plant, scheduled for October 1983. The discussion is presented under the following section headings: need for resumption of L-Reactor operations and purpose of this environmental assessment; proposed action and alternative; affected environment (including, site location and description, land use, historic and archeological resources, socioeconomic and community characteristics, geology and seismology, hydrology, meteorology and climatology, ecology, and radiation environment); environmental consequences; summary of projected L-Reactor releases and impacts; and Federal and State permits and approval. The three appendices are entitled: radiation dose calculation methods and assumptions; floodplain/wetlands assessment - L-Reactor operations; and, conversion table. A list of references is included at the end of each chapter

  6. The Moroccan solar plan. A comparative analysis of CSP and PV utilization until 2020

    Energy Technology Data Exchange (ETDEWEB)

    Richts, Christoph

    2012-02-15

    The present master thesis conducts technical and economic simulations of large-scale Photovoltaic (PV) and Concentrated Solar Power (CSP) plants for the Moroccan Solar Plan. It provides a database of performance indicators such as energy yields, capacity factors, typical efficiencies and losses of technical components, LCOE, and difference costs (DC: LCOE minus avoided costs of the conventional power system) for fixed tilted, 1-axis horizontal, 1-axis vertical and 2-axis tracking PV and CSP with no, 6, 12 and 18 full load hours of thermal storage. HelioClim irradiation data of 2005 for the sites in Ouarzazate, Ain Ben Mathar, Boujdour, Laayoune and Tarfaya is used ranging between 1,927 - 2,428 kWh/m{sup 2}/y (DNI) and 1,968 - 2,154 kWh/m{sup 2}/y (GHI). In the base scenario minimum LCOE are 9.6 - 5.4 EURct/kWh for PV (2012 - 2020) varying between 0.90 - 1.55 EURct/kWh among sites and technologies. CSP reaches 12.8 - 9.2 EURct/kWh and a bandwidth of 2.3 - 1.6 EURct/kWh. Average DC are lowest for horizontal 1-axis tracking (0.4 and -7.7 EURct/kWh for plants built in 2012 and 2020 respectively) and CSP with 6 hours of storage (1.3 and -3.5 EURct/kWh). PV is cheaper for all sites and technologies due to higher learning curves and less initial investment, but cannot contribute to coverage of the daily evening peak in Morocco. Four different MSP-scenarios with 2000 MW of solar energy require total investments of 3.7 - 7.5 billion EUR and yield 7.9% - 12.8% of the electricity demand in 2020 (given a growth 7%/y) depending on the ratio of PV and CSP utilization. The average LCOE are 8.3 - 11.7 EURct/kWh and the total discounted DC (10%/y) are -254 - 391 million EUR. Thus, solar energy is partly less expensive than a business-as-usual scenario. An extensive sensitivity analysis for WACC and price escalation of conventional energy shows that for only PV and only CSP scenarios in 55 and 22 out of 72 cases the DC are negative - although no environmental costs for conventional

  7. Cationic osteogenic peptide P15-CSP coatings promote 3-D osteogenesis in poly(epsilon-caprolactone) scaffolds of distinct pore size.

    Science.gov (United States)

    Li, Xian; Ghavidel Mehr, Nima; Guzmán-Morales, Jessica; Favis, Basil D; De Crescenzo, Gregory; Yakandawala, Nandadeva; Hoemann, Caroline D

    2017-08-01

    P15-CSP is a biomimetic cationic fusion peptide that stimulates osteogenesis and inhibits bacterial biofilm formation when coated on 2-D surfaces. This study tested the hypothesis that P15-CSP coatings enhance 3-D osteogenesis in a porous but otherwise hydrophobic poly-(ɛ-caprolactone) (PCL) scaffold. Scaffolds of 84 µm and 141 µm average pore size were coated or not with Layer-by-Layer polyelectrolytes followed by P15-CSP, seeded with adult primary human mesenchymal stem cells (MSCs), and cultured 10 days in proliferation medium, then 21 days in osteogenic medium. Atomic analyses showed that P15-CSP was successfully captured by LbL. After 2 days of culture, MSCs adhered and spread more on P15-CSP coated pores than PCL-only. At day 10, all constructs contained nonmineralized tissue. At day 31, all constructs became enveloped in a "skin" of tissue that, like 2-D cultures, underwent sporadic mineralization in areas of high cell density that extended into some 141 µm edge pores. By quantitative histomorphometry, 2.5-fold more tissue and biomineral accumulated in edge pores versus inner pores. P15-CSP specifically promoted tissue-scaffold integration, fourfold higher overall biomineralization, and more mineral deposits in the outer 84 µm and inner 141 µm pores than PCL-only (p pore surfaces with 3-D topography. Biomineralization deeper than 150 µm from the scaffold edge was optimally attained with the larger 141 µm peptide-coated pores. © 2017 Wiley Periodicals, Inc. J Biomed Mater Res Part A: 105A: 2171-2181, 2017. © 2017 Wiley Periodicals, Inc.

  8. Analysis of regulation and economic incentives of the hybrid CSP HYSOL

    DEFF Research Database (Denmark)

    Baldini, Mattia; Pérez, Cristian Hernán Cabrera

    2016-01-01

    The European HYSOL project, developed over the last three years in the solar thermal plant Manchasol (Ciudad Real, Spain), has been successfully completed, demonstrating that hybridisation of CSP with other energy sources (renewable and fossil) ensures power supply to the power grid in a stable...

  9. Deadlock Detection Based on Automatic Code Generation from Graphical CSP Models

    NARCIS (Netherlands)

    Jovanovic, D.S.; Liet, Geert K.; Broenink, Johannes F.; Karelse, F.

    2004-01-01

    The paper describes a way of using standard formal analysis tools for checking deadlock freedom in graphical models for CSP descriptions of concurrent systems. The models capture specification of a possible concurrent implementation of a system to be realized. Building the graphical models and

  10. Hanford B Reactor Building Hazard Assessment Report

    International Nuclear Information System (INIS)

    Griffin, P. W.

    1999-01-01

    The 105-B Reactor (hereinafter referred to as B Reactor) is located in the 100 Area of the Hanford Site near Richland, Washington. The B Reactor is one of nine plutonium production reactors that were constructed in the 1940s during the Cold War Era. Construction of the B Reactor began June 7, 1943, and operation began on September 26, 1944. The Environmental Restoration Contractor was requested by RL to provide an assessment/characterization of the B Reactor building to determine and document the hazards that are present and could pose a threat to the environment and/or to individuals touring the building. This report documents the potential hazards, determines the feasibility of mitigating the hazards, and makes recommendations regarding areas where public tour access should not be permitted

  11. Both the caspase CSP-1 and a caspase-independent pathway promote programmed cell death in parallel to the canonical pathway for apoptosis in Caenorhabditis elegans.

    Directory of Open Access Journals (Sweden)

    Daniel P Denning

    Full Text Available Caspases are cysteine proteases that can drive apoptosis in metazoans and have critical functions in the elimination of cells during development, the maintenance of tissue homeostasis, and responses to cellular damage. Although a growing body of research suggests that programmed cell death can occur in the absence of caspases, mammalian studies of caspase-independent apoptosis are confounded by the existence of at least seven caspase homologs that can function redundantly to promote cell death. Caspase-independent programmed cell death is also thought to occur in the invertebrate nematode Caenorhabditis elegans. The C. elegans genome contains four caspase genes (ced-3, csp-1, csp-2, and csp-3, of which only ced-3 has been demonstrated to promote apoptosis. Here, we show that CSP-1 is a pro-apoptotic caspase that promotes programmed cell death in a subset of cells fated to die during C. elegans embryogenesis. csp-1 is expressed robustly in late pachytene nuclei of the germline and is required maternally for its role in embryonic programmed cell deaths. Unlike CED-3, CSP-1 is not regulated by the APAF-1 homolog CED-4 or the BCL-2 homolog CED-9, revealing that csp-1 functions independently of the canonical genetic pathway for apoptosis. Previously we demonstrated that embryos lacking all four caspases can eliminate cells through an extrusion mechanism and that these cells are apoptotic. Extruded cells differ from cells that normally undergo programmed cell death not only by being extruded but also by not being engulfed by neighboring cells. In this study, we identify in csp-3; csp-1; csp-2 ced-3 quadruple mutants apoptotic cell corpses that fully resemble wild-type cell corpses: these caspase-deficient cell corpses are morphologically apoptotic, are not extruded, and are internalized by engulfing cells. We conclude that both caspase-dependent and caspase-independent pathways promote apoptotic programmed cell death and the phagocytosis of cell

  12. Design and prototyping of real-time systems using CSP and CML

    DEFF Research Database (Denmark)

    Rischel, Hans; Sun, Hong Yan

    1997-01-01

    A procedure for systematic design of event based systems is introduced by means of the Production Cell case study. The design is documented by CSP style processes, which allow both verification using formal techniques and also validation of a rapid prototype in the functional language CML...

  13. Enantioselective carbenoid insertion into C(sp3–H bonds

    Directory of Open Access Journals (Sweden)

    J. V. Santiago

    2016-05-01

    Full Text Available The enantioselective carbenoid insertion into C(sp3–H bonds is an important tool for the synthesis of complex molecules due to the high control of enantioselectivity in the formation of stereogenic centers. This paper presents a brief review of the early issues, related mechanistic studies and recent applications on this chemistry area.

  14. Platinum-Catalyzed, Terminal-Selective C(sp(3))-H Oxidation of Aliphatic Amines.

    Science.gov (United States)

    Lee, Melissa; Sanford, Melanie S

    2015-10-14

    This Communication describes the terminal-selective, Pt-catalyzed C(sp(3))-H oxidation of aliphatic amines without the requirement for directing groups. CuCl2 is employed as a stoichiometric oxidant, and the reactions proceed in high yield at Pt loadings as low as 1 mol%. These transformations are conducted in the presence of sulfuric acid, which reacts with the amine substrates in situ to form ammonium salts. We propose that protonation of the amine serves at least three important roles: (i) it renders the substrates soluble in the aqueous reaction medium; (ii) it limits binding of the amine nitrogen to Pt or Cu; and (iii) it electronically deactivates the C-H bonds proximal to the nitrogen center. We demonstrate that this strategy is effective for the terminal-selective C(sp(3))-H oxidation of a variety of primary, secondary, and tertiary amines.

  15. Guidelines for Self-assessment of Research Reactor Safety

    International Nuclear Information System (INIS)

    2018-01-01

    Self-assessment is an organization’s internal process to review its current status, processes and performance against predefined criteria and thereby to provide key elements for the organization’s continual development and improvement. Self-assessment helps the organization to think through what it is expected to do, how it is performing in relation to these expectations, and what it needs to do to improve performance, fulfil the expectations and achieve better compliance with the predefined criteria. This publication provides guidelines for a research reactor operating organization to perform a self-assessment of the safety management and the safety of the facility and to identify gaps between the current situation and the IAEA safety requirements for research reactors. These guidelines also provide a methodology for Member States, regulatory bodies and operating organizations to perform a self-assessment of their application of the provisions of the Code of Conduct on the Safety of Research Reactors. This publication also addresses planning, implementation and follow-up of actions to enhance safety and strengthen application of the Code. The guidelines are applicable to all types of research reactor and critical and subcritical assemblies, at all stages in their lifetimes, and to States, regulatory bodies and operating organizations throughout all phases of research reactor programmes. Research reactor operating organizations can use these guidelines at any time to support self-assessments conducted in accordance with the organization’s integrated management system. These guidelines also serve as a tool for an organization to prepare to receive an IAEA Integrated Safety Assessment of Research Reactors (INSARR) mission. An important result of this is the opportunity for an operating organization to identify focus areas and make safety improvements in advance of an INSARR mission, thereby increasing the effectiveness of the mission and efficiency of the

  16. Assessment of fusion reactor development. Proceedings

    International Nuclear Information System (INIS)

    Inoue, N.; Tazima, T.

    1994-04-01

    Symposium on assessment of fusion reactor development was held to make clear critical issues, which should be resolved for the commercial fusion reactor as a major energy source in the next century. Discussing items were as follows. (1) The motive force of fusion power development from viewpoints of future energy demand, energy resources and earth environment for 'Sustainable Development'. (2) Comparison of characteristics with other alternative energy sources, i.e. fission power and solar cell power. (3) Future planning of fusion research and advanced fuel fusion (D 3 He). (4) Critical issues of fusion reactor development such as Li extraction from the sea water, structural material and safety. (author)

  17. Space-reactor electric systems: subsystem technology assessment

    International Nuclear Information System (INIS)

    Anderson, R.V.; Bost, D.; Determan, W.R.

    1983-01-01

    This report documents the subsystem technology assessment. For the purpose of this report, five subsystems were defined for a space reactor electric system, and the report is organized around these subsystems: reactor; shielding; primary heat transport; power conversion and processing; and heat rejection. The purpose of the assessment was to determine the current technology status and the technology potentials for different types of the five subsystems. The cost and schedule needed to develop these potentials were estimated, and sets of development-compatible subsystems were identified

  18. Mise en Scene: Conversion of Scenarios to CSP Traces for the Requirements-to-Design-to-Code Project

    Science.gov (United States)

    Carter. John D.; Gardner, William B.; Rash, James L.; Hinchey, Michael G.

    2007-01-01

    The "Requirements-to-Design-to-Code" (R2D2C) project at NASA's Goddard Space Flight Center is based on deriving a formal specification expressed in Communicating Sequential Processes (CSP) notation from system requirements supplied in the form of CSP traces. The traces, in turn, are to be extracted from scenarios, a user-friendly medium often used to describe the required behavior of computer systems under development. This work, called Mise en Scene, defines a new scenario medium (Scenario Notation Language, SNL) suitable for control-dominated systems, coupled with a two-stage process for automatic translation of scenarios to a new trace medium (Trace Notation Language, TNL) that encompasses CSP traces. Mise en Scene is offered as an initial solution to the problem of the scenarios-to-traces "D2" phase of R2D2C. A survey of the "scenario" concept and some case studies are also provided.

  19. Reactor Safety Assessment System--A situation assessment aid for USNRC emergency response

    International Nuclear Information System (INIS)

    Bray, M.A.; Sebo, D.E.; Dixon, B.W.

    1985-01-01

    The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSAS is intended for use at the NRC's Operations Center in the event of a serious incident at a licensed nuclear power plant. The system uses plant parameter data and status information from the power plant. It has a rule base that uses the parametric values, the known operator actions, and the time sequence information in the data to generate situation assessment conclusions for use by the NRC Reactor Safety Team. RSAS rules currently cover one specific reactor type and use setpoints specific to one power plant

  20. Reactor Safety Assessment System: a situation assessment aid for USNRC emergency response

    International Nuclear Information System (INIS)

    Bray, M.A.; Sebo, D.E.; Dixon, B.W.

    1985-04-01

    The Reactor Safety Assessment System is an expert system under development for the United States Nuclear Regulatory Commission (NRC). RSAS is intended for use at the NRC's Operations Center in the event of a serious incident at a licensed nuclear power plant. The system uses plant parameter data and status information from the power plant. It has a rule base which uses the parametric values, the known operator actions and the time sequence information in the data to generate situation assessment conclusions for use by the NRC Reactor Safety Team. RSAS rules currently cover one specific reactor type and use setpoints specific to one power plant. 5 figs

  1. Space reactor system and subsystem investigations: assessment of technology issues for the reactor and shield subsystem. SP-100 Program

    International Nuclear Information System (INIS)

    Atkins, D.F.; Lillie, A.F.

    1983-01-01

    As part of Rockwell's effort on the SP-100 Program, preliminary assessment has been completed of current nuclear technology as it relates to candidate reactor/shield subsystems for the SP-100 Program. The scope of the assessment was confined to the nuclear package (to the reactor and shield subsystems). The nine generic reactor subsystems presented in Rockwell's Subsystem Technology Assessment Report, ESG-DOE-13398, were addressed for the assessment

  2. Preliminary Options Assessment of Versatile Irradiation Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sen, Ramazan Sonat [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-01-01

    The objective of this report is to summarize the work undertaken at INL from April 2016 to January 2017 and aimed at analyzing some options for designing and building a versatile test reactor; the scope of work was agreed upon with DOE-NE. Section 2 presents some results related to KNK II and PRISM Mod A. Section 3 presents some alternatives to the VCTR presented in [ ] as well as a neutronic parametric study to assess the minimum power requirement needed for a 235U metal fueled fast test reactor capable to generate a fast (>100 keV) flux of 4.0 x 1015 n /cm2-s at the test location. Section 4 presents some results regarding a fundamental characteristic of test reactors, namely displacement per atom (dpa) in test samples. Section 5 presents the INL assessment of the ANL fast test reactor design FASTER. Section 6 presents a summary.

  3. GEOTHERMAL / SOLAR HYBRID DESIGNS: USE OF GEOTHERMAL ENERGY FOR CSP FEEDWATER HEATING

    Energy Technology Data Exchange (ETDEWEB)

    Craig Turchi; Guangdong Zhu; Michael Wagner; Tom Williams; Dan Wendt

    2014-10-01

    This paper examines a hybrid geothermal / solar thermal plant design that uses geothermal energy to provide feedwater heating in a conventional steam-Rankine power cycle deployed by a concentrating solar power (CSP) plant. The geothermal energy represents slightly over 10% of the total thermal input to the hybrid plant. The geothermal energy allows power output from the hybrid plant to increase by about 8% relative to a stand-alone CSP plant with the same solar-thermal input. Geothermal energy is converted to electricity at an efficiency of 1.7 to 2.5 times greater than would occur in a stand-alone, binary-cycle geothermal plant using the same geothermal resource. While the design exhibits a clear advantage during hybrid plant operation, the annual advantage of the hybrid versus two stand-alone power plants depends on the total annual operating hours of the hybrid plant. The annual results in this draft paper are preliminary, and further results are expected prior to submission of a final paper.

  4. Risk-assessment techniques and the reactor licensing process

    International Nuclear Information System (INIS)

    Levine, S.

    1979-01-01

    A brief description of the Reactor Safety Study (WASH-1400), concentrating on the engineering aspects of the contribution to reactor accident risks is followed by some comments on how we have applied the insights and techniques developed in this study to prepare a program to improve the safety of nuclear power plants. Some new work we are just beginning on the application of risk-assessment techniques to stablize the reactor licensing process is also discussed

  5. Significance assessment of small-medium sized reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kanno, Minoru [Japan Atomic Power Co., Research and Development Dept., Tokyo (Japan)

    2002-12-01

    Preliminary assessment for deployment of small-medium sized reactor (S and M reactor) as a future option has been conducted at the JAPCO (Japan Atomic Power Company) under the cooperation with the CRIERI (Central Research Institute of Electric Power Industry). Significance of the S and M reactor introduction is listed as follows; lower investment cost, possible siting near demand side, enlarged freedom of siting, shorter transmission line, good compatibility with slow increase of demand and plain explanation of safety using simpler system such as integral type vessel without piping, natural convection core cooling and passive safety system. The deployment of simpler plant system, modular shop fabrication, ship-shell structured building and longer operation period can assure economics comparable with that of a large sized reactor, coping with scale-demerit. Also the S and M reactor is preferable in size for the nuclear heat utilization such as hydrogen production. (T. Tanaka)

  6. Thermal tests of a multi-tubular reactor for hydrogen production by using mixed ferrites thermochemical cycle

    Science.gov (United States)

    Gonzalez-Pardo, Aurelio; Denk, Thorsten; Vidal, Alfonso

    2017-06-01

    The SolH2 project is an INNPACTO initiative of the Spanish Ministry of Economy and Competitiveness, with the main goal to demonstrate the technological feasibility of solar thermochemical water splitting cycles as one of the most promising options to produce H2 from renewable sources in an emission-free way. A multi-tubular solar reactor was designed and build to evaluate a ferrite thermochemical cycle. At the end of this project, the ownership of this plant was transferred to CIEMAT. This paper reviews some additional tests with this pilot plant performed in the Plataforma Solar de Almería with the main goal to assess the thermal behavior of the reactor, evaluating the evolution of the temperatures inside the cavity and the relation between supplied power and reached temperatures. Previous experience with alumina tubes showed that they are very sensitive to temperature and flux gradients, what leads to elaborate an aiming strategy for the heliostat field to achieve a uniform distribution of the radiation inside the cavity. Additionally, the passing of clouds is a phenomenon that importantly affects all the CSP facilities by reducing their efficiency. The behavior of the reactor under these conditions has been studied.

  7. Application of fuzzy synthetic assessment to assess human factors design level on reactor control panel

    International Nuclear Information System (INIS)

    Peng Xuecheng

    1999-01-01

    Reactor control panel design level on human factors must be considered by designer. The author evaluated the human factor design level of arrangement and combinations including the switch buttons, meter dials and indication lamps on Minjiang Reactor and High-Flux Engineer Test Reactor (HFETR) critical device by application of fuzzy synthetic assessment method in mathematics. From the assessment results, the advantages and shortcomings are fount, and some modification suggestions have also been proposed

  8. A study of reactor vessel integrity assessment

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Hoon [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Kim, Jong Kyung; Shin, Chang Ho; Seo, Bo Kyun [Hanyang Univ., Seoul (Korea, Republic of)

    1999-02-15

    The fast neutron fluence at the Reactor Pressure Vessel(RPV) of KNGR designed for 60 years lifetime was calculated by full-scope Monte Carlo simulation for reactor vessel integrity assessment. KNGR core geometry was modeled on a three-dimensional representation of the one-sixteenth of the reactor in-vessel component. Each fuel assemblies were modeled explicitly, and each fuel pins were axially divided into 5 segments. The maximum flux of 4.3 x 10{sup 10} neutrons/cm{sup 2}. sec at the RPV was obtained by tallying neutrons crossing the beltline of inner surface of the RPV.

  9. Advanced Test Reactor probabilistic risk assessment

    International Nuclear Information System (INIS)

    Atkinson, S.A.; Eide, S.A.; Khericha, S.T.; Thatcher, T.A.

    1993-01-01

    This report discusses Level 1 probabilistic risk assessment (PRA) incorporating a full-scope external events analysis which has been completed for the Advanced Test Reactor (ATR) located at the Idaho National Engineering Laboratory

  10. Containment concepts assessment for the SEAFP reactor

    International Nuclear Information System (INIS)

    Di Pace, L.; Natalizio, A.

    2000-01-01

    A simple methodology has been developed for making relative comparisons of potential containment designs for future fusion reactors. The assessment methodology requires only conceptual design information. The application of this methodology, at the early stages of a fusion reactor design, provides designers useful information regarding the suitability of various containment designs and design features. Because the radiation hazard from the operation of future fusion power reactors is expected to be low, the containment design, in addition to public safety, needs to take into account worker safety considerations, as well as factors important to the reliable and economical operation of the power plant. Several containment concepts have been assessed with a methodology that takes into account public safety, worker safety, operability and maintainability as well as cost. This paper describes this methodology and presents the results of the assessment. The paper concludes that, to obtain a containment design that is optimised with respect to safety, operational and cost factors, designers should focus on a containment that is conceptually simple-that is, one utilising a single, large containment building without relying on special features such as expansion volumes, pressure suppression pools or spray systems

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

    International Nuclear Information System (INIS)

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

    1980-06-01

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

  12. Co-generation and innovative heat storage systems in small-medium CSP plants for distributed energy production

    Science.gov (United States)

    Giaconia, Alberto; Montagnino, Fabio; Paredes, Filippo; Donato, Filippo; Caputo, Giampaolo; Mazzei, Domenico

    2017-06-01

    CSP technologies can be applied for distributed energy production, on small-medium plants (on the 1 MW scale), to satisfy the needs of local communities, buildings and districts. In this perspective, reliable, low-cost, and flexible small/medium multi-generative CSP plants should be developed. Four pilot plants have been built in four Mediterranean countries (Cyprus, Egypt, Jordan, and Italy) to demonstrate the approach. In this paper, the plant built in Italy is presented, with specific innovations applied in the linear Fresnel collector design and the Thermal Energy Storage (TES) system, based on a single the use of molten salts but specifically tailored for small scale plants.

  13. Platinum-Catalyzed Terminal-Selective C(sp3)–H Oxidation of Aliphatic Amines

    Science.gov (United States)

    Lee, Melissa; Sanford, Melanie S.

    2016-01-01

    This paper describes the terminal-selective Pt-catalyzed C(sp3)–H oxidation of aliphatic amines without the requirement for directing groups. CuCl2 is employed as a stoichiometric oxidant, and the reactions proceed in high yield at Pt loadings as low as 1 mol %. These transformations are conducted in the presence of sulfuric acid, which reacts with the amine substrates in situ to form ammonium salts. We propose that protonation of the amine serves at least three important roles: (i) it renders the substrates soluble in the aqueous reaction medium; (ii) it limits binding of the amine nitrogen to Pt or Cu; and (ii) it electronically deactivates the C–H bonds proximal to the nitrogen center. We demonstrate that this strategy is effective for the terminal-selective C(sp3)–H oxidation of a variety of primary, secondary and tertiary amines. PMID:26439251

  14. Assessment of the enhanced DHRS configuration for MYRRHA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bubelis, E.; Jaeger, W. [KIT, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Bandini, G. [ENEA, via Martiri di Monte Sole, 4, 40129 Bologna (Italy); Alemberti, A.; Palmero, M. [ANSALDO, Corso Perrone 25, 16152 Genova (Italy)

    2016-10-15

    Highlights: • Innovative decay heat removal system (DHRS). • Heavy liquid metal cooled reactor. • Avoiding of lead bismuth eutectic (LBE) freezing. • Numerical assessment and proof of operational principles of innovative DHRS. - Abstract: This paper deals with the assessment of an innovative decay heat removal system for the MYRRHA reactor, based on the analysis of the selected transients with two different system codes. The application to liquid metal cooled reactors has the disadvantage of adding overcooling transients to the transient spectrum. Under these circumstances, freezing of the coolant can occur if no corrective or operator actions are taken in the medium and long term. Therefore, ANSALDO Nucleare invented an enhanced decay heat removal system which avoids the risk of freezing. The numerical assessment and proof of operational principles are performed by KIT and ENEA. The simulation results show that the freezing can be avoided. Moreover, both institutions calculate similar behavior during overcooling transients. This study will help to implement the novel decay heat removal system and the overall safety philosophy of innovative reactor concepts.

  15. Reactor technology assessment and selection utilizing systems engineering approach

    Science.gov (United States)

    Zolkaffly, Muhammed Zulfakar; Han, Ki-In

    2014-02-01

    The first Nuclear power plant (NPP) deployment in a country is a complex process that needs to consider technical, economic and financial aspects along with other aspects like public acceptance. Increased interest in the deployment of new NPPs, both among newcomer countries and those with expanding programs, necessitates the selection of reactor technology among commercially available technologies. This paper reviews the Systems Decision Process (SDP) of Systems Engineering and applies it in selecting the most appropriate reactor technology for the deployment in Malaysia. The integrated qualitative and quantitative analyses employed in the SDP are explored to perform reactor technology assessment and to select the most feasible technology whose design has also to comply with the IAEA standard requirements and other relevant requirements that have been established in this study. A quick Malaysian case study result suggests that the country reside with PWR (pressurized water reactor) technologies with more detailed study to be performed in the future for the selection of the most appropriate reactor technology for Malaysia. The demonstrated technology assessment also proposes an alternative method to systematically and quantitatively select the most appropriate reactor technology.

  16. The Potential of Coconut Shell Powder (CSP) and Coconut Shell Activated Carbon (CSAC) Composites as Electromagnetic Interference (EMI) Absorbing Material

    International Nuclear Information System (INIS)

    Siti Nurbazilah Abdul Jabal; Seok, Y.B.; Hoon, W.F.

    2016-01-01

    Agriculture waste is potentially useful as an alternative material to absorb and attenuate electromagnetic interference (EMI). This research highlights the use of coconut shell powder (CSP) and coconut shell activated carbon (CSAC) as raw materials with epoxy resin and amine hardener composite to absorb microwave signals over frequency of 1 - 8 GHz. In order to investigate the suitability of these raw materials as EMI absorbing material, carbon composition of the raw materials is determined through CHNS Elemental Analysis. The surface morphology of the raw materials in term of porosity is investigated by using TM3000 Scanning Electron Microscope (SEM). The complex permittivity of the composites is determined by using high temperature dielectric probe in conjunction with Network Analyzer. From the result, the Carbon% of CSP and CSAC is 46.70 % and 84.28 % respectively. In term of surface morphology, the surface porosity of CSP and CSAC is in the range of 2 μm and 1 μm respectively. For the dielectric properties, the dielectric constant and the dielectric loss factor for CSP and CSAC is 4.5767 and 64.8307 and 1.2144 and 13.8296 respectively. The materials more potentially useful as substitute materials for electromagnetic interference (EMI) absorbing are discussed. (author)

  17. Iridium complexes containing mesoionic C donors: selective C(sp3)-H versus C(sp2)-H bond activation, reactivity towards acids and bases, and catalytic oxidation of silanes and water.

    Science.gov (United States)

    Petronilho, Ana; Woods, James A; Mueller-Bunz, Helge; Bernhard, Stefan; Albrecht, Martin

    2014-11-24

    Metalation of a C2-methylated pyridylimidazolium salt with [IrCp*Cl2]2 affords either an ylidic complex, resulting from C(sp(3))-H bond activation of the C2-bound CH3 group if the metalation is performed in the presence of a base, such as AgO2 or Na2CO3, or a mesoionic complex via cyclometalation and thermally induced heterocyclic C(sp(2))-H bond activation, if the reaction is performed in the absence of a base. Similar cyclometalation and complex formation via C(sp(2))-H bond activation is observed when the heterocyclic ligand precursor consists of the analogous pyridyltriazolium salt, that is, when the metal bonding at the C2 position is blocked by a nitrogen rather than a methyl substituent. Despite the strongly mesoionic character of both the imidazolylidene and the triazolylidene, the former reacts rapidly with D(+) and undergoes isotope exchange at the heterocyclic C5 position, whereas the triazolylidene ligand is stable and only undergoes H/D exchange under basic conditions, where the imidazolylidene is essentially unreactive. The high stability of the Ir-C bond in aqueous solution over a broad pH range was exploited in catalytic water oxidation and silane oxidation. The catalytic hydrosilylation of ketones proceeds with turnover frequencies as high as 6,000 h(-1) with both the imidazolylidene and the triazolylidene system, whereas water oxidation is enhanced by the stronger donor properties of the imidazol-4-ylidene ligands and is more than three times faster than with the triazolylidene analogue. © 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  18. System assessment of helical reactors in comparison with tokamaks

    International Nuclear Information System (INIS)

    Yamazaki, K.; Imagawa, S.; Muroga, T.; Sagara, A.; Okamura, S.

    2002-10-01

    A comparative assessment of tokamak and helical reactors has been performed using equivalent physics/engineering model and common costing model. Higher-temperature plasma operation is required in tokamak reactors to increase bootstrap current fraction and to reduce current-drive (CD) power. In helical systems, lower-temperature operation is feasible and desirable to reduce helical ripple transport. The capital cost of helical reactor is rather high, however, the cost of electricity (COE) is almost same as that of tokamak reactor because of smaller re-circulation power (no CD power) and less-frequent blanket replacement (lower neutron wall loading). The standard LHD-type helical reactor with 5% beta value is economically equivalent to the standard tokamak with 3% beta. The COE of lower-aspect ratio helical reactor is on the same level of high-β N tokamak reactors. (author)

  19. Thermal energy storage for CSP (Concentrating Solar Power)

    Science.gov (United States)

    Py, Xavier; Sadiki, Najim; Olives, Régis; Goetz, Vincent; Falcoz, Quentin

    2017-07-01

    The major advantage of concentrating solar power before photovoltaic is the possibility to store thermal energy at large scale allowing dispatchability. Then, only CSP solar power plants including thermal storage can be operated 24 h/day using exclusively the solar resource. Nevertheless, due to a too low availability in mined nitrate salts, the actual mature technology of the two tanks molten salts cannot be applied to achieve the expected international share in the power production for 2050. Then alternative storage materials are under studies such as natural rocks and recycled ceramics made from industrial wastes. The present paper is a review of those alternative approaches.

  20. Thermal energy storage for CSP (Concentrating Solar Power

    Directory of Open Access Journals (Sweden)

    Py Xavier

    2017-01-01

    Full Text Available The major advantage of concentrating solar power before photovoltaic is the possibility to store thermal energy at large scale allowing dispatchability. Then, only CSP solar power plants including thermal storage can be operated 24 h/day using exclusively the solar resource. Nevertheless, due to a too low availability in mined nitrate salts, the actual mature technology of the two tanks molten salts cannot be applied to achieve the expected international share in the power production for 2050. Then alternative storage materials are under studies such as natural rocks and recycled ceramics made from industrial wastes. The present paper is a review of those alternative approaches.

  1. Rhodium(III)-Catalyzed Activation of C(sp3)-H Bonds and Subsequent Intermolecular Amidation at Room Temperature.

    Science.gov (United States)

    Huang, Xiaolei; Wang, Yan; Lan, Jingbo; You, Jingsong

    2015-08-03

    Disclosed herein is a Rh(III)-catalyzed chelation-assisted activation of unreactive C(sp3)-H bonds, thus enabling an intermolecular amidation to provide a practical and step-economic route to 2-(pyridin-2-yl)ethanamine derivatives. Substrates with other N-donor groups are also compatible with the amidation. This protocol proceeds at room temperature, has a relatively broad functional-group tolerance and high selectivity, and demonstrates the potential of rhodium(III) in the promotive functionalization of unreactive C(sp3)-H bonds. A rhodacycle having a SbF6(-) counterion was identified as a plausible intermediate. © 2015 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  2. INPRO economic assessment of the IRIS nuclear reactor for deployment in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Goncalves Filho, Orlando Joao Agostinho, E-mail: orlando@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN - RJ), Rua Helio de Almeida, 75, Cidade Universitaria, Ilha do Fundao, 21941-906 Rio de Janeiro, RJ (Brazil)

    2011-06-15

    Highlights: > First INPRO evaluation of IRIS economic competitiveness for deployment in Brazil. > Plant arrangement of three independent IRIS single units constructed in series. > Angra 3 reactor used as reference design for judgment of IRIS economic potential. > IRIS economically competes with 2nd generation nuclear power plants in Brazil - Abstract: This paper presents the results of the economic assessment of the International Reactor Innovative and Secure (IRIS) for deployment in Brazil using the assessment methodology developed under the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO), co-ordinated by the International Atomic Energy Agency (IAEA). INPRO initiated in 2001 and has the main objective of helping to ensure that nuclear energy will be available to contribute in a sustainable manner to the energy needs of the 21st century. Among its missions is the development of a methodology to assess innovative nuclear energy systems (INSs) on a global, regional and national basis. In 2005, Brazil submitted a proposal for the assessment of two small-size reactors as components of an INS, completed with a conventional open nuclear fuel cycle based on enriched uranium. One of the reactors assessed was IRIS, a small-size, modular, integral-type PWR reactor. IRIS was evaluated with regard to the areas of reactor safety and economics only. This paper outlines the rationale for the study and summarizes the results of the economic assessment. The study concluded that the reference design of IRIS complies with most of INPRO economics criteria and has potential to comply with the remaining ones.

  3. Further structural insights into the binding of complement factor H by complement regulator-acquiring surface protein 1 (CspA) of Borrelia burgdorferi

    International Nuclear Information System (INIS)

    Caesar, Joseph J. E.; Wallich, Reinhard; Kraiczy, Peter; Zipfel, Peter F.; Lea, Susan M.

    2013-01-01

    B. burgdorferi binds complement factor H using a dimeric surface protein, CspA (BbCRASP-1). Presented here is a new structure of CspA that suggests that there is a degree of flexibility between subunits which may have implications for complement regulator binding. Borrelia burgdorferi has evolved many mechanisms of evading the different immune systems across its range of reservoir hosts, including the capture and presentation of host complement regulators factor H and factor H-like protein-1 (FHL-1). Acquisition is mediated by a family of complement regulator-acquiring surface proteins (CRASPs), of which the atomic structure of CspA (BbCRASP-1) is known and shows the formation of a homodimeric species which is required for binding. Mutagenesis studies have mapped a putative factor H binding site to a cleft between the two subunits. Presented here is a new atomic structure of CspA which shows a degree of flexibility between the subunits which may be critical for factor H scavenging by increasing access to the binding interface and allows the possibility that the assembly can clamp around the bound complement regulators

  4. Assessment of Smart Reactor Utilization for Barelang

    International Nuclear Information System (INIS)

    Sahala-M-Lumbanraja; Yuliastuti

    2007-01-01

    This paper assesses the feasibility of SMART reactor utilization in BARELANG region. BARELANG region is an industrial area located in Riau Islands Province. The need of electricity and fresh water, whether for industry growth or people, are the main problem of this region. Until now, the National Electricity Company (PLN) has not able to supply the electricity needed by industrial sector. The use of oil as a main electricity generation resource of the entire power plant has caused a tremendous generation cost. On dry seasons, the fresh water supplied by PDAM is reducing drastically. This situation occurs because water source of PDAM extremely depends on the water storage during rainy seasons. SMART reactor is a modular light reactor developed by KAERI for dual purposes, producing electricity and fresh water at the same time. The total thermal power generated by this type of reactor is about 330 M Wth with 33 % efficiency, as 90 M We connected to the electricity grid and rest is used in producing potable water with capacity 40,000 m 3 /day. Compare to the conventional reactor, SMART reactor is based on simple operation and maintenance principles, enhanced safety, easy to inspect, a relatively short construction time, small investment cost, competitive generation cost, and a flexible design to fit with the existing infrastructure. The main characteristic of SMART reactor is an integral design concept where the entire main cooling system components are located in the pressurize vessel. (author)

  5. Guidelines for the review research reactor safety. Reference document for IAEA Integrated Safety Assessment of Research Reactors (INSARR)

    International Nuclear Information System (INIS)

    1997-01-01

    In 1992, the IAEA published new safety standards for research reactors as part of the set of publications considered by its Research Reactor Safety Programme (RRSP). This set also includes publications giving guidance for all safety aspects related to the lifetime of a research reactor. In addition, the IAEA has also revised the Safety Standards for radiation protection. Consequently, it was considered advisable to revise the Integrated Safety Assessment of Research Reactors (INSARR) procedures to incorporate the new requirements and guidance as well as to extend the scope of the safety reviews to currently operating research reactors. The present report is the result of this revision. The purpose of this report is to give guidance on the preparation, execution, reporting and follow-up of safety review mission to research reactors as conducted by the IAEA under its INSARR missions safety service. However, it will also be of assistance to operators and regulators in conducting: (a) ad hoc safety assessments of research reactors to address individual issues such as ageing or safety culture; and (b) other types of safety reviews such as internal and peer reviews and regulatory inspections

  6. Advanced Test Reactor outage risk assessment

    International Nuclear Information System (INIS)

    Thatcher, T.A.; Atkinson, S.A.

    1997-01-01

    Beginning in 1997, risk assessment was performed for each Advanced Test Reactor (ATR) outage aiding the coordination of plant configuration and work activities (maintenance, construction projects, etc.) to minimize the risk of reactor fuel damage and to improve defense-in-depth. The risk assessment activities move beyond simply meeting Technical Safety Requirements to increase the awareness of risk sensitive configurations, to focus increased attention on the higher risk activities, and to seek cost-effective design or operational changes that reduce risk. A detailed probabilistic risk assessment (PRA) had been performed to assess the risk of fuel damage during shutdown operations including heavy load handling. This resulted in several design changes to improve safety; however, evaluation of individual outages had not been performed previously and many risk insights were not being utilized in outage planning. The shutdown PRA provided the necessary framework for assessing relative and absolute risk levels and assessing defense-in-depth. Guidelines were written identifying combinations of equipment outages to avoid. Screening criteria were developed for the selection of work activities to receive review. Tabulation of inherent and work-related initiating events and their relative risk level versus plant mode has aided identification of the risk level the scheduled work involves. Preoutage reviews are conducted and post-outage risk assessment is documented to summarize the positive and negative aspects of the outage with regard to risk. The risk for the outage is compared to the risk level that would result from optimal scheduling of the work to be performed and to baseline or average past performance

  7. Overview of fourth generation reactors. Assessment in terms of safety and radiation protection

    International Nuclear Information System (INIS)

    Couturier, J.; Baudrand, O.; Blanc, D.; Bourgois, T.; Hache, G.; Ivanov, E.; Bonneville, H.; Meignen, R.; Nicaise, G.; Bruna, G.; Clement, B.; Kissane, M.; Monhardt, B.

    2012-01-01

    Based on a systematic analysis of the different concepts of fourth generation nuclear reactors, this report gives an overview of specific aspects regarding safety and radiation protection for six concepts: sodium fast reactors (SFR), gas fast reactors (GFR), lead fast reactors (LFR), molten salt reactors (MSR), very high or high temperature reactors (V/HTR) and supercritical water reactors (SCWR). This assessment is based on different studies and researches performed by the IRSN at an international level. For each reactor concept, the report proposes a presentation of the current status of development and its perspectives, describes the safety aspects which are specific to this concept, identifies and discusses elements for safety analysis, and assesses the concept with respect to the Fukushima accident and IAEA recommendations and predefined themes

  8. 75 FR 13740 - Office of Innovation and Improvement; Overview Information; Charter Schools Program (CSP) Grants...

    Science.gov (United States)

    2010-03-23

    ... DEPARTMENT OF EDUCATION Office of Innovation and Improvement; Overview Information; Charter Schools Program (CSP) Grants for National Leadership Activities; Notice Inviting Applications for New... of public schools have been identified for improvement, corrective action, or restructuring under...

  9. Modeling issues associated with production reactor safety assessment

    International Nuclear Information System (INIS)

    Stack, D.W.; Thomas, W.R.

    1990-01-01

    This paper describes several Probabilistic Safety Assessment (PSA) modeling issues that are related to the unique design and operation of the production reactors. The identification of initiating events and determination of a set of success criteria for the production reactors is of concern because of their unique design. The modeling of accident recovery must take into account the unique operation of these reactors. Finally, a more thorough search and evaluation of common-cause events is required to account for combinations of unique design features and operation that might otherwise not be included in the PSA. It is expected that most of these modeling issues also would be encountered when modeling some of the other more unique reactor and nonreactor facilities that are part of the DOE nuclear materials production complex. 9 refs., 2 figs

  10. Assessment of tritium breeding requirements for fusion power reactors

    International Nuclear Information System (INIS)

    Jung, J.

    1983-12-01

    This report presents an assessment of tritium-breeding requirements for fusion power reactors. The analysis is based on an evaluation of time-dependent tritium inventories in the reactor system. The method presented can be applied to any fusion systems in operation on a steady-state mode as well as on a pulsed mode. As an example, the UWMAK-I design was analyzed and it has been found that the startup inventory requirement calculated by the present method significantly differs from those previously calculated. The effect of reactor-parameter changes on the required tritium breeding ratio is also analyzed for a variety of reactor operation scenarios

  11. Assessment of beam tube performance for the maple research reactor

    International Nuclear Information System (INIS)

    Lee, A.G.

    1986-06-01

    The MAPLE research reactor is a versatile new research facility that can be adapted to meet the requirements of a variety of reactor applications. A particular group of reactor applications involves the use of beams of radiation extracted from the reactor core via tubes that penetrate through the biological shield and terminate in the reflector surrounding the fuelled core. An assessment is given of the neutron and gamma radiation fields entering beam tubes that are located radially or tangentially with respect to the core

  12. Catalyzed deuterium fueled reversed-field pinch reactor assessment

    International Nuclear Information System (INIS)

    Dobrott, D.

    1985-01-01

    This study is part of a Department of Energy supported alternate fusion fuels program at Science Applications International Corporation. The purpose of this portion of the study is to perform an assessment of a conceptual compact reversed-field pinch reactor (CRFPR) that is fueled by the catalyzed-deuterium (Cat-d) fuel cycle with respect to physics, technology, safety, and cost. The Cat-d CRFPR is compared to a d-t fueled fusion reactor with respect to several issues in this study. The comparison includes cost, reactor performance, and technology requirements for a Cat-d fueled CRFPR and a comparable cost-optimized d-t fueled conceptual design developed by LANL

  13. Safety Management and Safety Culture Self Assessment of Kartini Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Syarip, S., E-mail: syarip@batan.go.id [Centre for Accelerator and Material Process Technology, National Nuclear Energy Agency (BATAN), Yogyakarta (Indonesia)

    2014-10-15

    The self-assessment of safety culture and safety management status of Kartini research reactor is a step to foster safety culture and management by identifying good practices and areas for improvement, and also to improve reactor safety in a whole. The method used in this assessment is based on questionnaires provided by the Forum for Nuclear Cooperation in Asia (FNCA), then reviewed by experts. Based on the assessment and evaluation results, it can be concluded that there were several good practices in maintaining the safety status of Kartini reactor such as: reactor operators and radiation protection workers were aware and knowledgeable of the safety standards and policies that apply to their operation, readily accept constructive criticism from their management and from the inspectors of regulatory body that address safety performance. As a proof, for the last four years the number of inspection/audit findings from Regulatory Body (BAPETEN) tended to decrease while the reactor utilization and its operating hour increased. On the other hands there were also some comments and recommendations for improvement of reactor safety culture, such as that there should be more frequent open dialogues between employees and managers, to grow and attain a mutual support to achieve safety goals. (author)

  14. The seismic assessment of fast reactor cores in the UK

    International Nuclear Information System (INIS)

    Duthie, J.C.; Dostal, M.

    1988-01-01

    The design of the UK Commercial Demonstration Fast Reactor (CDFR) has evolved over a number of years. The design has to meet two seismic requirements: (i) the reactor must cause no hazard to the public during or after the Safe Shutdown Earthquake (SSE); (ii) there must be no sudden reduction in safety for an earthquake exceeding the SSE. The core is a complicated component in the whole reactor. It is usually represented in a very simplified manner in the seismic assessment of the whole reactor station. From this calculation, a time history or response spectrum can be generated for the diagrid, which supports the core, and for the above core structure, which supports the main absorber rods. These data may then be used to perform a detailed assessment of the reactor core. A new simplified model of the core response may then be made and used in a further calculation of the whole reactor. The calculation of the core response only, is considered in the remainder of this paper. One important feature of the fast reactor core, compared with other reactors, is that the components are relatively thin and flexible to promote neutron economy and heat transfer. A further important feature is that there are very small gaps between the wrapper tubes. This leads to very strong fluid-coupling effects. These effects are likely to be beneficial, but adequate techniques to calculate them are only just being developed. 9 refs, figs

  15. Reactor vessel assessment and the development of a reactor vessel life extension program for Calvert Cliffs Units One and Two

    International Nuclear Information System (INIS)

    Montgomery, B.; Hijeck, P.J.

    1988-01-01

    A study has been undertaken to provide a general assessment of the life extension capabilities for the Calvert Cliffs Units One and Two reactor pressure vessels. The purpose of the study is to assess the general life extension capabilities for the Calvert Cliffs reactor pressure vessels based upon an extension and variation of the Surry pilot plant life extension study. This assessment provided a detailed reactor vessel surveillance program for plant life extension along with a hierarchy of specific tasks necessary for attaining maximum useful life. The assessment identified a number of critical issues which may impact life attainment and extension along with potential solutions to address these issues to ensure the life extension option is not precluded

  16. Atropisomerism about aryl-Csp(3) bonds: the electronic and steric influence of ortho-substituents on conformational exchange in cannabidiol and linderatin derivatives.

    Science.gov (United States)

    Berber, Hatice; Lameiras, Pedro; Denhez, Clément; Antheaume, Cyril; Clayden, Jonathan

    2014-07-03

    Terpenylation reactions of substituted phenols were used to prepare cannabidiol and linderatin derivatives, and their structure and conformational behavior in solution were investigated by NMR and, for some representative examples, by DFT. VT-NMR spectra and DFT calculations were used to determine the activation energies of the conformational change arising from restricted rotation about the aryl-Csp(3) bond that lead to two unequally populated rotameric epimers. The NBO calculation was applied to explain the electronic stabilization of one conformer over another by donor-acceptor charge transfer interactions. Conformational control arises from a combination of stereoelectronic and steric effects between substituents in close contact with each other on the two rings of the endocyclic epoxide atropisomers. This study represents the first exploration of the stereoelectronic origins of atropisomerism around C(sp(2))-C(sp(3)) single bonds through theoretical calculations.

  17. Decommissioning planning and the assessment of alternatives for the Hanford production reactors

    International Nuclear Information System (INIS)

    Miller, C.E. Jr.; Potter, R.F.

    1985-01-01

    Several years ago, the US Department of Energy began assessing alternatives and planning the decommissioning of eight shut-down plutonium production reactors located on the DOE Hanford Site in Washington State. The first of these graphite-moderated, water-cooled, reactors was built and started up in 1944 as part of the World War II Manhattan Project. The last of them started up in 1955. The eight reactors each operated for 12 to 24 years, with all eight operating simultaneously for about 10 years. In the 1960's, production needs declined and the reactors were one-by-one permanently shut down, the last of them in 1971. (A ninth Hanford production reactor, N Reactor, was started up in 1963; it is still operating and is not within the scope of the decommissioning planning and alternatives assessment work reported in this paper). This paper provides an overview description of the decommissioning plan for the eight shut-down Hanford production reactors and their associated fuel storage basins. Included are descriptions of the decommissioning alternatives considered for the facilities, along with discussions of National Environmental Policy Act (NEPA) process activities applicable to the Hanford decommissioning work. The criteria used in assessing decommissioning alternatives and the assumptions used in the decommissioning planning are identified. 4 refs., 8 figs., 3 tabs

  18. Optimizing the CSP Tower Air Brayton Cycle System to Meet the SunShot Objectives - Final Technical Report

    Energy Technology Data Exchange (ETDEWEB)

    Bryner, Elliott [Soutwest Research Inst., San Antonio, TX (United States); Brun, Klaus [Soutwest Research Inst., San Antonio, TX (United States); Coogan, Shane [Soutwest Research Inst., San Antonio, TX (United States); Cunningham, C. Seth [Soutwest Research Inst., San Antonio, TX (United States); Poerner, Nathan [Soutwest Research Inst., San Antonio, TX (United States)

    2016-02-26

    The objective of this project is to increase Concentrated Solar Power (CSP) tower air receiver and gas turbine temperature capabilities to 1,000ºC by the development of a novel gas turbine combustor, which can be integrated on a megawatt-scale gas turbine, such as the Solar Turbines Mercury 50™. No combustor technology currently available is compatible with the CSP application target inlet air temperature of 1,000°C. Autoignition and flashback at this temperature prevent the use of conventional lean pre-mix injectors that are currently employed to manage NOx emissions. Additional challenges are introduced by the variability of the high-temperature heat source provided by the field of solar collectors, the heliostat in CSP plants. For optimum energy generation from the power turbine, the turbine rotor inlet temperature (TRIT) should remain constant. As a result of changing heat load provided to the solar collector from the heliostat, the amount of energy input required from the combustion system must be adjusted to compensate. A novel multi-bank lean micro-mix injector has been designed and built to address the challenges of high-temperature combustion found in CSP applications. The multi-bank arrangement of the micro-mix injector selectively injects fuel to meet the heat addition requirements to maintain constant TRIT with changing solar load. To validate the design, operation, and performance of the multi-bank lean micro-mix injector, a novel combustion test facility has been designed and built at Southwest Research Institute® (SwRI®) in San Antonio, TX. This facility, located in the Turbomachinery Research Facility, provides in excess of two kilograms per second of compressed air at nearly eight bar pressure. A two-megawatt electric heater raises the inlet temperature to 800°C while a secondary gas-fired heater extends the operational temperature range of the facility to 1,000°C. A combustor test rig connected to the heater has been designed and built to

  19. On issues of constructing an exception handling mechanism for CSP-based process-oriented concurrent software

    NARCIS (Netherlands)

    Jovanovic, D.S.; Orlic, B.; Broenink, Johannes F.; Broenink, J.F.; Roebers, H.W.; Sunter, J.P.E.; Welch, P.H.; Wood, D.C.

    2005-01-01

    This paper discusses issues, possibilities and existing approaches for fitting an exception handling mechanism (EHM) in CSP-based process-oriented software architectures. After giving a survey on properties desired for a concurrent EHM, specific problems and a few principal ideas for including

  20. Application of probabilistic risk assessment to advanced liquid metal reactor designs

    International Nuclear Information System (INIS)

    Carroll, W.P.; Temme, M.I.

    1987-01-01

    The United States Department of Energy (US DOE) has been active in the development and application of probabilistic risk assessment methods within its liquid metal breeder reactor development program for the past eleven years. These methods have been applied to comparative risk evaluations, the selection of design features for reactor concepts, the selection and emphasis of research and development programs, and regulatory discussions. The application of probabilistic methods to reactors which are in the conceptual design stage presents unique data base, modeling, and timing challenges, and excellent opportunities to improve the final design. We provide here the background and insights on the experience which the US DOE liquid metal breeder reactor program has had in its application of probabilistic methods to the Clinch River Breeder Reactor Plant project, the Conceptual Design State of the Large Development Plant, and updates on this design. Plans for future applications of probabilistic risk assessment methods are also discussed. The US DOE is embarking on an innovative design program for liquid metal reactors. (author)

  1. Gas cooled reactor assessment. Volume II. Final report, February 9, 1976--June 30, 1976

    International Nuclear Information System (INIS)

    1976-08-01

    This report was prepared to document the estimated power plant capital and operating costs, and the safety and environmental assessments used in support of the Gas Cooled Reactor Assessment performed by Arthur D. Little, Inc. (ADL), for the U.S. Energy Research and Development Administration. The gas-cooled reactor technologies investigated include: the High Temperature Gas Reactor Steam Cycle (HTGR-SC), the HTGR Direct Cycle (HTGR-DC), the Very High Temperature Reactor (VHTR) and the Gas Cooled Fast Reactor (GCFR). Reference technologies used for comparison include: Light Water Reactors (LWR), the Liquid Metal Fast Breeder Reactor (LMFBR), conventional coal-fired steam plants, and coal combustion for process heat

  2. DESIGN OF A VIBRATION AND STRESS MEASUREMENT SYSTEM FOR AN ADVANCED POWER REACTOR 1400 REACTOR VESSEL INTERNALS COMPREHENSIVE VIBRATION ASSESSMENT PROGRAM

    OpenAIRE

    KO, DO-YOUNG; KIM, KYU-HYUNG

    2013-01-01

    In accordance with the US Nuclear Regulatory Commission (US NRC), Regulatory Guide 1.20, the reactor vessel internals comprehensive vibration assessment program (RVI CVAP) has been developed for an Advanced Power Reactor 1400 (APR1400). The purpose of the RVI CVAP is to verify the structural integrity of the reactor internals to flow-induced loads prior to commercial operation. The APR1400 RVI CVAP consists of four programs (analysis, measurement, inspection, and assessment). Thoughtful prepa...

  3. WTO Compliance Status of the Conservation Security Program (CSP) and the Conservation Reserve Program (CRP)

    National Research Council Canada - National Science Library

    Schnepf, Randy

    2007-01-01

    .... This report is not a legal opinion, but describes both the CSP and CRP programs, the WTO Annex II provisions that govern compliance, and the potential issues involved in evaluating the compliance status of the two programs. This report will be updated as events warrant.

  4. Rhenium-catalyzed dehydrogenative olefination of C(sp(3))-H bonds with hypervalent iodine(III) reagents.

    Science.gov (United States)

    Gu, Haidong; Wang, Congyang

    2015-06-07

    A dehydrogenative olefination of C(sp(3))-H bonds is disclosed here, by merging rhenium catalysis with an alanine-derived hypervalent iodine(III) reagent. Thus, cyclic and acyclic ethers, toluene derivatives, cycloalkanes, and nitriles are all successfully alkenylated in a regio- and stereoselective manner.

  5. Assessment methodology applicable to safe decommissioning of Romanian VVR-S research reactor

    International Nuclear Information System (INIS)

    Baniu, O.; Vladescu, G.; Vidican, D.; Penescu, M.

    2002-01-01

    The paper contains the results of research activity performed by CITON specialists regarding the assessment methodology intended to be applied to safe decommissioning of the research reactors, developed taking into account specific conditions of the Romanian VVR-S Research Reactor. The Romanian VVR-S Research Reactor is an old reactor (1957) and its Decommissioning Plan is under study. The main topics of paper are as follows: Safety approach of nuclear facilities decommissioning. Applicable safety principles; Main steps of the proposed assessment methodology; Generic content of Decommissioning Plan. Main decommissioning activities. Discussion about the proposed Decommissioning Plan for Romanian Research Reactor; Safety risks which may occur during decommissioning activities. Normal decommissioning operations. Fault conditions. Internal and external hazards; Typical development of a scenario. Features, Events and Processes List. Exposure pathways. Calculation methodology. (author)

  6. Double mutation of cell wall proteins CspB and PBP1a increases secretion of the antibody Fab fragment from Corynebacterium glutamicum

    Science.gov (United States)

    2014-01-01

    Background Among other advantages, recombinant antibody-binding fragments (Fabs) hold great clinical and commercial potential, owing to their efficient tissue penetration compared to that of full-length IgGs. Although production of recombinant Fab using microbial expression systems has been reported, yields of active Fab have not been satisfactory. We recently developed the Corynebacterium glutamicum protein expression system (CORYNEX®) and demonstrated improved yield and purity for some applications, although the system has not been applied to Fab production. Results The Fab fragment of human anti-HER2 was successfully secreted by the CORYNEX® system using the conventional C. glutamicum strain YDK010, but the productivity was very low. To improve the secretion efficiency, we investigated the effects of deleting cell wall-related genes. Fab secretion was increased 5.2 times by deletion of pbp1a, encoding one of the penicillin-binding proteins (PBP1a), mediating cell wall peptidoglycan (PG) synthesis. However, this Δpbp1a mutation did not improve Fab secretion in the wild-type ATCC13869 strain. Because YDK010 carries a mutation in the cspB gene encoding a surface (S)-layer protein, we evaluated the effect of ΔcspB mutation on Fab secretion from ATCC13869. The Δpbp1a mutation showed a positive effect on Fab secretion only in combination with the ΔcspB mutation. The ΔcspBΔpbp1a double mutant showed much greater sensitivity to lysozyme than either single mutant or the wild-type strain, suggesting that these mutations reduced cell wall resistance to protein secretion. Conclusion There are at least two crucial permeability barriers to Fab secretion in the cell surface structure of C. glutamicum, the PG layer, and the S-layer. The ΔcspBΔpbp1a double mutant allows efficient Fab production using the CORYNEX® system. PMID:24731213

  7. A view of technology maturity assessment to realize fusion reactor by Japanese young researchers

    International Nuclear Information System (INIS)

    Kasada, Ryuta; Goto, Takuya; Miyazawa, Junichi; Fujioka, Shinsuke; Hiwatari, Ryoji; Oyama, Naoyuki; Tanigawa, Hiroyasu

    2013-01-01

    Japanese young researchers who have interest in realizing fusion reactor have analyzed Technology Readiness Levels (TRL) in Young Scientists Special Interest Group on Fusion Reactor Realization. In this report, brief introduction to TRL assessment and a view of TRL assessment against fusion reactor projects conducting in Japan. (J.P.N.)

  8. Guidelines for the Review of Research Reactor Safety: Revised Edition. Reference Document for IAEA Integrated Safety Assessment of Research Reactors (INSARR)

    International Nuclear Information System (INIS)

    2013-01-01

    The Integrated Safety Assessment of Research Reactors (INSARR) is an IAEA safety review service available to Member States with the objective of supporting them in ensuring and enhancing the safety of their research reactors. This service consists of performing a comprehensive peer review and an assessment of the safety of the respective research reactor. The reviews are based on IAEA safety standards and on the provisions of the Code of Conduct on the Safety of Research Reactors. The INSARR can benefit both the operating organizations and the regulatory bodies of the requesting Member States, and can include new research reactors under design or operating research reactors, including those which are under a Project and Supply Agreement with the IAEA. The first IAEA safety evaluation of a research reactor operated by a Member State was completed in October 1959 and involved the Swiss 20 MW DIORIT research reactor. Since then, and in accordance with its programme on research reactor safety, the IAEA has conducted safety review missions in its Member States to enhance the safety of their research reactor facilities through the application of the Code of Conduct on the Safety of Research Reactors and the relevant IAEA safety standards. About 320 missions in 51 Member States were undertaken between 1972 and 2012. The INSARR missions and other limited scope safety review missions are conducted following the guidelines presented in this publication, which is a revision of Guidelines for the Review of Research Reactor Safety (IAEA Services Series No. 1), published in December 1997. This publication details those IAEA safety standards and guidance publications relevant to the safety of research reactors that have been revised or published since 1997. The purpose of this publication is to give guidance on the preparation, implementation, reporting and follow-up of safety review missions. It is also intended to be of assistance to operators and regulators in conducting

  9. Power probability density function control and performance assessment of a nuclear research reactor

    International Nuclear Information System (INIS)

    Abharian, Amir Esmaeili; Fadaei, Amir Hosein

    2014-01-01

    Highlights: • In this paper, the performance assessment of static PDF control system is discussed. • The reactor PDF model is set up based on the B-spline functions. • Acquaints of Nu, and Th-h. equations solve concurrently by reformed Hansen’s method. • A principle of performance assessment is put forward for the PDF of the NR control. - Abstract: One of the main issues in controlling a system is to keep track of the conditions of the system function. The performance condition of the system should be inspected continuously, to keep the system in reliable working condition. In this study, the nuclear reactor is considered as a complicated system and a principle of performance assessment is used for analyzing the performance of the power probability density function (PDF) of the nuclear research reactor control. First, the model of the power PDF is set up, then the controller is designed to make the power PDF for tracing the given shape, that make the reactor to be a closed-loop system. The operating data of the closed-loop reactor are used to assess the control performance with the performance assessment criteria. The modeling, controller design and the performance assessment of the power PDF are all applied to the control of Tehran Research Reactor (TRR) power in a nuclear process. In this paper, the performance assessment of the static PDF control system is discussed, the efficacy and efficiency of the proposed method are investigated, and finally its reliability is proven

  10. Lifetime assessment on PWR reactor vessel internals in Korea

    International Nuclear Information System (INIS)

    Jung, Sung-Gyu; Jin, Tae-Eun; Jeong, Ill-Seok

    2002-01-01

    In order to extend the operating time of the Kori Unit 1 reactor internals, a comprehensive review of the potential ageing problems and a safety assessment have been performed. As the plant ages, reactor internal components which are subject to various ageing mechanism should be identified and evaluated based on the systematic technical procedure. In this respect, technical procedure for lifetime evaluation had been developed and applied to reactor internals. This paper describes a overall assessment and ageing management procedure and evaluation results for reactor internals. Also this paper suggests the optimal ageing management programs to maintain the integrity of reactor internals beyond design life based on the evaluation results. A review of all known potential ageing mechanisms was performed for each of the reactor internal subcomponents. From these results, 8 ageing mechanisms such as void swelling, irradiation and thermal embrittlement, fatigue, stress corrosion cracking, IASCC, stress relaxation, and wear for the reactor internal components were expected to be of major concerns during the current or extended plant life. In this study, 8 ageing mechanisms were identified for lifetime evaluation. For these ageing mechanisms, lifetime assessment was performed. As a result of this evaluation, it is expected that core barrel will exceed the IASCC threshold value during 40 operating years, and baffle/former and baffle former bolts will exceed the threshold value for void swelling, irradiation embrittlement, IASCC, stress relaxation during 40 operating years. However, for all other reactor internals subcomponents, thermal embrittlement, fatigue, SCC, and wear were identified as nonsignificant. As a result of lifetime evaluations, 4 ageing mechanisms were established to be plausible for 3 subcomponents. These results are shown. The existing ageing management programs (AMPs) for Kori Unit 1, such as ISI, water chemistry control, rod drop time testing etc., were

  11. Assessment of the integrity of WWER type reactor pressure vessels

    International Nuclear Information System (INIS)

    Brumovsky, M.

    1995-01-01

    Procedures are given for the assessment of the residual lifetime of reactor pressure vessels with respect to a sudden failure, the lifetime of vessels with defects disclosed during in-service inspections, and the fatigue or corrosion-mechanical lifetime. Also outlined are the ways of assessing the effects of major degradation mechanisms, i.e. radiation embrittlement, thermal aging, and fatigue damage, including the use of calculated values and experimental examination, by means of surveillance specimens in particular. All results of assessment performed so far indicate that the life of reactor pressure vessels at the Dukovany, Jaslovske Bohunice, and Temelin nuclear power plants is well secured. 7 figs., 3 refs

  12. 77 FR 66649 - Proposed Revision to Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors

    Science.gov (United States)

    2012-11-06

    ... and Severe Accident Evaluation for New Reactors AGENCY: Nuclear Regulatory Commission. ACTION... Assessment and Severe Accident Evaluation for New Reactors.'' The NRC is extending the public comment period... assessment (PRA) information and severe accident assessments for new reactors submitted to support design...

  13. Cost assessment of demo fusion reactor with considering maintenance

    International Nuclear Information System (INIS)

    Hashizume, Hidetoshi; Kitagoh, Kazutoshi

    2003-01-01

    The purpose of this study is to perform cost assessment of nuclear fusion reactors in order to draw up commercial plants. A fusion reactor may have a complex configuration to achieve high beta value, which leads to low and instable availability when maintenance is taken into account. Therefore, reactor's availability must be evaluated with considering the influence of the configuration complexity. Furthermore the availability has the strong impact on COE (Cost of Electricity), that is, a fusion reactor with low availability will not be accepted as a commercial plant. Therefore, we developed a new method to calculate availabilities with random numbers, in which the complexity of reactor's configuration could become considered. In addition, we considered the reduction of superconducting coil's maintenance time by introducing remountable magnet system because the coil maintenance requires quite long time in the present technology. The results show that the availability becomes relatively large if the short maintenance time of coils could be achieved, for example, by remountable magnetic systems. (author)

  14. Proliferation resistance assessment of high temperature gas reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chikamatsu N, M. A. [Instituto Tecnologico y de Estudios Superiores de Monterrey, Campus Santa Fe, Av. Carlos Lazo No. 100, Santa Fe, 01389 Mexico D. F. (Mexico); Puente E, F., E-mail: midori.chika@gmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    The Generation IV International Forum has established different objectives for the new generation of reactors to accomplish. These objectives are focused on sustain ability, safety, economics and proliferation resistance. This paper is focused on how the proliferation resistance of the High Temperature Gas Reactors (HTGR) is assessed and the advantages that these reactors present currently. In this paper, the focus will be on explaining why such reactors, HTGR, can achieve the goals established by the GIF and can present a viable option in terms of proliferation resistance, which is an issue of great importance in the field of nuclear energy generation. The reason why the HTGR are being targeted in this writing is that these reactors are versatile, and present different options from modular reactors to reactors with the same size as the ones that are being operated today. Besides their versatility, the HTGR has designed features that might improve on the overall sustain ability of the nuclear reactors. This is because the type of safety features and materials that are used open up options for industrial processes to be carried out; cogeneration for instance. There is a small section that mentions how HTGR s are being developed in the international sector in order to present the current world view in this type of technology and the further developments that are being sought. For the proliferation resistance section, the focus is on both the intrinsic and the extrinsic features of the nuclear systems. The paper presents a comparison between the features of Light Water Reactors (LWR) and the HTGR in order to be able to properly compare the most used technology today and one that is gaining international interest. (Author)

  15. Proliferation resistance assessment of high temperature gas reactors

    International Nuclear Information System (INIS)

    Chikamatsu N, M. A.; Puente E, F.

    2014-10-01

    The Generation IV International Forum has established different objectives for the new generation of reactors to accomplish. These objectives are focused on sustain ability, safety, economics and proliferation resistance. This paper is focused on how the proliferation resistance of the High Temperature Gas Reactors (HTGR) is assessed and the advantages that these reactors present currently. In this paper, the focus will be on explaining why such reactors, HTGR, can achieve the goals established by the GIF and can present a viable option in terms of proliferation resistance, which is an issue of great importance in the field of nuclear energy generation. The reason why the HTGR are being targeted in this writing is that these reactors are versatile, and present different options from modular reactors to reactors with the same size as the ones that are being operated today. Besides their versatility, the HTGR has designed features that might improve on the overall sustain ability of the nuclear reactors. This is because the type of safety features and materials that are used open up options for industrial processes to be carried out; cogeneration for instance. There is a small section that mentions how HTGR s are being developed in the international sector in order to present the current world view in this type of technology and the further developments that are being sought. For the proliferation resistance section, the focus is on both the intrinsic and the extrinsic features of the nuclear systems. The paper presents a comparison between the features of Light Water Reactors (LWR) and the HTGR in order to be able to properly compare the most used technology today and one that is gaining international interest. (Author)

  16. Assessment of benefits of research reactors in less developed countries. A case study of the Dalat reactor in Vietnam

    International Nuclear Information System (INIS)

    Hien, P.D.

    1999-01-01

    The analysis of data on nuclear research reactor (NRR) and socio-economic conditions across countries reveals highly significant relationships of reactor power with GDP and R and D expenditure. The trends revealed can be used as preliminary guides for feasibility assessment of investment in a NRR. Concerning reactor performance, i.e. the number of reactor operation days per year, the covariation with R and D expenditure is most significant, but moderate, implying that there are other controlling factors, e.g. the engagement of country in nuclear power development. Thus, the size of the R and D fund is a most significant indicator to look at in reactor planning. Unfortunately, the lack of adequate R and D funding is a common and chronic problem in less developed countries. As NRR is among the biggest R and D investment in less developed countries, adequate cost benefit assessment is rightfully required. In the case of Vietnam, during 15 years of operation of a 500 kW NRR 2300 Ci of radioisotopes were delivered and 45,000 samples were analysed for multielemental compositions. From a pure financial viewpoint these figures would still be insignificant to justify the investment. However, the impact of the reactor on the technological development seems not to be a matter of pro and cons. The status of reactor utilization and lessons learned are presented and discussed. (author)

  17. Assessment of benefits of research reactors in less developed countries. A case study of the Dalat reactor in Vietnam

    Energy Technology Data Exchange (ETDEWEB)

    Hien, P.D. [Vietnam Atomic Energy Agency, Hanoi (Viet Nam)

    1999-08-01

    The analysis of data on nuclear research reactor (NRR) and socio-economic conditions across countries reveals highly significant relationships of reactor power with GDP and R and D expenditure. The trends revealed can be used as preliminary guides for feasibility assessment of investment in a NRR. Concerning reactor performance, i.e. the number of reactor operation days per year, the covariation with R and D expenditure is most significant, but moderate, implying that there are other controlling factors, e.g. the engagement of country in nuclear power development. Thus, the size of the R and D fund is a most significant indicator to look at in reactor planning. Unfortunately, the lack of adequate R and D funding is a common and chronic problem in less developed countries. As NRR is among the biggest R and D investment in less developed countries, adequate cost benefit assessment is rightfully required. In the case of Vietnam, during 15 years of operation of a 500 kW NRR 2300 Ci of radioisotopes were delivered and 45,000 samples were analysed for multielemental compositions. From a pure financial viewpoint these figures would still be insignificant to justify the investment. However, the impact of the reactor on the technological development seems not to be a matter of pro and cons. The status of reactor utilization and lessons learned are presented and discussed. (author)

  18. Safety assessment of research reactors and preparation of the safety analysis report

    International Nuclear Information System (INIS)

    1994-01-01

    This Safety Guide presents guidelines, approved by international consensus, for the preparation, review and assessment of safety documentation for research reactors such as the Safety Analysis Report. While the Guide is most applicable to research reactors in the design and construction stage, it is also recommended for use during relicensing or reassessment of existing reactors

  19. Assessment of Sensor Technologies for Advanced Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Korsah, Kofi [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ramuhalli, Pradeep [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Vlim, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Kisner, Roger A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Britton, Jr, Charles L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wootan, D. W. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Anheier, Jr, N. C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Diaz, A. A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hirt, E. H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Chien, H. T. [Argonne National Lab. (ANL), Argonne, IL (United States); Sheen, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Bakhtiari, Sasan [Argonne National Lab. (ANL), Argonne, IL (United States); Gopalsami, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Heifetz, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Tam, S. W. [Argonne National Lab. (ANL), Argonne, IL (United States); Park, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Upadhyaya, B. R. [Univ. of Tennessee, Knoxville, TN (United States); Stanford, A. [Univ. of Tennessee, Knoxville, TN (United States)

    2016-10-01

    Sensors and measurement technologies provide information on processes, support operations and provide indications of component health. They are therefore crucial to plant operations and to commercialization of advanced reactors (AdvRx). This report, developed by a three-laboratory team consisting of Argonne National Laboratory (ANL), Oak Ridge National Laboratory (ORNL) and Pacific Northwest National Laboratory (PNNL), provides an assessment of sensor technologies and a determination of measurement needs for AdvRx. It provides the technical basis for identifying and prioritizing research targets within the instrumentation and control (I&C) Technology Area under the Department of Energy’s (DOE’s) Advanced Reactor Technology (ART) program and contributes to the design and implementation of AdvRx concepts.

  20. Assessment of structural materials inside the reactor pool of the Dalat research reactor

    International Nuclear Information System (INIS)

    Nguyen Nhi Dien; Luong Ba Vien; Nguyen Minh Tuan; Trang Cao Su

    2010-01-01

    Originally the Dalat Nuclear Research Reactor (DNRR) was a 250-kW TRIGA MARK II reactor, started building from early 1960s and achieved the first criticality on February 26, 1963. During the 1982-1984 period, the reactor was reconstructed and upgraded to 500kW, and restarted operation on March 20, 1984. From the original TRIGA reactor, only the pool liner, beam ports, thermal columns, and graphite reflector have been remained. The structural materials of pool liner and other components of TRIGA were made of aluminum alloy 6061 and aluminum cladding fuel assemblies. Some other parts, such as reactor core, irradiation rotary rack around the core, vertical irradiation facilities, etc. were replaced by the former Soviet Union's design with structural materials of aluminum alloy CAV-1. WWR-M2 fuel assemblies of U-Al alloy 36% and 19.75% 235 U enrichment and aluminum cladding have been used. In its original version, the reactor was setting upon an all-welded aluminum frame supported by four legs attached to the bottom of the pool. After the modification made, the new core is now suspended from the top of the pool liner by means of three aluminum concentric cylindrical shells. The upper one has a diameter of 1.9m, a length of 3.5m and a thickness of 10mm. This shell prevents from any visual access to the upper part of the pool liner, but is provided with some holes to facilitate water circulation in the 4cm gap between itself and the reactor pool liner. The lower cylindrical shells act as an extracting well for water circulation. As reactor has been operated at low power of 500 kW, it was no any problem with degradation of core structural materials due to neutron irradiation and thermal heat, but there are some ageing issues with aluminum liner and other structures (for example, corrosion of tightening-up steel bolt in the fourth beam port and flood of neutron detector housing) inside the reactor pool. In this report, the authors give an overview and assessment of

  1. 105-B Reactor museum feasibility assessment (Phase 2) project

    International Nuclear Information System (INIS)

    Heckel, R. P.

    2000-01-01

    This 105-B Reactor Museum feasibility assessment project report documents project activities that have been performed, including a review and assessment of previously existing information, a walk-through of the facility, an assessment of potential hazards, and selection of mitigative measures deemed to be appropriate to allow unescorted access by members of the public to a specified primary tour route

  2. Draft genome of Kocuria polaris CMS 76or(T) isolated from cyanobacterial mats, McMurdo Dry Valley, Antarctica: an insight into CspA family of proteins from Kocuria polaris CMS 76or(T).

    Science.gov (United States)

    Gundlapally, Sathyanarayana Reddy; Ara, Srinivas; Sisinthy, Shivaji

    2015-10-01

    Kocuria polaris strain CMS 76or(T) is a gram-positive, orange-pigmented bacterium isolated from a cyanobacterial mat sample from a pond located in McMurdo Dry Valley, Antarctica. It is psychrotolerant, orange pigmented, hydrolyses starch and Tween 80 and reduces nitrate. We report the 3.78-Mb genome of K. polaris strain CMS 76or(T), containing 3416 coding sequences, including one each for 5S rRNA, 23S rRNA, 16S rRNA and 47 tRNA genes, and the G+C content of DNA is 72.8%. An investigation of Csp family of proteins from K. polaris strain CMS 76or(T) indicated that it contains three different proteins of CspA (peg.319, peg.2255 and 2832) and the length varied from 67 to 69 amino acids. The three different proteins contain all the signature amino acids and two RNA binding regions that are characteristic of CspA proteins. Further, the CspA from K. polaris strain CMS 76or(T) was different from CspA of four other species of the genus Kocuria, Cryobacterium roopkundense and E. coli indirectly suggesting the role of CspA of K. polaris strain CMS 76or(T) in psychrotolerant growth of the bacterium.

  3. Socio-economic effects of a HYSOL CSP plant located in different countries: An input output analysis

    NARCIS (Netherlands)

    Corona, B.; López, A.; San Miguel, G.

    2016-01-01

    The aim of this paper is to estimate the socioeconomic effects associated with the production of electricity by a CSP plant with HYSOL configuration, using Input Output Analysis. These effects have been estimated in terms of production of Goods and Services (G&S), multiplier effect, value added,

  4. Assessment of very high-temperature reactors in process application. Appendix I. Evaluation of the reactor system

    International Nuclear Information System (INIS)

    Jones, J.E. Jr.; Spiewak, I.

    1976-12-01

    In April 1974, the U.S. Atomic Energy Commission [now the Energy Research and Development Administration (ERDA)] authorized General Atomic Company, General Electric Company, and Westinghouse Electric Corp., Astronuclear Laboratory, to assess the available technology for producing heat using very high-temperature nuclear reactors. An evaulation of these studies and of the technical and economic potential of very high-temperature reactors (VHTR) is presented. The VHTR is a helium-cooled graphite-moderated reactor. The concepts and technology are evaluated for producing process stream temperatures of 649, 760, 871, 982, and 1093 0 C (1200, 1400, 1600, 1800, and 2000 0 F). There are a number of large industrial process heat applications that could utilize the VHTR

  5. 77 FR 61446 - Proposed Revision Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors

    Science.gov (United States)

    2012-10-09

    ... Severe Accident Evaluation for New Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Standard... its Standard Review Plan (SRP), Section 19.0, ``Probabilistic Risk Assessment and Severe Accident... assessment (PRA) information and severe accident assessments for new reactors submitted to support design...

  6. N reactor external events probabilistic risk assessment

    International Nuclear Information System (INIS)

    Baxter, J.T.

    1989-01-01

    An external events probabilistic risk assessment of the N Reactor has been completed. The methods used are those currently being proposed for external events analysis in NUREG-1150. Results are presented for the external hazards that survived preliminary screening. They are earthquake, fire, and external flood. Core damage frequencies for these hazards are shown to be comparable to those for commercial pressurized water reactors. Dominant fire sequences are described and related to 10 CFR 50, Appendix R design requirements. Potential remedial measures that reduce fire core damage risk are described including modifications to fire protection systems, procedure changes, and addition of new administrative controls. Dominant seismic sequences are described. The effect of non-safety support system dependencies on seismic risk is presented

  7. Improving the Reliability and Modal Stability of High Power 870 nm AlGaAs CSP Laser Diodes for Applications to Free Space Communication Systems

    Science.gov (United States)

    Connolly, J. C.; Alphonse, G. A.; Carlin, D. B.; Ettenberg, M.

    1991-01-01

    The operating characteristics (power-current, beam divergence, etc.) and reliability assessment of high-power CSP lasers is discussed. The emission wavelength of these lasers was optimized at 860 to 880 nm. The operational characteristics of a new laser, the inverse channel substrate planar (ICSP) laser, grown by metalorganic chemical vapor deposition (MOCVD), is discussed and the reliability assessment of this laser is reported. The highlights of this study include a reduction in the threshold current value for the laser to 15 mA and a degradation rate of less than 2 kW/hr for the lasers operating at 60 mW of peak output power.

  8. Preliminary nuclear power reactor technology qualitative assessment for Malaysia

    International Nuclear Information System (INIS)

    Shamsul Amri Sulaiman

    2011-01-01

    Since the worlds first nuclear reactor major breakthrough in December 02, 1942, the nuclear power industry has undergone tremendous development and evolution for more than half a century. After surpassing moratorium of nuclear power plant construction caused by catastrophic accidents at Three-mile island (1979) and Chernobyl (1986), today, nuclear energy is back on the policy agendas of many countries, both developed and developing, signaling nuclear revival or nuclear renaissance. Selection of suitable nuclear power technology has thus been subjected to primary attention. This short paper attempts to draw preliminary technology assessment for the first nuclear power reactor technology for Malaysia. Methodology employed is qualitative analysis collating recent finding of tnb-kepco preliminary feasibility study for nuclear power program in peninsular malaysia and other published presentations and/or papers by multiple experts. The results suggested that pressurized water reactor (PWR) is the prevailing technology in terms of numbers and plant performances, and while the commercialization of generation IV reactors is remote (e.g. Not until 2030), generation III/ III+ NPP models are commercially available on the market today. Five (5) major steps involved in reactor technology selection were introduced with a focus on introducing important aspects of selection criteria. Three (3) categories for the of reactor technology selection were used for the cursory evaluation. The outcome of these analyses shall lead to deeper and full analyses of the recommended reactor technologies for a comprehensive feasibility study in the near future. Recommendations for reactor technology option were also provided for both strategic and technical recommendations. The paper shall also implore the best way to select systematically the first civilian nuclear power reactor. (Author)

  9. 75 FR 39220 - Charter Schools Program (CSP) Grants for Replication and Expansion of High-Quality Charter Schools

    Science.gov (United States)

    2010-07-08

    ... DEPARTMENT OF EDUCATION Charter Schools Program (CSP) Grants for Replication and Expansion of High-Quality Charter Schools AGENCY: Office of Innovation and Improvement, Department of Education. ACTION... notice inviting applications for new awards for FY 2010 for the Charter Schools Program Grants for...

  10. DNA prime/Adenovirus boost malaria vaccine encoding P. falciparum CSP and AMA1 induces sterile protection associated with cell-mediated immunity.

    Directory of Open Access Journals (Sweden)

    Ilin Chuang

    Full Text Available BACKGROUND: Gene-based vaccination using prime/boost regimens protects animals and humans against malaria, inducing cell-mediated responses that in animal models target liver stage malaria parasites. We tested a DNA prime/adenovirus boost malaria vaccine in a Phase 1 clinical trial with controlled human malaria infection. METHODOLOGY/PRINCIPAL FINDINGS: The vaccine regimen was three monthly doses of two DNA plasmids (DNA followed four months later by a single boost with two non-replicating human serotype 5 adenovirus vectors (Ad. The constructs encoded genes expressing P. falciparum circumsporozoite protein (CSP and apical membrane antigen-1 (AMA1. The regimen was safe and well-tolerated, with mostly mild adverse events that occurred at the site of injection. Only one AE (diarrhea, possibly related to immunization, was severe (Grade 3, preventing daily activities. Four weeks after the Ad boost, 15 study subjects were challenged with P. falciparum sporozoites by mosquito bite, and four (27% were sterilely protected. Antibody responses by ELISA rose after Ad boost but were low (CSP geometric mean titer 210, range 44-817; AMA1 geometric mean micrograms/milliliter 11.9, range 1.5-102 and were not associated with protection. Ex vivo IFN-γ ELISpot responses after Ad boost were modest (CSP geometric mean spot forming cells/million peripheral blood mononuclear cells 86, range 13-408; AMA1 348, range 88-1270 and were highest in three protected subjects. ELISpot responses to AMA1 were significantly associated with protection (p = 0.019. Flow cytometry identified predominant IFN-γ mono-secreting CD8+ T cell responses in three protected subjects. No subjects with high pre-existing anti-Ad5 neutralizing antibodies were protected but the association was not statistically significant. SIGNIFICANCE: The DNA/Ad regimen provided the highest sterile immunity achieved against malaria following immunization with a gene-based subunit vaccine (27%. Protection

  11. Risk assessment of computer-controlled safety systems for fusion reactors

    International Nuclear Information System (INIS)

    Fryer, M.O.; Bruske, S.Z.

    1983-01-01

    The complexity of fusion reactor systems and the need to display, analyze, and react promptly to large amounts of information during reactor operation will require a number of safety systems in the fusion facilities to be computer controlled. Computer software, therefore, must be included in the reactor safety analyses. Unfortunately, the science of integrating computer software into safety analyses is in its infancy. Combined plant hardware and computer software systems are often treated by making simple assumptions about software performance. This method is not acceptable for assessing risks in the complex fusion systems, and a new technique for risk assessment of combined plant hardware and computer software systems has been developed. This technique is an extension of the traditional fault tree analysis and uses structured flow charts of the software in a manner analogous to wiring or piping diagrams of hardware. The software logic determines the form of much of the fault trees

  12. Advanced Reactor PSA Methodologies for System Reliability Analysis and Source Term Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, D.; Brunett, A.; Passerini, S.; Grelle, A.; Bucknor, M.

    2017-06-26

    Beginning in 2015, a project was initiated to update and modernize the probabilistic safety assessment (PSA) of the GE-Hitachi PRISM sodium fast reactor. This project is a collaboration between GE-Hitachi and Argonne National Laboratory (Argonne), and funded in part by the U.S. Department of Energy. Specifically, the role of Argonne is to assess the reliability of passive safety systems, complete a mechanistic source term calculation, and provide component reliability estimates. The assessment of passive system reliability focused on the performance of the Reactor Vessel Auxiliary Cooling System (RVACS) and the inherent reactivity feedback mechanisms of the metal fuel core. The mechanistic source term assessment attempted to provide a sequence specific source term evaluation to quantify offsite consequences. Lastly, the reliability assessment focused on components specific to the sodium fast reactor, including electromagnetic pumps, intermediate heat exchangers, the steam generator, and sodium valves and piping.

  13. Experimental assessment of computer codes used for safety analysis of integral reactors

    Energy Technology Data Exchange (ETDEWEB)

    Falkov, A.A.; Kuul, V.S.; Samoilov, O.B. [OKB Mechanical Engineering, Nizhny Novgorod (Russian Federation)

    1995-09-01

    Peculiarities of integral reactor thermohydraulics in accidents are associated with presence of noncondensable gas in built-in pressurizer, absence of pumped ECCS, use of guard vessel for LOCAs localisation and passive RHRS through in-reactor HX`s. These features defined the main trends in experimental investigations and verification efforts for computer codes applied. The paper reviews briefly the performed experimental investigation of thermohydraulics of AST-500, VPBER600-type integral reactors. The characteristic of UROVEN/MB-3 code for LOCAs analysis in integral reactors and results of its verification are given. The assessment of RELAP5/mod3 applicability for accident analysis in integral reactor is presented.

  14. Assessment of torsatrons as reactors

    International Nuclear Information System (INIS)

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

    1992-12-01

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

  15. Regulatory Assessment Technologies for Aging of Reactor Vessel Internals

    Energy Technology Data Exchange (ETDEWEB)

    Jhung, Myung Jo; Park, Jeong Soon; Ko, Hanok [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-10-15

    In order to develop the audit calculation system, it is required to develop crack evaluation, seismic analysis and thermal-hydraulic analysis techniques for RVIs so that integrity of RVIs under the aging environment can be evaluated and be assured. In addition, regulatory requirements including safety review and inspection guides should be developed in order to assure the quality and uniformity of safety reviews and inspections regarding aging assessment and management of RVIs. Since Reactor Vessel Internals (RVIs) are installed within the reactor pressure vessel and surround the fuel assemblies, some of them are exposed to the environment such as high neutron irradiation, high temperature and reactor coolant flow. Those environmental factors can cause damage to RVIs including cracks, loss of material, fatigue, loss of fracture toughness and change of dimension as the operation time of nuclear power plants (NPPs) increases. For long-term operation more than 40 years, aging management of RVIs is important. The final objectives of this study are to establish the audit calculation system for RVIs and to develop regulatory requirements for aging assessment and management of RVIs considering their operating conditions, materials, and possible aging mechanisms.

  16. Development and implementation of a dynamic TES dispatch control component in a PV-CSP techno-economic performance modelling tool

    Science.gov (United States)

    Hansson, Linus; Guédez, Rafael; Larchet, Kevin; Laumert, Bjorn

    2017-06-01

    The dispatchability offered by thermal energy storage (TES) in concentrated solar power (CSP) and solar hybrid plants based on such technology presents the most important difference compared to power generation based only on photovoltaics (PV). This has also been one reason for recent hybridization efforts of the two technologies and the creation of Power Purchase Agreement (PPA) payment schemes based on offering higher payment multiples during daily hours of higher (peak or priority) demand. Recent studies involving plant-level thermal energy storage control strategies are however to a large extent based on pre-determined approaches, thereby not taking into account the actual dynamics of thermal energy storage system operation. In this study, the implementation of a dynamic dispatch strategy in the form of a TRNSYS controller for hybrid PV-CSP plants in the power-plant modelling tool DYESOPT is presented. In doing this it was attempted to gauge the benefits of incorporating a day-ahead approach to dispatch control compared to a fully pre-determined approach determining hourly dispatch only once prior to annual simulation. By implementing a dynamic strategy, it was found possible to enhance technical and economic performance for CSP-only plants designed for peaking operation and featuring low values of the solar multiple. This was achieved by enhancing dispatch control, primarily by taking storage levels at the beginning of every simulation day into account. The sequential prediction of the TES level could therefore be improved, notably for evaluated plants without integrated PV, for which the predicted storage levels deviated less than when PV was present in the design. While also featuring dispatch performance gains, optimal plant configurations for hybrid PV-CSP was found to present a trade-off in economic performance in the form of an increase in break-even electricity price when using the dynamic strategy which was offset to some extent by a reduction in

  17. No More "Magic Aprons": Longitudinal Assessment and Continuous Improvement of Customer Service at the University of North Dakota Libraries

    Science.gov (United States)

    Clark, Karlene T.; Walker, Stephanie R.

    2017-01-01

    The University of North Dakota (UND) Libraries have developed a multi-award winning Customer Service Program (CSP) involving longitudinal assessment and continuous improvement. The CSP consists of iterative training modules; constant reinforcement of Customer Service Principles with multiple communication strategies and tools, and incentives that…

  18. An independent safety assessment of Department of Energy nuclear reactor facilities: Procedures, operations and maintenance

    International Nuclear Information System (INIS)

    Toto, G.; Lindgren, A.J.

    1981-02-01

    The 1979 accident at the Three Mile Island commercial nuclear power plant has led to a number of studies of nuclear reactors, in both the public and private sectors. One of these is that of the Department of Energy's (DOE) Nuclear Facilities Personnel Qualification and Training (NFPQT) Committee, which has outlined tasks for assessment of 13 reactors owned by DOE and operated by contractors. This report covers one of the tasks, the assessment of procedures, operations, and maintenance at the DOE reactor facilities, based on a review of actual documents used at the reactor sites

  19. Assessing current and future techno-economic potential of concentrated solar power and photovoltaic electricity generation

    International Nuclear Information System (INIS)

    Köberle, Alexandre C.; Gernaat, David E.H.J.; Vuuren, Detlef P. van

    2015-01-01

    CSP and PV technologies represent energy sources with large potentials. We present cost-supply curves for both technologies using a consistent methodology for 26 regions, based on geoexplicit information on solar radiation, land cover type and slope, exploring individual potential and interdependencies. For present day, both CSP and PV supply curves start at $0.18/kWh, in North Africa, South America, and Australia. Applying accepted learning rates to official capacity targets, we project prices to drop to $0.11/kWh for both technologies by 2050. In an alternative “fast-learning” scenario, generation costs drop to $0.06–0.07/kWh for CSP, and $0.09/kWh for PV. Competition between them for best areas is explored along with sensitivities of their techno-economic potentials to land use restrictions and land cover type. CSP was found to be more competitive in desert sites with highest direct solar radiation. PV was a clear winner in humid tropical regions, and temperate northern hemisphere. Elsewhere, no clear winner emerged, highlighting the importance of competition in assessments of potentials. Our results show there is ample potential globally for both technologies even accounting for land use restrictions, but stronger support for RD&D and higher investments are needed to make CSP and PV cost-competitive with established power technologies by 2050. - Highlights: • A consistent assessment of global potential for CSP and PV, with cost-supply curves for 26 regions. • Combined global CSP and PV potential below US$0.35/kWh estimated at 135,128 TWh per year. • Competition for same land-based solar resource implies that potentials cannot be added. • Attractive areas are MENA, Northern Chile, Australia, China and Southwestern USA. • Costs are projected to go down over time, reaching US$0.06–0.11/KWh for attractive sites in 2050

  20. Thermionic conversion reactor technology assessment. Final report

    International Nuclear Information System (INIS)

    1984-02-01

    The in-core thermionic space nuclear power supply may be the only identified reactor-power concept that can meet the SP-100 size functional requirements with demonstrated state-of-the-art reactor system and space-qualified power system component temperatures. The SP-100 configuration limits provide a net 40 m 2 of primary non-deployed radiator area. If a reasonable 7-year degradation allowance of 15% to 20% is provided then the beginning of life (BOL) net power output requirement is about 120 kWe. Consequently, the SP-100 power system must produce a P/A of 2.7 kWe/m 2 . This non-deployed radiator area power density performance can only be reasonably achieved by the thermionic in-core convertr system, the potassium Rankine turbine system and the Stirling engine system. The purpose of this study is to examine past and current tests and data, and to assess the potential for successful development of suitable fueled-thermionic converters that will meet SP-100 and growth requirements. The basis for the assessment will be provided and the recommended key developments plan set forth

  1. Polymorphisms in Plasmodium vivax Circumsporozoite Protein (CSP) Influence Parasite Burden and Cytokine Balance in a Pre-Amazon Endemic Area from Brazil

    Science.gov (United States)

    Ribeiro, Bruno de Paulo; Cassiano, Gustavo Capatti; de Souza, Rodrigo Medeiros; Cysne, Dalila Nunes; Grisotto, Marcos Augusto Grigolin; de Azevedo dos Santos, Ana Paula Silva; Marinho, Cláudio Romero Farias; Machado, Ricardo Luiz Dantas; Nascimento, Flávia Raquel Fernandes

    2016-01-01

    Mechanisms involved in severe P. vivax malaria remain unclear. Parasite polymorphisms, parasite load and host cytokine profile may influence the course of infection. In this study, we investigated the influence of circumsporozoite protein (CSP) polymorphisms on parasite load and cytokine profile in patients with vivax malaria. A cross-sectional study was carried out in three cities: São Luís, Cedral and Buriticupu, Maranhão state, Brazil, areas of high prevalence of P. vivax. Interleukin (IL)-2, IL-4, IL-10, IL-6, IL-17, tumor necrosis factor alpha (TNF-α, interferon gamma (IFN-γ and transforming growth factor beta (TGF-β were quantified in blood plasma of patients and in supernatants from peripheral blood mononuclear cell (PBMC) cultures. Furthermore, the levels of cytokines and parasite load were correlated with VK210, VK247 and P. vivax-like CSP variants. Patients infected with P. vivax showed increased IL-10 and IL-6 levels, which correlated with the parasite load, however, in multiple comparisons, only IL-10 kept this association. A regulatory cytokine profile prevailed in plasma, while an inflammatory profile prevailed in PBMC culture supernatants and these patterns were related to CSP polymorphisms. VK247 infected patients showed higher parasitaemia and IL-6 concentrations, which were not associated to IL-10 anti-inflammatory effect. By contrast, in VK210 patients, these two cytokines showed a strong positive correlation and the parasite load was lower. Patients with the VK210 variant showed a regulatory cytokine profile in plasma, while those infected with the VK247 variant have a predominantly inflammatory cytokine profile and higher parasite loads, which altogether may result in more complications in infection. In conclusion, we propose that CSP polymorphisms is associated to the increase of non-regulated inflammatory immune responses, which in turn may be associated with the outcome of infection. PMID:26943639

  2. MAPLE-X10 reactor safety assessment

    International Nuclear Information System (INIS)

    Cotnam, K.D.; Lounsbury, R.I.; Gillespie, G.E.

    1990-01-01

    This paper reports on the safety assessment of the 10 MW MAPLE-X10 reactor which has involved a substantial component of PSA analysis to supplement deterministic analysis. Initiating events are identified through the use of a master logic diagram. The events are then examined through event sequence diagrams, at the concept design stage, followed by a set of reliability analyses that are coordinated with the event sequence diagrams. Improvements identified through the reliability analyses are incorporated into the design to ensure that safety objectives are attained

  3. Assessment of the technical viability of reactor options for plutonium disposition

    International Nuclear Information System (INIS)

    Primm, R.T. III.

    1996-01-01

    Various reactor concepts for the disposition of surplus Pu have been proposed by reactor vendors; not all have attained the same level of technical viability. Studies were performed to differentiate between reactor concepts by devising a quantitative index for technical viability. For a quantitative assessment, three issues required resolution: the definition of a technical maturity scale, the treatment of ''subjective'' factors which cannot be easily represented in a quantitative format, and the protocol for producing a single technical viability figure-of-merit for each alternative. Alternatives involving the use of foreign facilities were found to be the most technically viable

  4. The selection of probabilistic safety assessment techniques for non-reactor nuclear facilities

    International Nuclear Information System (INIS)

    Vail, J.

    1992-01-01

    Historically, the probabilistic safety assessment (PSA) methodology of choice is the well known event tree/fault tree inductive technique. For reactor facilities is has stood the test of time. Some non-reactor nuclear facilities have found inductive methodologies difficult to apply. The stand-alone fault tree deductive technique has been used effectively to analyze risk in nuclear chemical processing facilities and waste handling facilities. The selection between the two choices suggest benefits from use of the deductive method for non-reactor facilities

  5. Modelling of a cross flow evaporator for CSP application

    DEFF Research Database (Denmark)

    Sørensen, Kim; Franco, Alessandro; Pelagotti, Leonardo

    2016-01-01

    ) applications. Heat transfer and pressure drop prediction methods are an important tool for design and modelling of diabatic, two-phase, shell-side flow over a horizontal plain tubes bundle for a vertical up-flow evaporator. With the objective of developing a model for a specific type of cross flow evaporator...... the available correlations for the definition of two-phase flow heat transfer, void fraction and pressure drop in connection with the operation of steam generators, focuses attention on a comparison of the results obtained using several different models resulting by different combination of correlations......Heat exchangers consisting of bundles of horizontal plain tubes with boiling on the shell side are widely used in industrial and energy systems applications. A recent particular specific interest for the use of this special heat exchanger is in connection with Concentrated Solar Power (CSP...

  6. Application of probabilistic safety assessment to research reactors

    International Nuclear Information System (INIS)

    1989-07-01

    This document has been prepared to assist in the performance of a research reactor probabilistic safety assessment (PSA). It offers examples of experience gained by a number of Member States in carrying out PSA for research reactors. These examples are illustrative of the types of approach adopted, the problems that arise and the judgements entered into when conducting a PSA. The illustrative examples of experiences gained are discussed in a series of thirteen chapters which address some of the issues that arise in a PSA. The examples are not exhaustive and offer evidence of how other analyses have approached the task of preparing a PSA, for their particular plant. The principles should be capable of being utilised and the various issues which are discussed should be translated into the needs of the analyst. Each PSA will make its own demands on the analyst depending on the reactor and so the illustrations must only be used as guidance and not adopted as published, without critical appreciation. Refs, figs and tabs

  7. Technical assessment: An independent safety assessment of Department of Energy nuclear reactor facilities

    International Nuclear Information System (INIS)

    Brodsky, R.S.

    1981-02-01

    Inherent in the design of DOE reactors under review are many features which provide significant protection against the likelihood of TMI-type accidents. In addition, other features in the design or operating characteristics would tend to limit or reduce the consequences of the accident. Some of these features were discussed earlier in this report. However, some of the events included within the TMI accident sequence contain technical implications for the DOE reactors. These implications were reviewed by this Assessment Team, and the results of this review are reported in this and the following sections of this report. It is also important to reemphasize that as a result of this review, no major TMI-related safety issues have been identified that would indicate that these DOE reactors cannot be operated in a safe manner. Rather, the findings of this report, by nature, generally reemphasize and support ongoing DOE efforts and identify areas for additional improvements

  8. Generic magnetic fusion reactor cost assessment

    International Nuclear Information System (INIS)

    Sheffield, J.

    1985-01-01

    The Fusion Energy Division of the Oak Ridge National Laboratory discusses ''generic'' magnetic fusion reactors. The author comments on DT burning magnetic fusion reactor models being possibly operational in the 21st century. Representative parameters from D-T reactor studies are given, as well as a shematic diagram of a generic fusion reactor. Values are given for winding pack current density for existing and future superconducting coils. Topics included are the variation of the cost of electricity (COE), the dependence of the COE on the net electric power of the reactor, and COE formula definitions

  9. Safety assessment of the advanced CANDU reactor in postulated LOCA/LOECC events

    International Nuclear Information System (INIS)

    Hazen Hezhi Fan; Zoran Bilanovic

    2005-01-01

    The Advanced CANDU Reactor TM (ACR TM ) retains the proven strengths and features of CANDU reactors, and incorporates innovative new features and state-of-the-art technology. In addition to the enhanced emergency core cooling system, the reserve water system is designed to be available to inject reserve water by gravity into the reactor inlet headers after a postulated loss-of-coolant accident (LOCA). To assist in the ACR design and analysis of beyond the design basis events, simulations are needed to demonstrate the effectiveness of these two independent systems on core cooling, and to assess the consequences of the postulated accident coincident with the impairment of either of the two systems. The current paper is subject to an assessment of a postulated large LOCA coincident with loss of the emergency core cooling (LOECC) system. A postulated LOCA/LOECC has very low probability, in the range usually associated with severe core damage events. However, in the CANDU design, including ACR, the presence of moderator water surrounding the fuel channels acts as an effective heat sink, together with other safety features, to prevents severe core damage following a postulated LOCA/LOECC. Therefore, it is possible to analyse LOCA/LOECC using the same deterministic tools that are used for analysis of events with much higher frequencies, in the design basis event range. The assessment is conducted based on the current ACR-700 design. However, the analysis methodology, scope, computer tools, and the results in principle, are applicable to larger ACR designs. This assessment includes system (circuit), fuel channel, and fuel analyses. Some assessment results are needed in subsequent moderator analysis and containment analysis. In the assessment, several simulations were performed to analyse the full circuit and individual fuel channel transient behaviours, as well as the fission product release behaviour. The assessment has captured the key responses of the reactor heat

  10. Assessment of the Technical Maturity of Generation IV Concepts for Test or Demonstration Reactor Applications, Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-10-01

    The United States Department of Energy (DOE) commissioned a study the suitability of different advanced reactor concepts to support materials irradiations (i.e. a test reactor) or to demonstrate an advanced power plant/fuel cycle concept (demonstration reactor). As part of the study, an assessment of the technical maturity of the individual concepts was undertaken to see which, if any, can support near-term deployment. A Working Group composed of the authors of this document performed the maturity assessment using the Technical Readiness Levels as defined in DOE’s Technology Readiness Guide . One representative design was selected for assessment from of each of the six Generation-IV reactor types: gas-cooled fast reactor (GFR), lead-cooled fast reactor (LFR), molten salt reactor (MSR), supercritical water-cooled reactor (SCWR), sodium-cooled fast reactor (SFR), and very high temperature reactor (VHTR). Background information was obtained from previous detailed evaluations such as the Generation-IV Roadmap but other technical references were also used including consultations with concept proponents and subject matter experts. Outside of Generation IV activity in which the US is a party, non-U.S. experience or data sources were generally not factored into the evaluations as one cannot assume that this data is easily available or of sufficient quality to be used for licensing a US facility. The Working Group established the scope of the assessment (which systems and subsystems needed to be considered), adapted a specific technology readiness scale, and scored each system through discussions designed to achieve internal consistency across concepts. In general, the Working Group sought to determine which of the reactor options have sufficient maturity to serve either the test or demonstration reactor missions.

  11. Problems in the assessment of inherent safety characteristics of nuclear reactors

    International Nuclear Information System (INIS)

    Garribba, S.F.; Vivante, C.

    1988-01-01

    A number of proposals are being made for an increased RD and D effort on advanced nuclear power reactors that would display outstanding safety performance. A common characteristic of the different reactor concepts would be their limited reliance upon active engineered systems under major accident conditions. However, when submitted to a more close scrutiny reactor concept options may reveal diverging safety behaviors and also development opportunities. In this respect, three issues are explored in this paper. A first question is the meaning of non-active, i.e. inherent and passive safety features. Next, is the ranking of advanced and new reactor concepts from the viewpoint of inherent and passive safety. Multiple correspondence analysis may provide a simple tool, whose use is shown for the case of HTR-500, AP600 and PRISM. Conversely, probabilistic risk assessment would allow quantitative comparisons, although lack of information and data is an obstacle. Finally, is demonstration of safety performances as a step toward market deployment of the new reactor systems

  12. 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

  13. Electro-Kinetic Pumping with Slip Irreversibility in Heat Exchange of CSP-Powered Bio-Digester Assemblies

    OpenAIRE

    Ogedengbe, Emmanuel; Rosen, Marc

    2012-01-01

    Parametric studies of the effects of slip irreversibility in concentrating solar power (CSP)-powered bio-digester assemblies are investigated. Complexities regarding the identification of the appropriate electro-kinetic phenomena for certain electrolyte phases are reviewed. The application of exergy analysis to the design of energy conversion devices, like solar thermal collectors, for the required heat of formation in a downdraft waste food bio-digester, is discussed. Thermal management in t...

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  15. High performance and thermally stable tandem solar selective absorber coating for concentrated solar thermal power (CSP) application

    Science.gov (United States)

    Prasad, M. Shiva; Kumar, K. K. Phani; Atchuta, S. R.; Sobha, B.; Sakthivel, S.

    2018-05-01

    A novel tandem absorber system (Mn-Cu-Co-Ox-ZrO2/SiO2) developed on an austenitic stainless steel (SS-304) substrate to show an excellent optical performance (αsol: 0.96; ɛ: 0.23@500 °C). In order to achieve this durable tandem, we experimented with two antireflective layers such as ZrO2-SiO2 and nano SiO2 layer on top of Mn-Cu-Co-Ox-ZrO2 layer. We optimized the thickness of antireflective layers to get good tandem system in terms of solar absorptance and emittance. Field emission scanning electron microscopy (FESEM), UV-Vis-NIR and Fourier transform infrared spectroscopy (FTIR) were used to characterize the developed coatings. Finally, the Mn-Cu-Co-Ox-ZrO2/SiO2 exhibits high temperature resistance up to 800 °C, thus allow an increase in the operating temperature of CSP which may lead to high efficiency. We successfully developed a high temperature resistant tandem layer with easy manufacturability at low cost which is an attractive candidate for concentrated solar power generation (CSP).

  16. Burning plasma simulation and environmental assessment of tokamak, spherical tokamak and helical reactors

    International Nuclear Information System (INIS)

    Yamazaki, K.; Uemura, S.; Oishi, T.; Arimoto, H.; Shoji, T.; Garcia, J.

    2009-01-01

    Reference 1-GWe DT reactors (tokamak TR-1, spherical tokamak ST-1 and helical HR-1 reactors) are designed using physics, engineering and cost (PEC) code, and their plasma behaviours with internal transport barrier operations are analysed using toroidal transport analysis linkage (TOTAL) code, which clarifies the requirement of deep penetration of pellet fuelling to realize steady-state advanced burning operation. In addition, economical and environmental assessments were performed using extended PEC code, which shows the advantage of high beta tokamak reactors in the cost of electricity (COE) and the advantage of compact spherical tokamak in life-cycle CO 2 emission reduction. Comparing with other electric power generation systems, the COE of the fusion reactor is higher than that of the fission reactor, but on the same level as the oil thermal power system. CO 2 reduction can be achieved in fusion reactors the same as in the fission reactor. The energy payback ratio of the high-beta tokamak reactor TR-1 could be higher than that of other systems including the fission reactor.

  17. Assessment of two small-sized innovative nuclear reactors for electricity generation in Brazil using INPRO methodology

    International Nuclear Information System (INIS)

    Goncalves Filho, Orlando Joao Agostinho; Sefidvash, Farhang

    2009-01-01

    This paper presents the main results of the assessment study of two small-sized innovative reactors for electricity generation in Brazil using the methodology developed under the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO), co-ordinated by the International Atomic Energy Agency (IAEA). INPRO was initiated in 2001 and has the main objective of helping to ensure that nuclear energy is available to contribute in a sustainable manner to the energy needs of the 21st century. Brazil joined the INPRO project since its beginning and in 2005 submitted a proposal for the assessment using INPRO methodology of two small-sized reactors (IRIS - International Reactor Innovative and Secure, and FBNR - Fixed Bed Nuclear Reactor) as potential components of an innovative nuclear energy system (INS) completed by a conventional open nuclear fuel cycle based on enriched uranium. The scope of this assessment study was restricted to the reactor component of the INS and to the methodology areas of economics and safety for IRIS, and proliferation resistance and safety for FBNR. The results indicate that both IRIS and FBNR innovative designs comply mostly with the basic principles of the areas assessed and have potential to comply with the remaining ones. (author)

  18. Thermo-Economic Assessment of Advanced,High-Temperature CANDU Reactors

    International Nuclear Information System (INIS)

    Spinks, Norman J.; Pontikakis, Nikos; Duffey, Romney B.

    2002-01-01

    Research underway on the advanced CANDU examines new, innovative, reactor concepts with the aim of significant cost reduction and resource sustainability through improved thermodynamic efficiency and plant simplification. The so-called CANDU-X concept retains the key elements of the current CANDU designs, including heavy-water moderator that provides a passive heat sink and horizontal pressure tubes. Improvement in thermodynamic efficiency is sought via substantial increases in both pressure and temperature of the reactor coolant. Following on from the new Next Generation (NG) CANDU, which is ready for markets in 2005 and beyond, the reactor coolant is chosen to be light water but at supercritical operating conditions. Two different temperature regimes are being studied, Mark 1 and Mark 2, based respectively on continued use of zirconium or on stainless-steel-based fuel cladding. Three distinct cycle options have been proposed for Mark 1: the High-Pressure Steam Generator (HPSG) cycle, the Dual cycle, and the Direct cycle. For Mark 2, the focus is on simplification via a Direct cycle. This paper presents comparative thermo-economic assessments of the CANDU-X cycle options, with the ultimate goal of ascertaining which particular cycle option is the best overall in terms of thermodynamics and economics. A similar assessment was already performed for the NG CANDU. The economic analyses entail obtaining cost estimates of major plant components, such as heat exchangers, turbines and pumps. (authors)

  19. Isolation and Characterization of Vaccine Candidate Genes Including CSP and MSP1 in Plasmodium yoelii.

    Science.gov (United States)

    Kim, Seon-Hee; Bae, Young-An; Seoh, Ju-Young; Yang, Hyun-Jong

    2017-06-01

    Malaria is an infectious disease affecting humans, which is transmitted by the bite of Anopheles mosquitoes harboring sporozoites of parasitic protozoans belonging to the genus Plasmodium . Despite past achievements to control the protozoan disease, malaria still remains a significant health threat up to now. In this study, we cloned and characterized the full-unit Plasmodium yoelii genes encoding merozoite surface protein 1 (MSP1), circumsporozoite protein (CSP), and Duffy-binding protein (DBP), each of which can be applied for investigations to obtain potent protective vaccines in the rodent malaria model, due to their specific expression patterns during the parasite life cycle. Recombinant fragments corresponding to the middle and C-terminal regions of PyMSP1 and PyCSP, respectively, displayed strong reactivity against P. yoelii -infected mice sera. Specific native antigens invoking strong humoral immune response during the primary and secondary infections of P. yoelii were also abundantly detected in experimental ICR mice. The low or negligible parasitemia observed in the secondary infected mice was likely to result from the neutralizing action of the protective antibodies. Identification of these antigenic proteins might provide the necessary information and means to characterize additional vaccine candidate antigens, selected solely on their ability to produce the protective antibodies.

  20. A review of Andasol 3 and perspective for parabolic trough CSP plants in South Africa

    Science.gov (United States)

    Dinter, Frank; Möller, Lucas

    2016-05-01

    Andasol 3 is a 50 MW parabolic trough concentrating solar power plant with thermal energy storage in Andalusia, southern Spain. Having started operating in 2011 as one of the first plants of its kind in Spain it has been followed by more than 50 in the country since. For the reason that CSP plants with storage have the potential to compete against fossil fuel fired plants much better than any other renewable energy source a long-term review of such a plant operating on a commercial scale is needed. With data at hand documenting Andasol 3's operation over the course of one year between July 2013 and June 2014 we intend to provide such a review. We calculated the plants overall efficiency, its capacity factor, the gross energy generation as well as auxiliary powers on a monthly basis to reflect upon its overall performance. It was also looked at the benefits caused by the thermal energy storage and especially how steadily and reliably the plant was able to operate. With basic background information about physical, geographical and meteorological aspects influencing the solar resource, its variation and a CSP plant's performance a qualitative estimation for a parabolic trough plant located in South Africa was made.

  1. Adapting a reactor safety assessment system for specific plants

    International Nuclear Information System (INIS)

    Ballard, T.L.; Cordes, G.A.

    1991-01-01

    The Reactor Safety Assessment System (RSAS) is an expert system being developed by the Idaho National Engineering Laboratory, the University of Maryland (UofM) and US Nuclear Regulatory Commission (NRC) for use in the NRC Operations center. RSAS is designed to help the Reactor Safety Team monitor and project core status during an emergency at a licensed nuclear power plant. Analysis uses a hierarchical plant model based on equipment availability and automatically input parametric plant information. There are 3 families of designs of pressurized water reactors and 75 plants using modified versions of the basic design. In order to make an RSAS model for each power plant, a generic model for a given plant type is used with differences being specified by plant specific files. Graphical displays of this knowledge are flexible enough to handle any plant configuration. A variety of tools have been implemented to make it easy to modify a design to fit a given plant while minimizing chance for error. 3 refs., 4 figs

  2. Assessment of very high temperature reactors in process applications

    International Nuclear Information System (INIS)

    Jones, J.E. Jr.; Spiewak, I.; Gambill, W.R.

    1976-01-01

    In April 1974, the United States Energy Research and Development Administration (ERDA) authorized General Atomic Company, General Electric Company, and Westinghouse Astronuclear Laboratory to assess the available technology for producing process heat utilizing a very high temperature nuclear reactor (VHTR). The VHTR is defined as a gas-cooled graphite-moderated reactor. Oak Ridge National Laboratory has been given a lead role in evaluating the VHTR reactor studies and potential applications of the VHTR. Process temperatures up to the 760 to 871 0 C range appear to be achievable with near-term technology. The major development considerations are high temperature materials, the safety questions (especially regarding the need for an intermediate heat exchanger) and the process heat exchanger. The potential advantages of the VHTR over competing fossil energy sources are conservation of fossil fuels and reduced atmospheric impacts. Costs are developed for nuclear process heat supplied from a 3000-MW(th) VHTR. The range of cost in process applications is competitive with current fossil fuel alternatives

  3. Assessing reactor physics codes capabilities to simulate fast reactors on the example of the BN-600 benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Ivanov, Vladimir [Scientific and Engineering Centre for Nuclear and Radiation Safety (SES NRS), Moscow (Russian Federation); Bousquet, Jeremy [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany)

    2016-11-15

    This work aims to assess the capabilities of reactor physics codes (initially validated for thermal reactors) to simulate fast sodium cooled reactors. The BFS-62-3A critical experiment from the BN-600 Hybrid Core Benchmark Analyses was chosen for the investigation. Monte-Carlo codes (KENO from SCALE and SERPENT 2.1.23) and the deterministic diffusion code DYN3D-MG are applied to calculate the neutronic parameters. It was found that the multiplication factor and reactivity effects calculated by KENO and SERPENT using the ENDF/B-VII.0 continuous energy library are in a good agreement with each other and with the measured benchmark values. Few-groups macroscopic cross sections, required for DYN3D-MG, were prepared in applying different methods implemented in SCALE and SERPENT. The DYN3D-MG results of a simplified benchmark show reasonable agreement with results from Monte-Carlo calculations and measured values. The former results are used to justify DYN3D-MG implementation for sodium cooled fast reactors coupled deterministic analysis.

  4. Reactor safeguards system assessment and design. Volume I

    International Nuclear Information System (INIS)

    Varnado, G.B.; Ericson, D.M. Jr.; Daniel, S.L.; Bennett, H.A.; Hulme, B.L.

    1978-06-01

    This report describes the development and application of a methodology for evaluating the effectiveness of nuclear power reactor safeguards systems. Analytic techniques are used to identify the sabotage acts which could lead to release of radioactive material from a nuclear power plant, to determine the areas of a plant which must be protected to assure that significant release does not occur, to model the physical plant layout, and to evaluate the effectiveness of various safeguards systems. The methodology was used to identify those aspects of reactor safeguards systems which have the greatest effect on overall system performance and which, therefore, should be emphasized in the licensing process. With further refinements, the methodology can be used by the licensing reviewer to aid in assessing proposed or existing safeguards systems

  5. Response of Vibrio cholerae to Low-Temperature Shifts: CspV Regulation of Type VI Secretion, Biofilm Formation, and Association with Zooplankton.

    Science.gov (United States)

    Townsley, Loni; Sison Mangus, Marilou P; Mehic, Sanjin; Yildiz, Fitnat H

    2016-07-15

    The ability to sense and adapt to temperature fluctuation is critical to the aquatic survival, transmission, and infectivity of Vibrio cholerae, the causative agent of the disease cholera. Little information is available on the physiological changes that occur when V. cholerae experiences temperature shifts. The genome-wide transcriptional profile of V. cholerae upon a shift in human body temperature (37°C) to lower temperatures, 15°C and 25°C, which mimic those found in the aquatic environment, was determined. Differentially expressed genes included those involved in the cold shock response, biofilm formation, type VI secretion, and virulence. Analysis of a mutant lacking the cold shock gene cspV, which was upregulated >50-fold upon a low-temperature shift, revealed that it regulates genes involved in biofilm formation and type VI secretion. CspV controls biofilm formation through modulation of the second messenger cyclic diguanylate and regulates type VI-mediated interspecies killing in a temperature-dependent manner. Furthermore, a strain lacking cspV had significant defects for attachment and type VI-mediated killing on the surface of the aquatic crustacean Daphnia magna Collectively, these studies reveal that cspV is a major regulator of the temperature downshift response and plays an important role in controlling cellular processes crucial to the infectious cycle of V. cholerae Little is known about how human pathogens respond and adapt to ever-changing parameters of natural habitats outside the human host and how environmental adaptation alters dissemination. Vibrio cholerae, the causative agent of the severe diarrheal disease cholera, experiences fluctuations in temperature in its natural aquatic habitats and during the infection process. Furthermore, temperature is a critical environmental signal governing the occurrence of V. cholerae and cholera outbreaks. In this study, we showed that V. cholerae reprograms its transcriptome in response to

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

    International Nuclear Information System (INIS)

    1977-06-01

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

  7. Feasibility and Safety Assessment for Advanced Reactor Concepts Using Vented Fuel

    International Nuclear Information System (INIS)

    Klein, Andrew; Lenhof, Renae; Deason, Wesley; Harter, Jackson

    2015-01-01

    Recent interest in fast reactor technology has led to renewed analysis of past reactor concepts such as Gas Fast Reactors and Sodium Fast Reactors. In an effort to make these reactors more economic, the fuel is required to stay in the reactor for extended periods of time; the longer the fuel stays within the core, the more fertile material is converted into usable fissile material. However, as burnup of the fuel-rod increases, so does the internal pressure buildup due to gaseous fission products. In order to reach the 30 year lifetime requirements of some reactor designs, the fuel pins must have a vented-type design to allow the buildup of fission products to escape. The present work aims to progress the understanding of the feasibility and safety issues related to gas reactors that incorporate vented fuel. The work was separated into three different work-scopes: 1. Quantitatively determine fission gas release from uranium carbide in a representative helium cooled fast reactor; 2. Model the fission gas behavior, transport, and collection in a Fission Product Vent System; and, 3. Perform a safety analysis of the Fission Product Vent System. Each task relied on results from the previous task, culminating in a limited scope Probabilistic Risk Assessment (PRA) of the Fission Product Vent System. Within each task, many key parameters lack the fidelity needed for comprehensive or accurate analysis. In the process of completing each task, the data or methods that were lacking were identified and compiled in a Gap Analysis included at the end of the report.

  8. Feasibility and Safety Assessment for Advanced Reactor Concepts Using Vented Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Klein, Andrew [Oregon State Univ., Corvallis, OR (United States). Nuclear Engineering and Radiation Health Physics; Matthews, Topher [Oregon State Univ., Corvallis, OR (United States); Lenhof, Renae [Oregon State Univ., Corvallis, OR (United States); Deason, Wesley [Oregon State Univ., Corvallis, OR (United States); Harter, Jackson [Oregon State Univ., Corvallis, OR (United States)

    2015-01-16

    Recent interest in fast reactor technology has led to renewed analysis of past reactor concepts such as Gas Fast Reactors and Sodium Fast Reactors. In an effort to make these reactors more economic, the fuel is required to stay in the reactor for extended periods of time; the longer the fuel stays within the core, the more fertile material is converted into usable fissile material. However, as burnup of the fuel-rod increases, so does the internal pressure buildup due to gaseous fission products. In order to reach the 30 year lifetime requirements of some reactor designs, the fuel pins must have a vented-type design to allow the buildup of fission products to escape. The present work aims to progress the understanding of the feasibility and safety issues related to gas reactors that incorporate vented fuel. The work was separated into three different work-scopes: 1. Quantitatively determine fission gas release from uranium carbide in a representative helium cooled fast reactor; 2. Model the fission gas behavior, transport, and collection in a Fission Product Vent System; and, 3. Perform a safety analysis of the Fission Product Vent System. Each task relied on results from the previous task, culminating in a limited scope Probabilistic Risk Assessment (PRA) of the Fission Product Vent System. Within each task, many key parameters lack the fidelity needed for comprehensive or accurate analysis. In the process of completing each task, the data or methods that were lacking were identified and compiled in a Gap Analysis included at the end of the report.

  9. Measurement of ex vivo ELISpot interferon-gamma recall responses to Plasmodium falciparum AMA1 and CSP in Ghanaian adults with natural exposure to malaria.

    Science.gov (United States)

    Ganeshan, Harini; Kusi, Kwadwo A; Anum, Dorothy; Hollingdale, Michael R; Peters, Bjoern; Kim, Yohan; Tetteh, John K A; Ofori, Michael F; Gyan, Ben A; Koram, Kwadwo A; Huang, Jun; Belmonte, Maria; Banania, Jo Glenna; Dodoo, Daniel; Villasante, Eileen; Sedegah, Martha

    2016-02-01

    Malaria eradication requires a concerted approach involving all available control tools, and an effective vaccine would complement these efforts. An effective malaria vaccine should be able to induce protective immune responses in a genetically diverse population. Identification of immunodominant T cell epitopes will assist in determining if candidate vaccines will be immunogenic in malaria-endemic areas. This study therefore investigated whether class I-restricted T cell epitopes of two leading malaria vaccine antigens, Plasmodium falciparum circumsporozoite protein (CSP) and apical membrane antigen-1 (AMA1), could recall T cell interferon-γ responses from naturally exposed subjects using ex vivo ELISpot assays. Thirty-five subjects aged between 24 and 43 years were recruited from a malaria-endemic urban community of Ghana in 2011, and their peripheral blood mononuclear cells (PBMCs) were tested in ELISpot IFN-γ assays against overlapping 15mer peptide pools spanning the entire CSP and AMA1 antigens, and 9-10mer peptide epitope mixtures that included previously identified and/or predicted human leukocyte antigen (HLA) class 1-restricted epitopes from same two antigens. For CSP, 26 % of subjects responded to at least one of the nine 15mer peptide pools whilst 17 % responded to at least one of the five 9-10mer HLA-restricted epitope mixtures. For AMA1, 63 % of subjects responded to at least one of the 12 AMA1 15mer peptide pools and 51 % responded to at least one of the six 9-10mer HLA-restricted epitope mixtures. Following analysis of data from the two sets of peptide pools, along with bioinformatics predictions of class I-restricted epitopes and the HLA supertypes expressed by a subset of study subjects, peptide pools that may contain epitopes recognized by multiple HLA supertypes were identified. Collectively, these results suggest that natural transmission elicits ELISpot IFN-γ activities to class 1-restricted epitopes that are largely HLA-promiscuous. These

  10. An independent safety assessment of Department of Energy nuclear reactor facilities: Safety overview and management function

    International Nuclear Information System (INIS)

    Booth, M.; Brodsky, R.S.; Frankhouser, W.L.

    1981-02-01

    The Under Secretary of Energy established the Nuclear Facilities Personnel Qualification and Training (NFPQT) Committee in October, 1979, in the aftermath of the Three Mile Island (TMI) nuclear accident, to assess the adequacy of training of personnel at DOE nuclear facilities. Subsequently, in February, 1980, the charge to this Committee was modified to assess all implications of the Kemeny Commission report on TMI with regard to DOE nuclear reactors, excluding those in the Division of Naval Reactors. The modified charge was also limited, for the time being, to reactor facilities instead of all nuclear facilities. This report describes the portion of the revised assessment activities that was assigned to the Assessment Support Team

  11. Development of a numerical tool for safety assessment and emergency management of experimental reactors

    International Nuclear Information System (INIS)

    Maas, L.; Beuter, A.; Seropian, C.

    2010-01-01

    The Institute of Radiological Protection and Nuclear Safety (IRSN) acts as technical support to French public authorities. Among its duties, one important item is to provide help for emergency situations management in case of an accident occurring in a French nuclear facility. In this framework, IRSN develops and applies numerical tools dealing with containment management issues. Up to now IRSN has not got any specific tool for experimental reactors. Accordingly, it has been then decided to extend the ASTEC code, devoted to severe accident scenarios for Pressurized Water Reactors, to this kind of reactors. This lumped-parameter code, co-developed by IRSN and GRS (Germany), covers the entire phenomenology from the initiating event up to fission products release outside the reactor containment, except for the steam explosion and the mechanical integrity of the containment. A first application to experimental reactors was carried out to assess the High Flux Reactor (HFR) operator's improvement proposal concerning the containment management during accidental situations. This reactor, located in Grenoble (France), is composed of a double wall containment with a pressurized containment annulus preventing any direct leakage into the environment. Until now, in case of severe accidents (mainly core melting in pool, explosive reactivity accident called BORAX), the HFR emergency management consisted in isolating the containment building in the early stage of the accident, to prevent any radioactive products release to the environment. The operator decided to improve this containment management during accidental situations by using an air filtering venting system able to maintain a slight sub-atmospheric pressure in the reactor building. The operator's demonstration of the efficiency of this new system is mainly based on containment pressure evaluations during accidental transients. IRSN assessed these calculations through ASTEC calculations. Finally, a global agreement was

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

    International Nuclear Information System (INIS)

    Tinnes, S.P.; Cramer, D.S.; Logan, V.E.; Topp, S.V.; Smith, J.A.; Brandyberry, M.D.

    1990-01-01

    A full-scope probabilistic risk assessment (PRA) is being performed for the Savannah River site (SRS) production reactors. The Level 1 PRA for the K Reactor has been completed and includes the assessment of reactor systems response to accidents and estimates of the severe core melt frequency (SCMF). The internal events spectrum includes those events related directly to plant systems and safety functions for which transients or failures may initiate an accident. The SRS PRA has three principal objectives: improved understanding of SRS reactor safety issues through discovery and understanding of the mechanisms involved. Improved risk management capability through tools for assessing the safety impact of both current standard operations and proposed revisions. A quantitative measure of the risks posed by SRS reactor operation to employees and the general public, to allow comparison with declared goals and other societal risks

  13. Assessment of residual life of fast breeder test reactor

    International Nuclear Information System (INIS)

    Srinivasan, G.

    2016-01-01

    The Fast Breeder Test Reactor (FBTR) is a loop type sodium cooled fast reactor and has been in operation since 1985. As a part of regulatory requirement for relicensing, residual life assessment had to be carried out. The systems are made of SS 316, and designed for creep and fatigue. The design life for creep is 100,000 h at 550°C. The design fatigue cycle for operation from shutdown to full power varies from component to component. In general, most of the components are designed for 2000 cycles. The reactor has operated mostly below the design temperatures. It is seen that enough creep-fatigue life is available for the non-replaceable, permanent components. The residual life was found to be governed by the residual ductility of the Grid Plate supporting the core after neutron irradiation. Fast flux measurements were carried out at the grid plate location. Samples were irradiated and tensile tested. Results indicate the allowable dpa for a 10% residual ductility criterion as 4.37. This gave a residual life of ~ 6 Effective Full Power Years for the reactor as of Feb 2012. Measures to reduce the neutron dose on the grid plate are being taken. (author)

  14. Probability safety assessment of LOOP accident to molten salt reactor

    International Nuclear Information System (INIS)

    Mei Mudan; Shao Shiwei; Yu Zhizhen; Chen Kun; Zuo Jiaxu

    2013-01-01

    Background: Loss of offsite power (LOOP) is a possible accident to any type of reactor, and this accident can reflect the main idea of reactor safety design. Therefore, it is very important to conduct a study on probabilistic safety assessment (PSA) of the molten salt reactor that is under LOOP circumstance. Purpose: The aim is to calculate the release frequency of molten salt radioactive material to the core caused by LOOP, and find out the biggest contributor to causing the radioactive release frequency. Methods: We carried out the PSA analysis of the LOOP using the PSA process risk spectrum, and assumed that the primary circuit had no valve and equipment reliability data based on the existing mature power plant equipment reliability data. Results: Through the PSA analysis, we got the accident sequences of the release of radioactive material to the core caused by LOOP and its frequency. The results show that the release frequency of molten salt radioactive material to the core caused by LOOP is about 2×10 -11 /(reactor ·year), which is far below that of the AP1000 LOOP. In addition, through the quantitative analysis, we obtained the point estimation and interval estimation of uncertainty analysis, and found that the biggest contributor to cause the release frequency of radioactive material to the core is the reactor cavity cooling function failure. Conclusion: This study provides effective help for the design and improvement of the following molten salt reactor system. (authors)

  15. Recent advances in C(sp3–H bond functionalization via metal–carbene insertions

    Directory of Open Access Journals (Sweden)

    Bo Wang

    2016-04-01

    Full Text Available The recent development of intermolecular C–H insertion in the application of C(sp3–H bond functionalizations, especially for light alkanes, is reviewed. The challenging problem of regioselectivity in C–H bond insertions has been tackled by the use of sterically bulky metal catalysts, such as metal porphyrins and silver(I complexes. In some cases, high regioselectivity and enantioselectivity have been achieved in the C–H bond insertion of small alkanes. This review highlights the most recent accomplishments in this field.

  16. Assessment of Power Quality Problems for TRIGA PUSPATI Reactor (RTP)

    International Nuclear Information System (INIS)

    Mohd Fazli Zakaria; Ramachandaramurthy, V.K.

    2016-01-01

    The electrical power systems are exposed to different types of power quality disturbances. Investigation and monitoring of power quality is necessary to maintain accurate operation of sensitive equipment especially for nuclear installations. This paper will discuss the power quality problems observed at the electrical sources of PUSPATI TRIGA Reactor (RTP). Assessment of power quality requires the identification of any anomalous behavior on a power system, which adversely affects the normal operation of electrical or electronic equipment. A power quality assessment involves gathering data resources; analyzing the data (with reference to power quality standards) then, if problems exist, recommendation of mitigation techniques must be considered. Field power quality data is collected by power quality recorder and analyzed with reference to power quality standards. Normally the electrical power is supplied to the RTP via two sources in order to keep a good reliability where each of them is designed to carry the full load. The assessment of power quality during reactor operation was performed for both electrical sources. There were several disturbances such as voltage harmonics and flicker that exceeded the thresholds. (author)

  17. An investigation into multi-dimensional prediction models to estimate the pose error of a quadcopter in a CSP plant setting

    Science.gov (United States)

    Lock, Jacobus C.; Smit, Willie J.; Treurnicht, Johann

    2016-05-01

    The Solar Thermal Energy Research Group (STERG) is investigating ways to make heliostats cheaper to reduce the total cost of a concentrating solar power (CSP) plant. One avenue of research is to use unmanned aerial vehicles (UAVs) to automate and assist with the heliostat calibration process. To do this, the pose estimation error of each UAV must be determined and integrated into a calibration procedure. A computer vision (CV) system is used to measure the pose of a quadcopter UAV. However, this CV system contains considerable measurement errors. Since this is a high-dimensional problem, a sophisticated prediction model must be used to estimate the measurement error of the CV system for any given pose measurement vector. This paper attempts to train and validate such a model with the aim of using it to determine the pose error of a quadcopter in a CSP plant setting.

  18. External event Probabilistic Risk Assessment for the High Flux Isotope Reactor (HFIR)

    International Nuclear Information System (INIS)

    Flanagan, G.F.; Johnson, D.H.; Buttemer, D.; Perla, H.F.; Chien, S.H.

    1989-01-01

    The High Flux Isotope Reactor (HFIR) is a high performance isotope production and research reactor which has been in operation at Oak Ridge National Laboratory (ORNL) since 1965. In late 1986 the reactor was shut down as a result of discovery of unexpected neutron embrittlement of the reactor vessel. In January of 1988 a level 1 Probabilistic Risk Assessment (PRA) (excluding external events) was published as part of the response to the many reviews that followed the shutdown and for use by ORNL to prioritize action items intended to upgrade the safety of the reactor. A conservative estimate of the core damage frequency initiated by internal events for HFIR was 3.11 x 10 -4 . In June 1989 a draft external events initiated PRA was published. The dominant contributions from external events came from seismic, wind, and fires. The overall external event contribution to core damage frequency is about 50% of the internal event initiated contribution and is dominated by seismic events

  19. Benefits of production extension and shifting with thermal storage for a 1MW CSP-ORC plant in Morocco

    Science.gov (United States)

    Bennouna, El Ghali; Mimet, Abdelaziz; Frej, Hicham

    2016-05-01

    The importance of thermal storage for commercial CSP (concentrated Solar Power) plants has now become obvious, this regardless of the solar technology used and the power cycle. The availability of a storage system to a plant operator brings a lot of possibilities for production management, cash flow optimization and grid stabilizing. In particular, and depending on plant location and local grid strategy, thermal storage can contribute, when wisely used, to control production and adapt it to the demand and / or power unbalances and varying prices. Storage systems design, sizing and configuration are proper to each power plant, hence systems that are now widely installed within large commercial solar plants are not necessarily suited for small scale decentralized production, and will not have the same effects. In this paper the benefits of thermal storage are studied for a 1MWe CSP plant with an ORC (Organic Rankine Cycle), this plant has many specific features which call for a detail analysis about the appropriate storage design and optimum operating strategies for decentralized solutions.

  20. Assessment of Loss-of-Coolant Effect on Pressurized Heavy Water Reactors

    International Nuclear Information System (INIS)

    Kim, Won Young; Park, Joo Hwan; Kim, Bong Ghi

    2009-01-01

    A CANDU reactor is a heavy-water-moderated, natural uranium fuelled reactor with a pressure tube. The reactor contains a horizontal cylindrical vessel (calandria) and each pressure tube is isolated from the heavy-water moderator in a calandria. This allows the moderator system to be operated of a high-pressure and of a high-temperature coolant in pressure tube. This causes the pressurized liquid coolant in the channel to void and therefore give rise to a reactivity transient in the event of a break or fault in the coolant circuit. In particular, all CANDU reactors are well known to have a positive void reactivity coefficient and thus this phenomenon may lead to a positive feedback, which can cause a large power pulse. We assess the loss-of-coolant effect by coolant void reactivity versus fuel burnup, four factor parameters for fresh fuel and equilibrium fuel, reactivity change due to the change of coolant density and reactivity change in the case of half- and full-core coolant

  1. Experience in the implementation of quality assurance program and safety culture assessment of research reactor operation and maintenance

    International Nuclear Information System (INIS)

    Syarip; Suryopratomo, K.

    2001-01-01

    The implementation of quality assurance program and safety culture for research reactor operation are of importance to assure its safety status. It comprises an assessment of the quality of both technical and organizational aspects involved in safety. The method for the assessment is based on judging the quality of fulfillment of a number of essential issues for safety i.e. through audit, interview and/or discussions with personnel and management in plant. However, special consideration should be given to the data processing regarding the fuzzy nature of the data i.e. in answering the questionnaire. To accommodate this situation, the SCAP, a computer program based on fuzzy logic for assessing plant safety status, has been developed. As a case study, the experience in the assessment of Kartini research reactor safety status shows that it is strongly related to the implementation of quality assurance program in reactor operation and awareness of reactor operation staffs to safety culture practice. It is also shown that the application of the fuzzy rule in assessing reactor safety status gives a more realistic result than the traditional approach. (author)

  2. A probabilistic consequence assessment for a very high temperature reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Joeun; Kim, Jintae; Jae, Moosung [Hanyang Univ., Seoul (Korea, Republic of). Dept. of Nuclear Engineering

    2017-02-15

    Currently, fossil fuel is globally running out. If current trends continue, crude oil will be depleted in 20 years and natural gas in 40 years. In addition, the use of fossil resource has increased emissions of green gas such as carbon dioxide. Therefore, there has been a strong demand in recent years for producing large amounts of hydrogen as an alternative energy [1]. To generate hydrogen energy, very high temperature more than 900 C is required but this level is not easy to reach. Because a Very High Temperature Reactor (VHTR), one of next generation reactor, is able to make the temperature, it is regarded as a solution of the problem. Also, VHTR has an excellent safety in comparison with existing and other next generation reactors. Especially, a passive system, Reactor Cavity Cooling System (RCCS), is adopted to get rid of radiant heat in case of accidents. To achieve variety requirements of new designed-reactors, however, it needs to develop new methodologies and definitions different with existing method. At the same time, an application of probability safety assessment (PSA) has been proposed to ensure the safety of next generation NPPs. For this, risk-informed designs of structures have to be developed and verified. Particularly, the passive system requires to be evaluated for its reliability. The objective of this study is to improve safety of VIITR by conducting risk profile.

  3. Safety case methodology for decommissioning of research reactors. Assessment of the long term impact of a flooding scenario

    International Nuclear Information System (INIS)

    Vladescu, G.; Banciu, O.

    1999-01-01

    The paper contains the assessment methodology of a Safety Case fuel decommissioning of research reactors, taking into account the international approach principles. The paper also includes the assessment of a flooding scenario for a decommissioned research reactor (stage 1 of decommissioning). The scenario presents the flooding of reactor basement, radionuclide migration through environment and long term radiological impact for public. (authors)

  4. Procedures for conducting probabilistic safety assessment for non-reactor nuclear facilities

    International Nuclear Information System (INIS)

    2002-01-01

    A well performed and adequately documented safety assessment of a nuclear facility will serve as a basis to determine whether the facility complies with the safety objectives, principles and criteria as stipulated by the national regulatory body of the country where the facility is in operation. International experience shows that the practices and methodologies used to perform safety assessments and periodic safety re-assessment for non-reactor nuclear facilities differ significantly from county to country. Most developing countries do not have methods and guidance for safety assessment that are prescribed by the regulatory body. Typically the safety evaluation for the facility is based on a case by case assessment. Whilst conservative deterministic analyses are predominantly used as a licensing basis in many countries, recently probabilistic safety assessment (PSA) techniques have been applied as a useful complementary tool to support safety decision making. The main benefit of PSA is to provide insights into the safety aspects of facility design and operation. PSA points up the potential environmental impacts of postulated accidents, including the dominant risk contributors, and enables safety analysts to compare options for reducing risk. In order to advise on how to apply PSA methodology for the safety assessment of non-reactor nuclear facilities, the IAEA organized several consultants meetings, which led to the preparation of this TECDOC. This document is intended as guidance for the conduct of PSA in non-nuclear facilities. The main emphasis here is on the general procedural steps of a PSA that is specific for a non-reactor nuclear facility, rather than the details of the specific methods. The report is directed at technical staff managing or performing such probabilistic assessments and to promote a standardized framework, terminology and form of documentation for these PSAs. It is understood that the level of detail implied in the tasks presented in this

  5. The future prospect of PV and CSP solar technologies: An expert elicitation survey

    International Nuclear Information System (INIS)

    Bosetti, Valentina; Catenacci, Michela; Fiorese, Giulia; Verdolini, Elena

    2012-01-01

    In this paper we present and discuss the results of an expert elicitation survey on solar technologies. Sixteen leading European experts from the academic world, the private sector and international institutions took part in this expert elicitation survey on Photovoltaic (PV) and Concentrated Solar Power (CSP) technologies. The survey collected probabilistic information on (1) how Research, Development and Demonstration (RD and D) investments will impact the future costs of solar technologies and (2) the potential for solar technology deployment both in OECD and non-OECD countries. Understanding the technological progress and the potential of solar PV and CPS technologies is crucial to draft appropriate energy policies. The results presented in this paper are thus relevant for the policy making process and can be used as better input data in integrated assessment and energy models. - Highlights: ► With constant public support at least one solar technology will become cost-competitive with fossil fuels. ► Demonstration should become a key area of funding. ► Without climate policy (carbon price), by 2030 solar technologies will not be cost-competitive. ► The EU will first achieve a breakthrough in production costs. ► The share of electricity production from solar will never exceed 30%.

  6. Soil slurry reactors for the assessment of contaminant biodegradation

    Science.gov (United States)

    Toscano, G.; Colarieti, M. L.; Greco, G.

    2012-04-01

    Slurry reactors are frequently used in the assessment of feasibility of biodegradation in natural soil systems. The rate of contaminant removal is usually quantified by zero- or first-order kinetics decay constants. The significance of such constants for the evaluation of removal rate in the field could be questioned because the slurry reactor is a water-saturated, well-stirred system without resemblance with an unsaturated fixed bed of soil. Nevertheless, a kinetic study with soil slurry reactors can still be useful by means of only slightly more sophisticated kinetic models than zero-/first-order decay. The use of kinetic models taking into account the role of degrading biomass, even in the absence of reliable experimental methods for its quantification, provides further insight into the effect of nutrient additions. A real acceleration of biodegradation processes is obtained only when the degrading biomass is in the growth condition. The apparent change in contaminant removal course can be useful to diagnose biomass growth without direct biomass measurement. Even though molecular biology techniques are effective to assess the presence of potentially degrading microorganism in a "viable-but-nonculturable" state, the attainment of conditions for growth is still important to the development of enhanced remediation techniques. The methodology is illustrated with reference to data gathered for two test sites, Oslo airport Gardermoen in Norway (continuous contamination by aircraft deicing fluids) and the Trecate site in Italy (aged contamination by crude oil spill). This research is part of SoilCAM project (Soil Contamination, Advanced integrated characterisation and time-lapse Monitoring 2008-2012, EU-FP7).

  7. Reports and operational engineering: An independent safety assessment of Department of Energy nuclear reactor facilities

    International Nuclear Information System (INIS)

    Rochman, A.; Washburn, B.W.

    1981-02-01

    The Nuclear Facilities Personnel Qualification and Training (NFPQT) Committee, established via an October 24, 1979 memorandum from the Department of Energy (DOE) Under Secretary, was instructed to review the ''Kemeny Commission'' recommendations and to identify possible implications for DOE's nuclear facilities. As a result of this review, the Committee recommended that DOE carry out assessments in seven categories. The assessments would address specific topics identified for each category as delineated in the NFPQT ''Guidelines for Assessing the Safe Operation of DOE-Owned Reactors,'' dated May 7, 1980. The Committee recognized that similar assessments had been ongoing in the DOE program and safety overview organizations since the Three Mile Island nuclear accident and it was the Committee's intent to use the results of those ongoing assessments as an input to their evaluations. This information would be supplemented by additional studies consisting of the subject-related documents used at each reactor facility studied, and an on-site review of these reactor facilities by professional personnel within the Department of Energy, its operating contractors and independent consultants. 1 tab

  8. Reliability assessment of Passive Containment Cooling System of an Advanced Reactor using APSRA methodology

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Mukesh, E-mail: mukeshd@barc.gov.in [Reactor Engineering Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Chakravarty, Aranyak [School of Nuclear Studies and Application, Jadavpur University, Kolkata 700032 (India); Nayak, A.K. [Reactor Engineering Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Prasad, Hari; Gopika, V. [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2014-10-15

    Highlights: • The paper deals with the reliability assessment of Passive Containment Cooling System of Advanced Heavy Water Reactor. • Assessment of Passive System ReliAbility (APSRA) methodology is used for reliability assessment. • Performance assessment of the PCCS is initially performed during a postulated design basis LOCA. • The parameters affecting the system performance are then identified and considered for further analysis. • The failure probabilities of the various components are assessed through a classical PSA treatment using generic data. - Abstract: Passive Systems are increasingly playing a prominent role in the advanced nuclear reactor systems and are being utilised in normal operations as well as safety systems of the reactors following an accident. The Passive Containment Cooling System (PCCS) is one of the several passive safety features in an Advanced Reactor (AHWR). In this paper, the APSRA methodology has been employed for reliability evaluation of the PCCS of AHWR. Performance assessment of the PCCS is initially performed during a postulated design basis LOCA using the best-estimate code RELAP5/Mod 3.2. The parameters affecting the system performance are then identified and considered for further analysis. Based on some pre-determined failure criterion, the failure surface for the system is predicted using the best-estimate code taking into account the deviations of the identified parameters from their nominal states as well as the model uncertainties inherent to the best estimate code. Root diagnosis is then carried out to determine the various failure causes, which occurs mainly due to malfunctioning of mechanical components. The failure probabilities of the various components are assessed through a classical PSA treatment using generic data. The reliability of the PCCS is then evaluated from the probability of availability of these components.

  9. Reliability assessment of Passive Containment Cooling System of an Advanced Reactor using APSRA methodology

    International Nuclear Information System (INIS)

    Kumar, Mukesh; Chakravarty, Aranyak; Nayak, A.K.; Prasad, Hari; Gopika, V.

    2014-01-01

    Highlights: • The paper deals with the reliability assessment of Passive Containment Cooling System of Advanced Heavy Water Reactor. • Assessment of Passive System ReliAbility (APSRA) methodology is used for reliability assessment. • Performance assessment of the PCCS is initially performed during a postulated design basis LOCA. • The parameters affecting the system performance are then identified and considered for further analysis. • The failure probabilities of the various components are assessed through a classical PSA treatment using generic data. - Abstract: Passive Systems are increasingly playing a prominent role in the advanced nuclear reactor systems and are being utilised in normal operations as well as safety systems of the reactors following an accident. The Passive Containment Cooling System (PCCS) is one of the several passive safety features in an Advanced Reactor (AHWR). In this paper, the APSRA methodology has been employed for reliability evaluation of the PCCS of AHWR. Performance assessment of the PCCS is initially performed during a postulated design basis LOCA using the best-estimate code RELAP5/Mod 3.2. The parameters affecting the system performance are then identified and considered for further analysis. Based on some pre-determined failure criterion, the failure surface for the system is predicted using the best-estimate code taking into account the deviations of the identified parameters from their nominal states as well as the model uncertainties inherent to the best estimate code. Root diagnosis is then carried out to determine the various failure causes, which occurs mainly due to malfunctioning of mechanical components. The failure probabilities of the various components are assessed through a classical PSA treatment using generic data. The reliability of the PCCS is then evaluated from the probability of availability of these components

  10. Economic assessment of the IRIS reactor for deployment in Brazil using INPRO methodology

    International Nuclear Information System (INIS)

    Goncalves Filho, Orlando Joao Agostinho

    2009-01-01

    This paper presents the main results of the evaluation of the economic competitiveness of the International Reactor Innovative and Secure (IRIS) for deployment in Brazil using the assessment methodology developed under the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO), co-ordinated by the International Atomic Energy Agency (IAEA). INPRO was initiated in 2001 and has the main objective of helping to ensure that nuclear energy will be available to contribute in a sustainable manner to the energy needs of the 21st century. Among its missions are the development of a methodology to assess innovative nuclear energy systems (INS) on a global, regional and national basis, and to facilitate the co-operation among IAEA Member States for planning the development and deployment of INS. Brazil joined INPRO since its beginning and in 2005 submitted a proposal for the screening assessment of two small-sized integral-type PWR reactors as alternative components of an INS completed with a conventional open nuclear fuel cycle based on enriched uranium. This paper outlines the rationale and the main results of the economic assessment of the IRIS-based INS completed in August 2008. The study concluded that IRIS reference design satisfies most of INPRO criteria in the area of economics. (author)

  11. Seismic Margin Assessment for Research Reactor using Fragility based Fault Tree Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kwag, Shinyoung; Oh, Jinho; Lee, Jong-Min; Ryu, Jeong-Soo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The research reactor has been often subjected to external hazards during the design lifetime. Especially, a seismic event can be one of significant threats to the failure of structure system of the research reactor. This failure is possibly extended to the direct core damage of the reactor. For this purpose, the fault tree for structural system failure leading to the core damage under an earthquake accident is developed. The failure probabilities of basic events are evaluated as fragility curves of log-normal distributions. Finally, the plant-level seismic margin is investigated by the fault tree analysis combining with fragility data and the critical path is identified. The plant-level probabilistic seismic margin assessment using the fragility based fault tree analysis was performed for quantifying the safety of research reactor to a seismic hazard. For this, the fault tree for structural system failure leading to the core damage of the reactor under a seismic accident was developed. The failure probabilities of basic events were evaluated as fragility curves of log-normal distributions.

  12. Carbon Dioxide-Mediated C(sp3)-H Arylation of Amine Substrates.

    Science.gov (United States)

    Kapoor, Mohit; Liu, Daniel; Young, Michael C

    2018-05-25

    Elaborating amines via C-H functionalization has been an important area of research over the past decade but has generally relied on an added directing group or sterically hindered amine approach. Since free-amine-directed C(sp 3 )-H activation is still primarily limited to cyclization reactions and to improve the sustainability and reaction scope of amine-based C-H activation, we present a strategy using CO 2 in the form of dry ice that facilitates intermolecular C-H arylation. This methodology has been used to enable an operationally simple procedure whereby 1° and 2° aliphatic amines can be arylated selectively at their γ-C-H positions. In addition to potentially serving as a directing group, CO 2 has also been demonstrated to curtail the oxidation of sensitive amine substrates.

  13. Probabilistic Safety Assessment Of It TRIGA Mark-II Reactor

    International Nuclear Information System (INIS)

    Ergun, E; Kadiroglu, O.S.

    1999-01-01

    The probabilistic safety assessment for Istanbul Technical University (ITU) TRIGA Mark-II reactor is performed. Qualitative analysis, which includes fault and event trees and quantitative analysis which includes the collection of data for basic events, determination of minimal cut sets, calculation of quantitative values of top events, sensitivity analysis and importance measures, uncertainty analysis and radiation release from fuel elements are considered

  14. Safety-licensing assessment of NASAP reactor concepts and fuel cycle facilities

    International Nuclear Information System (INIS)

    Lipinski, W.C.; Prohammer, F.G.; van Erp, J.B.; Seefeldt, W.B.

    1978-06-01

    Assessments are presented of the safety/licensability of reactor concepts based on information supplied by the Nonproliferation Alternative Systems Assessment Program (NASAP) characterization contractors in their updated responses to the data package for NASAP Rolling Report II. The assessment of the LMFBR includes information from a characterization contractor on alternate fuel cycles but does not include information provided by a characterization contractor on plant-related safety issues. The information provided by the characterization contractors was supplemented by assessments provided by the U. S. Nuclear Regulatory Commission

  15. Assessment of very high-temperature reactors in process applications

    International Nuclear Information System (INIS)

    Spiewak, I.; Jones, J.E. Jr.; Gambill, W.R.; Fox, E.C.

    1976-11-01

    An overview is presented of the technical and economic feasibility for the development of a very high-temperature reactor (VHTR) and associated processes. A critical evaluation of VHTR technology for process temperatures of 1400 and 2000 0 F is made. Additionally, an assessment of potential market impact is made to determine the commercial viability of the reactor system. It is concluded that VHTR process heat in the range of 1400 to 1500 0 F is attainable with near-term technology. However, process heat in excess of 1600 0 F would require considerably more materials development. The potential for the VHTR could include a major contribution to synthetic fuel, hydrogen, steel, and fertilizer production and to systems for transport and storage of high-temperature heat. A recommended development program including projected costs is presented

  16. Assessment of Silicon Carbide Composites for Advanced Salt-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Katoh, Yutai [ORNL; Wilson, Dane F [ORNL; Forsberg, Charles W [ORNL

    2007-09-01

    The Advanced High-Temperature Reactor (AHTR) is a new reactor concept that uses a liquid fluoride salt coolant and a solid high-temperature fuel. Several alternative fuel types are being considered for this reactor. One set of fuel options is the use of pin-type fuel assemblies with silicon carbide (SiC) cladding. This report provides (1) an initial viability assessment of using SiC as fuel cladding and other in-core components of the AHTR, (2) the current status of SiC technology, and (3) recommendations on the path forward. Based on the analysis of requirements, continuous SiC fiber-reinforced, chemically vapor-infiltrated SiC matrix (CVI SiC/SiC) composites are recommended as the primary option for further study on AHTR fuel cladding among various industrially available forms of SiC. Critical feasibility issues for the SiC-based AHTR fuel cladding are identified to be (1) corrosion of SiC in the candidate liquid salts, (2) high dose neutron radiation effects, (3) static fatigue failure of SiC/SiC, (4) long-term radiation effects including irradiation creep and radiation-enhanced static fatigue, and (5) fabrication technology of hermetic wall and sealing end caps. Considering the results of the issues analysis and the prospects of ongoing SiC research and development in other nuclear programs, recommendations on the path forward is provided in the order or priority as: (1) thermodynamic analysis and experimental examination of SiC corrosion in the candidate liquid salts, (2) assessment of long-term mechanical integrity issues using prototypical component sections, and (3) assessment of high dose radiation effects relevant to the anticipated operating condition.

  17. 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)

  18. Synthesis of benzimidazoles by PIDA-promoted direct C(sp2)-H imidation of N-arylamidines.

    Science.gov (United States)

    Huang, Jinbo; He, Yimiao; Wang, Yong; Zhu, Qiang

    2012-10-29

    A metal-free synthesis of diversified benzimidazoles from N-arylamidines through a phenyliodine(III) diacetate (PIDA) promoted intramolecular direct C(sp(2))-H imidation has been developed. The reaction proceeds smoothly at 0 °C or ambient temperature to provide the desired products in good to excellent yields. The synthesis of 2-alkyl- or 2-alkyl-fused benzimidazoles, which are generally inaccessible by similar Pd- or Cu-catalyzed approaches, can also be achieved. Copyright © 2012 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  19. Advanced neutron source reactor probabilistic flow blockage assessment

    International Nuclear Information System (INIS)

    Ramsey, C.T.

    1995-08-01

    The Phase I Level I Probabilistic Risk Assessment (PRA) of the conceptual design of the Advanced Neutron Source (ANS) Reactor identified core flow blockage as the most likely internal event leading to fuel damage. The flow blockage event frequency used in the original ANS PRA was based primarily on the flow blockage work done for the High Flux Isotope Reactor (HFIR) PRA. This report examines potential flow blockage scenarios and calculates an estimate of the likelihood of debris-induced fuel damage. The bulk of the report is based specifically on the conceptual design of ANS with a 93%-enriched, two-element core; insights to the impact of the proposed three-element core are examined in Sect. 5. In addition to providing a probability (uncertainty) distribution for the likelihood of core flow blockage, this ongoing effort will serve to indicate potential areas of concern to be focused on in the preliminary design for elimination or mitigation. It will also serve as a loose-parts management tool

  20. Assessment of a RELAP5 model for the IPR-R1 TRIGA research reactor

    International Nuclear Information System (INIS)

    Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Mesquita, Amir Z.; Soares, Humberto V.

    2010-01-01

    RELAP5 code was developed at the Idaho National Environmental and Engineering Laboratory and it is widely used for thermal hydraulic studies of commercial nuclear power plants and, currently, it has been also applied for thermal hydraulic analysis of nuclear research systems with good predictions. This work is a contribution to the assessment of RELAP5/3.3 code for research reactors analysis. It presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor conditions operating at 50 and 100 kW. The reactor is located at the Nuclear Technology Development Centre (CDTN), Brazil. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of code-to-data validation. The RELAP5 results were also compared with calculation performed using the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The use of a cross flow model has been essential to improve results in the transient condition respect to preceding investigations.

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

    International Nuclear Information System (INIS)

    Tinnes, S.P.; Cramer, D.S.; Logan, V.E.; Topp, S.V.; Smith, J.A.; Brandyberry, M.D.

    1990-01-01

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

  2. Inverse kinetics technique for reactor shutdown measurement: an experimental assessment. [AGR

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, T. A.; McDonald, D.

    1975-09-15

    It is proposed to use the Inverse Kinetics Technique to measure the subcritical reactivity as a function of time during the testing of the nitrogen injection systems on AGRs. A description is given of an experimental assessment of the technique by investigating known transients created by control rod movements on a small experimental reactor, (2m high, 1m radius). Spatial effects were observed close to the moving rods but otherwise derived reactivities were independent of detector position and agreed well with the existing calibrations. This prompted the suggestion that data from installed reactor instrumentation could be used to calibrate CAGR control rods.

  3. Design of A Vibration and Stress Measurement System for an Advanced Power Reactor 1400 Reactor Vessel Internals Comprehensive Vibration Assessment Program

    International Nuclear Information System (INIS)

    Ko, Doyoung; Kim, Kyuhyung

    2013-01-01

    In accordance with the US Nuclear Regulatory Commission (US NRC), Regulatory Guide 1.20, the reactor vessel internals comprehensive vibration assessment program (RVI CVAP) has been developed for an Advanced Power Reactor 1400 (APR1400). The purpose of the RVI CVAP is to verify the structural integrity of the reactor internals to flow-induced loads prior to commercial operation. The APR1400 RVI CVAP consists of four programs (analysis, measurement, inspection, and assessment). Thoughtful preparation is essential to the measurement program, because data acquisition must be performed only once. The optimized design of a vibration and stress measurement system for the RVI CVAP is essential to verify the integrity of the APR1400 RVI. We successfully designed a vibration and stress measurement system for the APR1400 RVI CVAP based on the design materials, the hydraulic and structural analysis results, and performance tests of transducers in an extreme environment. The measurement system designed in this paper will be utilized for the APR1400 RVI CVAP as part of the first construction project in Korea

  4. DESIGN OF A VIBRATION AND STRESS MEASUREMENT SYSTEM FOR AN ADVANCED POWER REACTOR 1400 REACTOR VESSEL INTERNALS COMPREHENSIVE VIBRATION ASSESSMENT PROGRAM

    Directory of Open Access Journals (Sweden)

    DO-YOUNG KO

    2013-04-01

    Full Text Available In accordance with the US Nuclear Regulatory Commission (US NRC, Regulatory Guide 1.20, the reactor vessel internals comprehensive vibration assessment program (RVI CVAP has been developed for an Advanced Power Reactor 1400 (APR1400. The purpose of the RVI CVAP is to verify the structural integrity of the reactor internals to flow-induced loads prior to commercial operation. The APR1400 RVI CVAP consists of four programs (analysis, measurement, inspection, and assessment. Thoughtful preparation is essential to the measurement program, because data acquisition must be performed only once. The optimized design of a vibration and stress measurement system for the RVI CVAP is essential to verify the integrity of the APR1400 RVI. We successfully designed a vibration and stress measurement system for the APR1400 RVI CVAP based on the design materials, the hydraulic and structural analysis results, and performance tests of transducers in an extreme environment. The measurement system designed in this paper will be utilized for the APR1400 RVI CVAP as part of the first construction project in Korea.

  5. Thermal stability of multilayered Pt-Al2O3 nanocoatings for high temperature CSP systems

    CSIR Research Space (South Africa)

    Nuru, ZY

    2015-10-01

    Full Text Available B), 115-120 Thermal stability of multilayered Pt-Al2O3 nanocoatings for high temperature CSP systems Z.Y. Nuru a, b, *, L. Kotsedi a, b, C.J. Arendse c, D. Motaung d, B. Mwakikunga d, K. Roro d, e, M. Maaza a, b a UNESCO-UNISA Africa Chair... Pretoria, South Africa e R&D Core-Energy, Council for Scientific and Industrial Research, P O Box 395, 0001 Pretoria, South Africa Abstract This contribution reports on the effect of thermal annealing on sputtered Pt–Al(sub2)O(sub3) multilayered...

  6. Probablistic risk assessment methodology application to Indian pressurised heavy water reactors

    International Nuclear Information System (INIS)

    Babar, A.K.; Grover, R.B.; Mehra, V.K.; Gangwal, D.K.; Chakraborty, G.

    1987-01-01

    Probabilistic risk assessment in the context of nuclear power plants is associated with models that predict the offsite radiological releases resulting from reactor accidents. Level 1 PRA deals with the identification of accident sequences relevant to the design of a system and also with their quantitative estimation. It is characterised by event tree, fault tree analysis. The initiating events applicable to pressurised heavy water reactors have been considered and the dominating initiating events essential for detailed studies are identified in this paper. Reliability analysis and the associated problems encountered during the case studies are mentioned briefly. It is imperative to validate the failure data used for analysis. Bayesian technique has been employed for the same and a brief account is included herein. A few important observations, e.g. effects of the presence of moderator, made during the application of probabilistic risk assessment methodology are also discussed. (author)

  7. Assessment of corrosion and fatigue damage to light water reactor metal containments

    International Nuclear Information System (INIS)

    Sinha, U.P.; Shah, V.N.; Smith, S.K.

    1991-01-01

    This paper presents a generic procedure for estimating aging damage, evaluating structural integrity, and identifying mitigation activities for safe operation of boiling water reactor (BWR) Mark I metal containments and ice-condenser type pressurized water reactor (PWR) cylindrical metal containments. The mechanisms of concern that can cause aging damage to these two types of containments are corrosion and fatigue. Assessment of fatigue damage to bellows is also described. Assessment of corrosion and fatigue damage described in this paper include: containment design features that are relevant to aging assessment, several corrosion and fatigue mechanisms, inspection of corrosion and fatigue damage, and mitigation of damage caused by these mechanisms. In addition, synergistic interaction between corrosion and fatigue is considered. Possible actions for mitigating aging include enhanced inspection methods, maintenance activities based on operating experience, and supplementary surveillance programs. Field experience related to aging of metal containments is reviewed. Finally, conclusions and recommendations are presented

  8. Reactor assessments of advanced bumpy torus configurations

    International Nuclear Information System (INIS)

    Uckan, N.A.; Owen, L.W.; Spong, D.A.; Miller, R.L.; Ard, W.B.; Pipkins, J.F.; Schmitt, R.J.

    1984-02-01

    Recently, several innovative approaches were introduced for enhancing the performance of the basic ELMO Bumpy Torus (EBT) concept and for improving its reactor potential. These include planar racetrack and square geometries, Andreoletti coil systems, and bumpy torus-stellarator hybrids (which include twisted racetrack and helical axis stellarator - snakey torus). Preliminary evaluations of reactor implications of each approach have been carried out based on magnetics (vacuum) calculations, transport and scaling relationships, and stability properties deduced from provisional configurations that implement the approach but are not necessarily optimized. Further optimization is needed in all cases to evaluate the full potential of each approach. Results of these studies indicate favorable reactor projections with a significant reduction in reactor physical size as compared to conventional EBT reactor designs carried out in the past

  9. LCOE reduction potential of parabolic trough and solar tower CSP technology until 2025

    Science.gov (United States)

    Dieckmann, Simon; Dersch, Jürgen; Giuliano, Stefano; Puppe, Michael; Lüpfert, Eckhard; Hennecke, Klaus; Pitz-Paal, Robert; Taylor, Michael; Ralon, Pablo

    2017-06-01

    Concentrating Solar Power (CSP), with an installed capacity of 4.9 GW by 2015, is a young technology compared to other renewable power generation technologies. A limited number of plants and installed capacity in a small challenging market environment make reliable and transparent cost data for CSP difficult to obtain. The International Renewable Energy Agency (IRENA) and the DLR German Aerospace Center gathered and evaluated available cost data from various sources for this publication in order to yield transparent, reliable and up-to-date cost data for a set of reference parabolic trough and solar tower plants in the year 2015 [1]. Each component of the power plant is analyzed for future technical innovations and cost reduction potential based on current R&D activities, ongoing commercial developments and growth in market scale. The derived levelized cost of electricity (LCOE) for 2015 and 2025 are finally contrasted with published power purchase agreements (PPA) of the NOOR II+III power plants in Morocco. At 7.5% weighted average cost of capital (WACC) and 25 years economic life time, the levelized costs of electricity for plants with 7.5 (trough) respectively 9 (tower) full-load hours thermal storage capacity decrease from 14-15 -ct/kWh today to 9-10 -ct/kWh by 2025 for both technologies at direct normal irradiation of 2500 kWh/(m².a). The capacity factor increases from 41.1% to 44.6% for troughs and from 45.5% to 49.0% for towers. Financing conditions are a major cost driver and offer potential for further cost reduction with the maturity of the technology and low interest rates (6-7 - ct/kWh for 2% WACC at 2500 kWh/(m2.a) in 2025).

  10. Assessment of the accident response of a light-water-moderated breeder-reactor system: AWBA development program

    International Nuclear Information System (INIS)

    High, H.M.

    1983-05-01

    The predicted accident response for a light water moderated, thorium/U-233 fueled, seed-blanket reactor concept was assessed. The first part of the assessment compared breeder accident response with that of a current commercial pressurized water reactor design for several different types of transients. Based on these comparisons and a review of the various parameter differences between the breeder and a U-235 fueled plant, the second part of the assessment studied the breeder accident behavior in more detail, particularly in areas of potential concern. Based on the two parts of the assessment, it was concluded that the breeder accident response would be very similar to that of present commercial pressurized water reactor plants. The large Doppler and moderator reactivity coefficients of the breeder would significantly reduce the severity of many of the accidents that must be considered. It is expected that the accident response of the breeder can be shown to meet regulatory criteria

  11. Pebble bed modular reactors versus other generation technologies. Costs and challenges for South Africa

    International Nuclear Information System (INIS)

    Grubert, Emily; Parks, Brian; Schneider, Erich; Sekar, Srinivas

    2011-01-01

    South Africa is Africa's major economy, with plans to double its electricity generation capacity by 2026. South Africa has spent almost two decades developing a nuclear reactor known as a Pebble Bed Modular Reactor (PBMR), which could provide substantial benefits to the electricity grid but was recently mothballed due to high costs. This work estimates the lifecycle financial costs of South African PBMRs, then compares these costs to those of five other generation options: coal, nuclear as pressurized water reactors (PWRs), wind, and solar as photovoltaics (PV) or concentrating solar power (CSP). Each technology is evaluated with low, base case, and high assumptions for capital costs, construction time, and interest rates. Decommissioning costs, project lifetime, capacity factors, and sensitivity to carbon price are also considered. PBMR could be cost competitive with coal under certain low cost conditions, even without a carbon price. However, international lending practices and other factors suggest that a high capital cost, high interest rate nuclear plant is likely to be competing with a low capital cost, low interest rate coal plant in a market where cost recovery is challenging. PBMR could potentially become more competitive if low rate international loans were available to nuclear projects or became unavailable to coal projects. (author)

  12. Probabilistic safety assessment framework of pebble-bed modular high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Liu Tao; Tong Jiejuan; Zhao Jun; Cao Jianzhu; Zhang Liguo

    2009-01-01

    After an investigation of similar reactor type probabilistic safety assessment (PSA) framework, Pebble-bed Modular High-Temperature Gas-cooled Reactor (HTR-PM) PSA framework was presented in correlate with its own design characteristics. That is an integral framework which spreads through event sequence structure with initiating events at the beginning and source term categories in the end. The analysis shows that it is HTR-PM design feature that determines its PSA framework. (authors)

  13. Probabilistic safety assessment of the PLUTO Research Reactor

    International Nuclear Information System (INIS)

    Preston, J.F.; Coates, D.A.

    1990-01-01

    The preliminary finding of a probabilistic safety assessment (PSA) carried out in support of a licensing submission are presented. The research reactor, a 25 MW highly enriched thermal reactor moderated and cooled by D 2 O, is housed in a steel containment building equipped with an active extract system to mitigate any possible release. A full PSA (to level 3) was performed based on the current operational plant making as much use of the plant operational records as possible. A medium sized event tree-fault tree approach was used to allow realistic modelling of operator actions. For reasons of practicality only plant damage states of core melt, fuel damage, and tritium release were defined, all release accident sequences being assigned to one of these states. Prior to discharge to the environment the releases were further sub-divided dependent upon the success of the active extract system. The individual and societal risks were calculated taking account of meterological and demographic conditions. The provisional results indicate that the core melt frequency is in the region of 1 x 10 -4 /yr, the dominant contributor being an unisolatable gross leakage beyond the capabilities of the recovery systems. The core melt frequency is comparable with those of power reactors of a similar age; however, the core inventory and hence release is much smaller; therefore the consequences are much reduced. The risk to an individual at any fixed location 100 m from the plant is assessed as 1 x 10 -6 ; the societal risk is estimated as 6 x 10 -4 . The main contributor to the dose received is from the released iodine. Additional benefit is being obtained from the PSA in several ways: the insights obtained into the function and operation are being incorporated into the operational safety document, whilst the source term results are being used to assist in the refurbishment/improvement of the active extract system

  14. Continuous Assessment of Safety Margin for the 14-MW TRIGA Reactor

    International Nuclear Information System (INIS)

    Ciocanescu, M.; Georgescu, D.; Doru, O.

    2008-01-01

    The assessment of reactor safety implies analyses of the reactor and its systems response to a range of postulated initiating events (such as malfunction or failures of equipment, operator errors, external events and so on which could lead to either anticipated operational occurrences or to accident conditions. Decreasing in heat removal by the reactor cooling system may be considered as a process disturbance which may lead to a postulated initiating event. The cold source for the reactor cooling system, in case of TRIGA-14 MW reactor is the atmosphere by the secondary cooling towers. The ability to evacuate the heat produced by the reactor core ranges between the outlet temperature of the core flow and the outdoors temperature in air, which is subject to season and day variation. Selected values for safety limits, safety system settings and limiting condition(s) are derived from safety analysis and are consistent with the operational state of the reactor. When a limiting condition for safe operation is not satisfied, the operating personal is supposed to take the appropriate action(s) to ensure safety. Operating requirements and the safety system are presented. The reactor operating safety parameters from the main Data Acquisition System are transferred to an AT personal computer. These selected parameters are the following: - average inlet temperature which is calculated as an average temperature measured by 20 type K thermocouples distributed within a 4 x 5 matrix located on the top of the reactor core; - average outlet temperature which is calculated as an average record from 10 type K thermocouples placed in the outlet pipe; - average flow rate which is calculated as an average value from four transducers (two for the inlet flow rate and two for the outlet flow rate). Due to its high instability, this value is also filtered using a two-pole low-pass filter (software); - reactor thermal power value derivable from the previous parameters or obtained from the

  15. Environmental assessment for the deactivation of the N Reactor facilities

    International Nuclear Information System (INIS)

    1995-05-01

    This environmental assessment (EA) provides information for the US Department of Energy (DOE) to decide whether the Proposed Action for the N Reactor facilities warrants a Finding of No Significant Impact or requires the preparation of an environmental impact statement (EIS). The EA describes current conditions at the N Reactor facilities, the need to take action at the facilities, the elements of the Proposed Action and alternatives, and the potential environmental impacts. As required by the National Environmental Policy Act of 1969 (NEPA), this EA complies with Title 40, Code of Federal Regulations (CFR), parts 1500--1508, ''Regulations for Implementing the Procedural Provisions of NEPA. '' It also implements the ''National Environmental Policy Act; Implementing Procedures and Guidelines'' (10 CFR 1021)

  16. Uncertainty Quantification in the Reliability and Risk Assessment of Generation IV Reactors: Final Scientific/Technical Report

    International Nuclear Information System (INIS)

    Vierow, Karen; Aldemir, Tunc

    2009-01-01

    The project entitled, 'Uncertainty Quantification in the Reliability and Risk Assessment of Generation IV Reactors', was conducted as a DOE NERI project collaboration between Texas A and M University and The Ohio State University between March 2006 and June 2009. The overall goal of the proposed project was to develop practical approaches and tools by which dynamic reliability and risk assessment techniques can be used to augment the uncertainty quantification process in probabilistic risk assessment (PRA) methods and PRA applications for Generation IV reactors. This report is the Final Scientific/Technical Report summarizing the project.

  17. Emergency planning and response: An independent safety assessment of Department of Energy nuclear reactor facilities

    International Nuclear Information System (INIS)

    Knuth, D.; Boyd, R.

    1981-02-01

    The Department of Energy (DOE) has formed a Nuclear Facilities Personnel Qualification and Training (NFPQT) Committee to assess the implications of the recommendations contained in the President's Commission Report on the Three Mile Island (TMI) Accident (the Kemeny Commission report) that are applicable to DOE's nuclear reactor operations. Thirteen DOE nuclear reactors have been reviewed. The assessments of the 13 facilities are based on information provided by the individual operator organizations and/or cognizant DOE Field Offices. Additional clarifying information was supplied in some, but not all, instances. This report indicates how these 13 reactor facilities measure up in light of the Kemeny and other TMI-related studies and recommendations, particularly those that have resulted in upgraded Nuclear Regulatory Commission (NRC) requirements in the area of emergency planning and response

  18. Dose assessment around TR-2 reactor due to maximum credible accident

    International Nuclear Information System (INIS)

    Turgut, M. H.; Adalioglu, U.; Aytekin, A.

    2001-01-01

    The revision of safety analysis report of TR-2 research reactor had been initiated in 1995. The whole accident analysis and accepted scenario for maximum credible accident has been revised according to the new safety concepts and the impact to be given to the environment due to this scenario has been assessed. This paper comprises all results of these calculations. The accepted maximum credible accident scenario is the partial blockage of the whole reactor core which resulted in the release of 25% of the core inventory. The DOSER code which uses very conservative modelling of atmospheric distributions were modified for the assessment calculations. Pasquill conditions based on the local weather observations, topography, and building affects were considered. The thyroid and whole body doses for 16 sectors and up to 10 km of distance around CNAEM were obtained. Release models were puff and a prolonged one of two hours of duration. Release fractions for the active isotopes were chosen from literature which were realistic

  19. Preliminary Assessment of Two Alternative Core Design Concepts for the Special Purpose Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Werner, James E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hummel, Andrew J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kennedy, John C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); O' Brien, Robert C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Dion, Axel M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wright, Richard N. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ananth, Krishnan P. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-11-01

    The Special Purpose Reactor (SPR) is a small 5 MWt, heat pipe-cooled, fast reactor based on the Los Alamos National Laboratory (LANL) Mega-Power concept. The LANL concept features a stainless steel monolithic core structure with drilled channels for UO2 pellet stacks and evaporator sections of the heat pipes. Two alternative active core designs are presented here that replace the monolithic core structure with simpler and easier to manufacture fuel elements. The two new core designs are simply referred to as Design A and Design B. In addition to ease of manufacturability, the fuel elements for both Design A and Design B can be individually fabricated, assembled, inspected, tested, and qualified prior to their installation into the reactor core leading to greater reactor system reliability and safety. Design A fuel elements will require the development of a new hexagonally-shaped UO2 fuel pellet. The Design A configuration will consist of an array of hexagonally-shaped fuel elements with each fuel element having a central heat pipe. This hexagonal fuel element configuration results in four radial gaps or thermal resistances per element. Neither the fuel element development, nor the radial gap issue are deemed to be serious and should not impact an aggressive reactor deployment schedule. Design B uses embedded arrays of heat pipes and fuel pins in a double-wall tank filled with liquid metal sodium. Sodium is used to thermally bond the heat pipes to the fuel pins, but its usage may create reactor transportation and regulatory challenges. An independent panel of U.S. manufacturing experts has preliminarily assessed the three SPR core designs and views Design A as simplest to manufacture. Herein are the results of a preliminary neutronic, thermal, mechanical, material, and manufacturing assessment of both Design A and Design B along with comparisons to the LANL concept (monolithic core structure). Despite the active core differences, all three reactor concepts behave

  20. Molten Salt Reactor in the Overview and Perspective of Technological Assessment

    International Nuclear Information System (INIS)

    Julia Abdul Karim; Khaironie Md Takip; Muhammad Khairul Arif Mustafa; Mohd Hairie Rabir; Lanyau, T.; Tom, P.P.

    2016-01-01

    Full text: A Molten Salt Reactor (MSR) is unique in its characteristics that offer safer operation, deliver efficient power output that can assure in the sustainable energy production without CO_2 emissions. Several concepts of this kind of reactor have been proposed by stake holder with different design and configuration and up to date they are exasperating to obtain an optimum workable solution to the fuel salt composition in the foresee of neutronic properties, operating temperature, actinide and fission products solubility, chemical control and processing, materials compatibility and handling of waste. Hence, these key issues are wide open as the potential Research and Development in the specific areas of studies. In addition to that, concern arise in the viewpoint of socioeconomic, politics, public acceptance, safety and security, proven technology, proliferation resistance and physical protection that also need to give special attention in problem solving. The worldwide collaboration through Gen IV International Forum has discussed the potential of MSR and addresses on the issues globally. Recently, Malaysia has taken an initiative aiming to participate in MSR studies due to its potential as an energy source using thorium. Therefore, this paper is focusing on the technology assessment for Thorium-breeding Molten Salt Reactor (TMSR) especially on the ability of utilizing thorium as fuel. This assessment also will help to enhance the understanding of thorium beneficiation to cater for the energy demand. (author)

  1. Preliminary assessment of an S.G.H.W. type research reactor

    International Nuclear Information System (INIS)

    Bicevskis, A.; Chapman, A.G.; Hesse, E.W.

    1970-08-01

    A preliminary design study has been made of a research reactor, based on the enriched S.G.H.W.R. concept, to be used for power reactor fuel irradiation, isotope production, basic research, and training in nuclear technology. A reactor physics assessment established a core size which would allow uninterrupted operation for the required irradiation period consistent with low capital and operating costs. A design was selected with 24 channels, a D 2 O calandria diameter of 2.7 m and an overall core height of 4.0 m. The capital cost was estimated as $750,000 for the fuel and $1,600,000 for the moderator, the refuelling cost being $340,000 per annum. A thermal design study showed that the fission heat of 65 MW could be transmitted to pressurised light water at 200 lb/in 2 abs. and rejected to sea water in two conventional U-tube heat exchangers. The basic design is flexible and can be adapted to meet many special requirements. (author)

  2. Aging assessment and mitigation for major LWR [light water reactor] components

    International Nuclear Information System (INIS)

    Shah, Y.N.; Ware, A.G.; Conley, D.A.; MacDonald, P.E.; Burns, J.J. Jr.

    1989-01-01

    This paper summarizes some of the results of the Aging Assessment and Mitigation Project sponsored by the US Nuclear Regulatory Commission (USNRC), Office of Nuclear Regulatory Research. The objective of the project is to develop an understanding of the aging degradation of the major light water reactor (LWR) structures and components and to develop methods for predicting the useful life of these components so that the impact of aging on the safe operation of nuclear power plants can be evaluated and addressed. The research effort consists of integrating, evaluating, and updating the available aging-related information. This paper discusses current accomplishments and summarizes the significant degradation processes active in two major components: pressurized water reactor pressurizer surge and spray lines and nozzles, and light water reactor primary coolant pumps. This paper also evaluates the effectiveness of the current inservice inspection programs and presents conclusions and recommendations related to aging of these two major components. 37 refs., 7 figs., 3 tabs

  3. Assessment of the thorium fuel cycle in power reactors

    International Nuclear Information System (INIS)

    Kasten, P.R.; Homan, F.J.; Allen, E.J.

    1977-01-01

    A study was conducted at Oak Ridge National Laboratory to evaluate the role of thorium fuel cycles in power reactors. Three thermal reactor systems were considered: Light Water Reactors (LWRs); High-Temperature Gas-Cooled Reactors (HTGRs); and Heavy Water Reactors (HWRs) of the Canadian Deuterium Uranium Reactor (CANDU) type; most of the effort was on these systems. A summary comparing thorium and uranium fuel cycles in Fast Breeder Reactors (FBRs) was also compiled

  4. Safety assessment principles for reactor protection systems in the United Kingdom

    International Nuclear Information System (INIS)

    Philp, W.

    1990-01-01

    The duty of Nuclear Installations Inspectorate (NII) is to see that the appropriate standards are developed, achieved and maintained by the plant operators, and to monitor and regulate the safety of the plant by means of its powers under the licence. It does not issue standards or codes of practice for NPPs, but it requires each plant operator to develop its own safety criteria and requirements. The following relevant issues are described: NII assessment principles and societal risks; principles and guidance for the assessment of rector protection systems; assessment of reactor shutdown systems

  5. Safety assessment principles for reactor protection systems in the United Kingdom

    Energy Technology Data Exchange (ETDEWEB)

    Philp, W

    1990-07-01

    The duty of Nuclear Installations Inspectorate (NII) is to see that the appropriate standards are developed, achieved and maintained by the plant operators, and to monitor and regulate the safety of the plant by means of its powers under the licence. It does not issue standards or codes of practice for NPPs, but it requires each plant operator to develop its own safety criteria and requirements. The following relevant issues are described: NII assessment principles and societal risks; principles and guidance for the assessment of rector protection systems; assessment of reactor shutdown systems.

  6. Procedures to relate the NII safety assessment principles for nuclear reactors to risk

    CERN Document Server

    Kelly, G N; Hemming, C R

    1985-01-01

    Within the framework of the Public Inquiry into the proposed pressurised water reactor (PWR) at Sizewell, estimates were made of the levels of individual and societal risk from a PWR designed in a manner which would conform to the safety assessment principles formulated by the Nuclear Installations Inspectorate (NII). The procedures used to derive these levels of risk are described in this report. The opportunity has also been taken to revise the risk estimates made at the time of the Inquiry by taking account of additional data which were not then available, and to provide further quantification of the likely range of uncertainty in the predictions. This re-analysis has led to small changes in the levels of risk previously evaluated, but these are not sufficient to affect the broad conclusions reached before. For a reactor just conforming to the NII safety assessment principles a maximum individual risk of fatal cancer of about 10 sup - sup 6 per year of reactor operation has been estimated; the societal ris...

  7. SequenceL: Automated Parallel Algorithms Derived from CSP-NT Computational Laws

    Science.gov (United States)

    Cooke, Daniel; Rushton, Nelson

    2013-01-01

    With the introduction of new parallel architectures like the cell and multicore chips from IBM, Intel, AMD, and ARM, as well as the petascale processing available for highend computing, a larger number of programmers will need to write parallel codes. Adding the parallel control structure to the sequence, selection, and iterative control constructs increases the complexity of code development, which often results in increased development costs and decreased reliability. SequenceL is a high-level programming language that is, a programming language that is closer to a human s way of thinking than to a machine s. Historically, high-level languages have resulted in decreased development costs and increased reliability, at the expense of performance. In recent applications at JSC and in industry, SequenceL has demonstrated the usual advantages of high-level programming in terms of low cost and high reliability. SequenceL programs, however, have run at speeds typically comparable with, and in many cases faster than, their counterparts written in C and C++ when run on single-core processors. Moreover, SequenceL is able to generate parallel executables automatically for multicore hardware, gaining parallel speedups without any extra effort from the programmer beyond what is required to write the sequen tial/singlecore code. A SequenceL-to-C++ translator has been developed that automatically renders readable multithreaded C++ from a combination of a SequenceL program and sample data input. The SequenceL language is based on two fundamental computational laws, Consume-Simplify- Produce (CSP) and Normalize-Trans - pose (NT), which enable it to automate the creation of parallel algorithms from high-level code that has no annotations of parallelism whatsoever. In our anecdotal experience, SequenceL development has been in every case less costly than development of the same algorithm in sequential (that is, single-core, single process) C or C++, and an order of magnitude less

  8. MANAGEMENT OF RESEARCH AND TEST REACTOR ALUMINUM SPENT NUCLEAR FUEL - A TECHNOLOGY ASSESSMENT

    Energy Technology Data Exchange (ETDEWEB)

    Vinson, D.

    2010-07-11

    The Department of Energy's Environmental Management (DOE-EM) Program is responsible for the receipt and storage of aluminum research reactor spent nuclear fuel or used fuel until ultimate disposition. Aluminum research reactor used fuel is currently being stored or is anticipated to be returned to the U.S. and stored at DOE-EM storage facilities at the Savannah River Site and the Idaho Nuclear Technology and Engineering Center. This paper assesses the technologies and the options for safe transportation/receipt and interim storage of aluminum research reactor spent fuel and reviews the comprehensive strategy for its management. The U.S. Department of Energy uses the Appendix A, Spent Nuclear Fuel Acceptance Criteria, to identify the physical, chemical, and isotopic characteristics of spent nuclear fuel to be returned to the United States under the Foreign Research Reactor Spent Nuclear Fuel Acceptance Program. The fuel is further evaluated for acceptance through assessments of the fuel at the foreign sites that include corrosion damage and handleability. Transport involves use of commercial shipping casks with defined leakage rates that can provide containment of the fuel, some of which are breached. Options for safe storage include wet storage and dry storage. Both options must fully address potential degradation of the aluminum during the storage period. This paper focuses on the various options for safe transport and storage with respect to technology maturity and application.

  9. Response tree evaluation: experimental assessment of an expert system for nuclear reactor operators

    International Nuclear Information System (INIS)

    Nelson, W.R.; Blackman, H.S.

    1985-09-01

    The United States Nuclear Regulatory Commission (USNRC) sponsored a project performed by EG and G Idaho, Inc., at the Idaho National Engineering Laboratory (INEL) to evaluate different display concepts for use in nuclear reactor control rooms. Included in this project was the evaluation of the response tree computer-based decision aid and its associated displays. The response tree evaluation task was deisgned to (1) assess the merit of the response tree decision aid and (2) develop a technical basis for recommendations, guidelines, and criteria for the design and evaluation of computerized decision aids for use in reactor control rooms. Two major experiments have been conducted to evaluate the response tree system. This report emphasizes the conduct and results of the second experiment. An enhanced version of the response tree system, known as the automated response tree system, was used in a controlled experiment using trained reactor operators as test subjects. This report discusses the automated response tree system, the design of the evaluation experiment, and the quantitative results of the experiment. The results of the experiment are compared to the results of the previous experiment to provide an integrated perspective of the response tree evaluation project. In addition, a subjective assessment of the results addresses the implications for the use of advanced, ''intelligent'' decision aids in the reactor control room

  10. Self-assessment of application of the Code of Conduct on the safety of research reactors - Mexico

    International Nuclear Information System (INIS)

    Mamani-Alegria, Y.R.; Salgado-Gonzalez, J.R.; Miranda-Aldaco, J.

    2009-01-01

    In Mexico, the nuclear regulatory body is the National Commission on Nuclear Safety and Safeguards (CNSNS), and there is one research reactor, a TRIGA MARK III, operated by the National Institute for Nuclear Research (ININ). The main aspects of the Self-assessment of application of The Code of Conduct on the Safety of Research Reactor are given for the case of the TRIGA reactor. Furthermore, in this paper we give a brief description of the legal framework of the licensing process, for nuclear activities in a research reactor, there are also highlights of the major reactor features, the uses of the reactor for isotope production, the management and verification of safety, the radiation protection management program, the emergency planning and the training and qualification of the operation personnel. (author)

  11. Plasma, a plant safety monitoring and assessment system for VVER-440 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hornaes, A.; Hulsund, J. E. [Institutt for energiteknikk (IFE), OECD Halden Reactor Project, Halden (Norway); Lipcsei, S.; Major, Cs.; Racz, A.; Vegh, J. [KFKI, Atomic Energy Research Institute, Budapest (Hungary); Eiler, J. [Paks, Nuclear Power Plant Ltd, Paks (Hungary)

    1999-05-15

    The objective with the Plant Safety Monitoring and Assessment System (PLASMA) is to develop an operator support system to support the execution of new symptom-based Emergency Operating Procedures for application in VVER reactors, with the Paks NPP in Hungary as the target plant. Many of the VVER reactors are rewriting their EOPs to comply more with Western standards of symptom-based EOPs. In this connection it is desirable to improve the data validation, information integration and presentation for operators when executing the EOPs. The entry-point to a symptom-oriented procedure is defined by the occurrence of a well-defined reactor operation status, with all its symptoms. However, the application of the EOF benefits from an operator support system, which performs plant status and symptom identification reliably and accurately. The development of the PLASMA system is a joint venture between Institutt for energiteknikk (IFE) and KFKI with the NPP Paks as the target plant. The project has been initiated and partly funded by the Science and Technology Agency (STA), Japan through the OECD NEA assistance program. In Hungary, considerable effort has concentrated on the safety reassessment of the Paks NPP and new EOPs are being written, but no comprehensive Operator Support System (OSS) for plant safety assessment is installed. Some safety parameter display functions are incorporated into diverse operator support systems, but an online 'plant safety monitoring and assessment system' is still missing. The present project comprises designing, constructing, testing and installing such an OSS, which to a great extent could support plant operators in their safety assessment work (author) (ml)

  12. Quality assurance: an independent safety assessment of Department of Energy nuclear reactor facilities

    International Nuclear Information System (INIS)

    Frankhouser, W.L.; Bass, W. Jr.; Langston, M.E.

    1981-02-01

    This report presents the assessments of QA programs at eight DOE reactor sites as performed by three members of the NFPQT Support Team. A summation of assessments is presented. That summation includes discussion of findings and recommendations for follow-on actions. The detailed record of contractor-by-contractor reviews is provided. A discussion of the approach in performing the QA assessments is presented, and the basis and limitations of the assessments are discussed

  13. Shielding assessment of the ETRR-1 Reactor Under power upgrading

    Energy Technology Data Exchange (ETDEWEB)

    Ahmad, E E [Reactor Department, Nuclear Research Center, Atomic Energy Authority, Cairo (Egypt)

    1997-12-31

    The assessment of existing shielding of the ETRR-1 reactor in case of power upgrading is presented and discussed. It was carried out using both the present EK-10 type fuel elements and some other types of fuel elements with different enrichments. The shielding requirements for the ETRR-1 when power is upgraded are also discussed. The optimization curves between the upgraded reactor power and the shield thickness are presented. The calculation have been made using the ANISN code with the DLC-75 data library. The results showed that the present shield necessitates an additional layer of steel with thickness of 10.20 and 25 cm. When its power is upgraded to 3, 6 and 10 MWt in order to cutoff all neutron energy groups to be adequately safe under normal operating conditions. 4 figs.

  14. Progress in Methodologies for the Assessment of Passive Safety System Reliability in Advanced Reactors. Results from the Coordinated Research Project on Development of Advanced Methodologies for the Assessment of Passive Safety Systems Performance in Advanced Reactors

    International Nuclear Information System (INIS)

    2014-09-01

    Strong reliance on inherent and passive design features has become a hallmark of many advanced reactor designs, including several evolutionary designs and nearly all advanced small and medium sized reactor (SMR) designs. Advanced nuclear reactor designs incorporate several passive systems in addition to active ones — not only to enhance the operational safety of the reactors but also to eliminate the possibility of serious accidents. Accordingly, the assessment of the reliability of passive safety systems is a crucial issue to be resolved before their extensive use in future nuclear power plants. Several physical parameters affect the performance of a passive safety system, and their values at the time of operation are unknown a priori. The functions of passive systems are based on basic physical laws and thermodynamic principals, and they may not experience the same kind of failures as active systems. Hence, consistent efforts are required to qualify the reliability of passive systems. To support the development of advanced nuclear reactor designs with passive systems, investigations into their reliability using various methodologies are being conducted in several Member States with advanced reactor development programmes. These efforts include reliability methods for passive systems by the French Atomic Energy and Alternative Energies Commission, reliability evaluation of passive safety system by the University of Pisa, Italy, and assessment of passive system reliability by the Bhabha Atomic Research Centre, India. These different approaches seem to demonstrate a consensus on some aspects. However, the developers of the approaches have been unable to agree on the definition of reliability in a passive system. Based on these developments and in order to foster collaboration, the IAEA initiated the Coordinated Research Project (CRP) on Development of Advanced Methodologies for the Assessment of Passive Safety Systems Performance in Advanced Reactors in 2008. The

  15. 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

  16. 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

  17. Studies on the assessment and validation of reactor dynamics models used in Finland

    International Nuclear Information System (INIS)

    Vanttola, T.

    1993-10-01

    Two reactor dynamics related computer codes of the calculation system at the Technical Research Centre of Finland have been assessed. The codes TRAB and SMATRA, have been examined from two points of view. First, models of some critical phenomena determining the worst fuel rod conditions during reactor transients have been evaluated on the basis of experimental information. Second, the the overall behaviour of the codes describing the dynamics of the reactor core and its cooling system has been studied on the basis of simulation of real transients and of performed safety analyses of selected accidents. The emphasis is on the VVER-440 reactors, but the generality of the methods has been demonstrated by showing that the key phenomena of the Chernobyl accident can be reproduced and analysed using the same calculation system. In the study the separate phenomena examined are single- and two-phase friction, post DNB heat transfer and critical heat flux in the VVER rod bundle. (60 refs., 11 figs., 4 tabs.)

  18. Structure shielding from cloud and fallout gamma ray sources for assessing the consequences of reactor accidents

    International Nuclear Information System (INIS)

    Burson, Z.G.; Profio, A.E.

    1975-12-01

    Radiation shielding provided by transportation vehicles and structures typical of where people live and work were estimated for cloud and fallout gamma-ray sources resulting from a hypothetical reactor accident. Dose reduction factors are recommended for a variety of situations for realistically assessing the consequences of reactor accidents

  19. IgG2 antibodies against a clinical grade Plasmodium falciparum CSP vaccine antigen associate with protection against transgenic sporozoite challenge in mice.

    Directory of Open Access Journals (Sweden)

    Robert Schwenk

    Full Text Available The availability of a highly purified and well characterized circumsporozoite protein (CSP is essential to improve upon the partial success of recombinant CSP-based malaria vaccine candidates. Soluble, near full-length, Plasmodium falciparum CSP vaccine antigen (CS/D was produced in E. coli under bio-production conditions that comply with current Good Manufacturing Practices (cGMP. A mouse immunogenicity study was conducted using a stable oil-in-water emulsion (SE of CS/D in combination with the Toll-Like Receptor 4 (TLR4 agonist Glucopyranosyl Lipid A (GLA/SE, or one of two TLR7/8 agonists: R848 (un-conjugated or 3M-051 (covalently conjugated. Compared to Alum and SE, GLA/SE induced higher CS/D specific antibody response in Balb/c mice. Subclass analysis showed higher IgG2:IgG1 ratio of GLA/SE induced antibodies as compared to Alum and SE. TLR synergy was not observed when soluble R848 was mixed with GLA/SE. Antibody response of 3M051 formulations in Balb/c was similar to GLA/SE, except for the higher IgG2:IgG1 ratio and a trend towards higher T cell responses in 3M051 containing groups. However, no synergistic enhancement of antibody and T cell response was evident when 3M051 conjugate was mixed with GLA/SE. In C57Bl/6 mice, CS/D adjuvanted with 3M051/SE or GLA/SE induced higher CSP repeat specific titers compared to SE. While, 3M051 induced antibodies had high IgG2c:IgG1 ratio, GLA/SE promoted high levels of both IgG1 and IgG2c. GLA/SE also induced more potent T-cell responses compared to SE in two independent C57/BL6 vaccination studies, suggesting a balanced and productive T(H1/T(H2 response. GLA and 3M-051 similarly enhanced the protective efficacy of CS/D against challenge with a transgenic P. berghei parasite and most importantly, high levels of cytophilic IgG2 antibodies were associated with protection in this model. Our data indicated that the cGMP-grade, soluble CS/D antigen combined with the TLR4-containing adjuvant GLA/SE warrants

  20. Environmental impact assessment of Ar-41 released by the normal operation of TRIGA-Mark 2 research reactor

    International Nuclear Information System (INIS)

    Qassoud, D.; Soufi, I.; Ziagos, J.; Demir, Z

    2007-01-01

    Full text: In accordance with the international regulation of nuclear safety and radiological protection of the environment applicable to the basic nuclear installations, category in which the Triga-Mark 2 research reactor is considered, an assesment of the impact in to the environment of the Ar-41 radioelement is accomplished. This radioelement is released by the normal operation of this reactor. The assessment is based on the characteristics of a Moroccan site (where the reactor is installed). It is carried out using CEA Gaussian models and mathematical models developed in LLNL. Considering the assumptions of impact assessments of the radioactivity in the atmosphere, the most important exposure is relatively corresponding to 1 Km from the reactor. This exposure is approximately 0,07% of the lawful limit. Beyond this locality, the exposure becomes lower than 0,02% of this limit. Beyond 5 Km, it becomes lower than ten nono-Sivert. In the basis of the site radiological baseline, the environmental impact of Ar-41 released in normal operation of the reactor is negligible in the studied case. [fr

  1. Assessment of the Capability of Molten Salt Reactors as a Next Generation High Temperature Reactors

    International Nuclear Information System (INIS)

    Elsheikh, B.M.

    2017-01-01

    Molten Salt Reactor according to Aircraft Reactor Experiment (ARE) and the Molten Salt Reactor Experiment (MSRE) programs, was designed to be the first full-scale, commercial nuclear power plant utilizing molten salt liquid fuels that can be used for producing electricity, and producing fissile fuels (breeding)burning actinides. The high temperature in the primary cycle enables the realization of efficient thermal conversion cycles with net thermal efficiencies reach in some of the designs of nuclear reactors greater than 45%. Molten salts and liquid salt because of their low vapor pressure are excellent candidates for meeting most of the requirements of these high temperature reactors. There is renewed interest in MSRs because of changing goals and new technologies in the use of high-temperature reactors. Molten Salt Reactors for high temperature create substantial technical challenges to have high effectiveness intermediate heat transfer loop components. This paper will discuss and investigate the capability and compatibility of molten salt reactors, toward next generation high temperature energy system and its technical challenges

  2. Fracture assessment of Savannah River Reactor carbon steel piping

    International Nuclear Information System (INIS)

    Mertz, G.E.; Stoner, K.J.; Caskey, G.R.; Begley, J.A.

    1991-01-01

    The Savannah River Site (SRS) production reactors have been in operation since the mid-1950's. One postulated failure mechanism for the reactor piping is brittle fracture of the original A285 and A53 carbon steel piping. Material testing of archival piping determined (1) the static and dynamic tensile properties; (2) Charpy impact toughness; and (3) the static and dynamic compact tension fracture toughness properties. The nil-ductility transition temperature (NDTT), determined by Charpy impact test, is above the minimum operating temperature for some of the piping materials. A fracture assessment was performed to demonstrate that potential flaws are stable under upset loading conditions and minimum operating temperatures. A review of potential degradation mechanisms and plant operating history identified weld defects as the most likely crack initiation site for brittle fracture. Piping weld defects, as characterized by radiographic and metallographic examination, and low fracture toughness material properties were postulated at high stress locations in the piping. Normal operating loads, upset loads, and residual stresses were assumed to act on the postulated flaws. Calculated allowable flaw lengths exceed the size of observed weld defects, indicating adequate margins of safety against brittle fracture. Thus, a detailed fracture assessment was able to demonstrate that the piping systems will not fail by brittle fracture, even though the NDTT for some of the piping is above the minimum system operating temperature

  3. Assessment of the TRINO reactor pressure vessel integrity: theoretical analysis and NDE

    Energy Technology Data Exchange (ETDEWEB)

    Milella, P P; Pini, A [ENEA, Rome (Italy)

    1988-12-31

    This document presents the method used for the capability assessment of the Trino reactor pressure vessel. The vessel integrity assessment is divided into the following parts: transients evaluation and selection, fluence estimate for the projected end of life of the vessel, characterization of unirradiated and irradiated materials, thermal and stress analysis, fracture mechanics analysis and eventually fracture input to Non Destructive Examination (NDE). For each part, results are provided. (TEC).

  4. PREFACE: International conference on Computer Simulation in Physics and beyond (CSP2015)

    Science.gov (United States)

    2016-02-01

    The International conference on Computer Simulations in Physics and beyond (CSP2015) was held from 6-10 September 2015 at the campus of the Moscow Institute for Electronics and Mathematics (MIEM), National Research University Higher School of Economics, Moscow. Computer simulations are in increasingly popular tool for scientific research, supplementing experimental and analytical research. The main goal of the conference is contributing to the development of methods and algorithms which take into account trends in hardware development, which may help with intensive research. The conference also allowed senior scientists and students to have the opportunity to speak each other and exchange ideas and views on the developments in the area of high-performance computing in science. We would like to take this opportunity to thank our sponsors: the Russian Foundation for Basic Research, Federal Agency of Scientific Organizations, and Higher School of Economics.

  5. Nuclear proliferation and civilian nuclear power: report of the Nonproliferation Alternative Systems Assessment Program. Volume IX. Reactor and fuel cycle descriptions

    Energy Technology Data Exchange (ETDEWEB)

    1979-12-01

    The Nonproliferation Alternative Systems Assessment Program (NASAP) has characterized and assessed various reactor/fuel-cycle systems. Volume IX provides, in summary form, the technical descriptions of the reactor/fuel-cycle systems studied. This includes the status of the system technology, as well as a discussion of the safety, environmental, and licensing needs from a technical perspective. This information was then used in developing the research, development, and demonstration (RD and D) program, including its cost and time frame, to advance the existing technology to the level needed for commercial use. Wherever possible, the cost data are given as ranges to reflect the uncertainties in the estimates. Volume IX is divided into three sections: Chapter 1, Reactor Systems; Chapter 2, Fuel-Cycle Systems; and the Appendixes. Chapter 1 contains the characterizations of the following 12 reactor types: light-water reactor; heavy-water reactor; water-cooled breeder reactor; high-temperature gas-cooled reactor; gas-cooled fast reactor; liquid-metal fast breeder reactor; spectral-shift-controlled reactor; accelerator-driven reactor; molten-salt reactor; gaseous-core reactor; tokamak fusion-fisson hybrid reactor; and fast mixed-spectrum reactor. Chapter 2 contains similar information developed for fuel-cycle facilities in the following categories: mining and milling; conversion and enrichment; fuel fabrication; spent fuel reprocessing; waste handling and disposal; and transportation of nuclear materials.

  6. Nuclear proliferation and civilian nuclear power: report of the Nonproliferation Alternative Systems Assessment Program. Volume IX. Reactor and fuel cycle descriptions

    International Nuclear Information System (INIS)

    1979-12-01

    The Nonproliferation Alternative Systems Assessment Program (NASAP) has characterized and assessed various reactor/fuel-cycle systems. Volume IX provides, in summary form, the technical descriptions of the reactor/fuel-cycle systems studied. This includes the status of the system technology, as well as a discussion of the safety, environmental, and licensing needs from a technical perspective. This information was then used in developing the research, development, and demonstration (RD and D) program, including its cost and time frame, to advance the existing technology to the level needed for commercial use. Wherever possible, the cost data are given as ranges to reflect the uncertainties in the estimates. Volume IX is divided into three sections: Chapter 1, Reactor Systems; Chapter 2, Fuel-Cycle Systems; and the Appendixes. Chapter 1 contains the characterizations of the following 12 reactor types: light-water reactor; heavy-water reactor; water-cooled breeder reactor; high-temperature gas-cooled reactor; gas-cooled fast reactor; liquid-metal fast breeder reactor; spectral-shift-controlled reactor; accelerator-driven reactor; molten-salt reactor; gaseous-core reactor; tokamak fusion-fisson hybrid reactor; and fast mixed-spectrum reactor. Chapter 2 contains similar information developed for fuel-cycle facilities in the following categories: mining and milling; conversion and enrichment; fuel fabrication; spent fuel reprocessing; waste handling and disposal; and transportation of nuclear materials

  7. D-D tokamak reactor assessment

    International Nuclear Information System (INIS)

    Baxter, D.C.; Dabiri, A.E.

    1983-01-01

    A quantitative comparison of the physics and technology requirements, and the cost and safety performance of a d-d tokamak relative to a d-t tokamak has been performed. The first wall/blanket and energy recovery cycle for the d-d tokamak is simpler, and has a higher efficiency than the d-t tokamak. In most other technology areas (such as magnets, RF, vacuum, etc.) d-d requirements are more severe and the systems are more complex, expensive and may involve higher technical risk than d-t tokamak systems. Tritium technology for processing the plasma exhaust, and tritium refueling technology are required for d-d reactors, but no tritium containment around the blanket or heat transport system is needed. Cost studies show that for high plasma beta and high magnetic field the cost of electricity from d-d and d-t tokamaks is comparable. Safety analysis shows less radioactivity in a d-d reactor but larger amounts of stored energy and thus higher potential for energy release. Consequences of all postulated d-d accidents are significantly smaller than those from d-t reactor tritium releases

  8. Complementary Strategies for Directed C(sp3 )-H Functionalization: A Comparison of Transition-Metal-Catalyzed Activation, Hydrogen Atom Transfer, and Carbene/Nitrene Transfer.

    Science.gov (United States)

    Chu, John C K; Rovis, Tomislav

    2018-01-02

    The functionalization of C(sp 3 )-H bonds streamlines chemical synthesis by allowing the use of simple molecules and providing novel synthetic disconnections. Intensive recent efforts in the development of new reactions based on C-H functionalization have led to its wider adoption across a range of research areas. This Review discusses the strengths and weaknesses of three main approaches: transition-metal-catalyzed C-H activation, 1,n-hydrogen atom transfer, and transition-metal-catalyzed carbene/nitrene transfer, for the directed functionalization of unactivated C(sp 3 )-H bonds. For each strategy, the scope, the reactivity of different C-H bonds, the position of the reacting C-H bonds relative to the directing group, and stereochemical outcomes are illustrated with examples in the literature. The aim of this Review is to provide guidance for the use of C-H functionalization reactions and inspire future research in this area. © 2018 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

  9. Environmental assessment for the deactivation of the N Reactor facilities. Revision 1

    International Nuclear Information System (INIS)

    1994-11-01

    This environmental assessment (EA) provides information for the US Department of Energy (DOE) to decide whether the Proposed Action for the N Reactor facilities warrants a Finding of No Significant Impact or requires the preparation of an environmental impact statement (EIS). The EA describes current conditions at the N Reactor facilities, the need to take action at the facilities, the elements of the Proposed Action and alternatives, and the potential environmental impacts. The N Reactor facilities are currently in a surveillance and maintenance program, and will eventually be decontaminated and decommissioned (D and D). Operation and maintenance of the facilities resulted in conditions that could adversely impact human health or the environment if left as is until final D and D. The Proposed Action would deactivate the facilities to remove the conditions that present a potential threat to human health and the environment and to reduce surveillance and maintenance requirements. The action would include surveillance and maintenance after deactivation. Deactivation would take about three years and would involve about 80 facilities. Surveillance and maintenance would continue until final D and D, which is expected to be complete for all facilities except the N Reactor itself by the year 2018

  10. The Assessment Of High Temperature Reactor Fuel (Characteristics Of HTTR Fuel)

    International Nuclear Information System (INIS)

    Dewita, Erlan; Tuka, Veronica; Gunandjar

    1996-01-01

    HTTR is one of the reactor type with Helium coolant and outlet coolant temperature of 950 o C. One possibility of HTTR application is the coo generation of steam in high temperature and electric power for supply energy to industry in the future. Considering to the high operating temperature of HTTR, therefore it is needed the reactor fuel which have good mechanical, chemical and physical stability to the high temperature, and stable to the influence of fission fragment and neutron during irradiation. This assessment of the HTTR fuel characteristic based on the experiment data to find information of HTTR operation feasibility. Result of the assessment indicated that fission gas release at burn-up of 3.6 % FIMA which was the same as the maximum burn up in the HTTR design was fairly lower than the maximum release estimated in the design (5 x 10 - 4), which is R/B from the fuel fabricated by the prismatic block fuel method would be low (between 10 - 9 dan 10 - 8)

  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. Report on Reactor Physics Assessment of Candidate Accident Tolerant Fuel Cladding Materials in LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); George, Nathan [Univ. of Tennessee, Knoxville, TN (United States); Maldonado, G. Ivan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-08-28

    This work focuses on ATF concepts being researched at Oak Ridge National Laboratory (ORNL), expanding on previous studies of using alternate cladding materials in pressurized water reactors (PWRs). The neutronic performance of two leading alternate cladding materials were assessed in boiling water reactors (BWRs): iron-chromium-aluminum (FeCrAl) cladding, and silicon carbide (SiC)-based composite cladding. This report fulfills ORNL Milestone M3FT-15OR0202332 within the fiscal year 2015 (FY15)

  13. Safety assessment to support NUE fuel full core implementation in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Fan, H.Z.; Laurie, T.; Siddiqi, A.; Li, Z.P.; Rouben, D.; Zhu, W.; Lau, V.; Cottrell, C.M. [CANDU Energy Inc., Mississauga, Ontario (Canada)

    2013-07-01

    The Natural Uranium Equivalent (NUE) fuel contains a combination of recycled uranium and depleted uranium, in such a manner that the resulting mixture is similar to the natural uranium currently used in CANDU® reactors. Based on successful preliminary results of 24 bundles of NUE fuel demonstration irradiation in Qinshan CANDU 6 Unit 1, the NUE full core implementation program has been developed in cooperation with the Third Qinshan Nuclear Power Company and Candu Energy Inc, which has recently received Chinese government policy and funding support from their National-Level Energy Innovation program. This paper presents the safety assessment results to technically support NUE fuel full core implementation in CANDU reactors. (author)

  14. Comparison of methodologies for assessing the risks from nuclear weapons and from nuclear reactors

    International Nuclear Information System (INIS)

    Benjamin, A.S.

    1996-01-01

    There are important differences between the safety principles for nuclear weapons and for nuclear reactors. For example, a principal concern for nuclear weapons is to prevent electrical energy from reaching the nuclear package during accidents produced by crashes, fires, and other hazards, whereas the foremost concern for nuclear reactors is to maintain coolant around the core in the event of certain system failures. Not surprisingly, new methods have had to be developed to assess the risk from nuclear weapons. These include fault tree transformations that accommodate time dependencies, thermal and structural analysis techniques that are fast and unconditionally stable, and parameter sampling methods that incorporate intelligent searching. This paper provides an overview of the new methods for nuclear weapons and compares them with existing methods for nuclear reactors. It also presents a new intelligent searching process for identifying potential nuclear detonation vulnerabilities. The new searching technique runs very rapidly on a workstation and shows promise for providing an accurate assessment of potential vulnerabilities with far fewer physical response calculations than would be required using a standard Monte Carlo sampling procedure

  15. Analysis of dynamic stability and safety of reactor system by reactor simulator

    International Nuclear Information System (INIS)

    Raisic, N.

    1963-11-01

    In order to enable qualitative analysis of dynamic properties of reactors RA and RB, mathematical models of these reactors were formulated and adapted for solution on analog computer. This report contains basic assessments for creating the model and complete equations for each reactor. Model was used to analyse three possible accidents at the RA reactor and possible hypothetical accidents at the RB reactor

  16. The assessment of technological and safety aspects of small power reactor SMART

    International Nuclear Information System (INIS)

    Antariksawan, A.R.; Ekariansyah, Andi S.; Sony, D.T.; Suharno; Hastowo, Hudi

    2002-01-01

    This paper describes and discusses the technology and safety of small nuclear power plant SMART. The reactor SMART produces 300 MWth of power is cooled and moderated with light water and integral PWR type developed by KAERI. At present, the development activities had reached the end of basic design stage. The concept design of reactor SMART is based on safety enhancement, economic competitiveness and high performance. The fuel is uranium oxide with approximately 5% w/o enrichment. The safety characteristics of the core are shown with low power density around 62.6 W/cc, high negative reactivity coefficient, and high shutdown and thermal margin. Besides the inherent safety characteristics, SMART is equipped with engineered safety features and severe accident management system which are in compliance with the IAEA recommendations. The application of SMART for dual-purpose produces 90 Mwe and 40,000 to fresh water a day. Based on the technology and core characteristics of the reactor SMART, it is very interesting to be deeply assessed

  17. Approaches to Assess Competitiveness of Small and Medium Sized Reactors

    International Nuclear Information System (INIS)

    Kuznetsov, V.; Barkatullah, N.

    2011-01-01

    investment related factors affecting SMR competitiveness, as well as the tools available to assess these factors. In parallel with this, IAEA is conducting a series of case studies on SMR competitive deployment and application, making use of these tools and fostering their further development. The case studies that would highlight comparative benefits and disadvantages of SMRs versus larger reactors for different deployment strategies and application conditions, will be finalized in the end of 2009, and their final results are not yet available Therefore, the present paper summarizes the major outputs of the report titled 'Approaches to Assess Competitiveness of SMRs' and discusses approaches that need to be taken into account in the assessment rather than the results of their application. However, several illustrations of the application of such approaches are provided. (author)

  18. Risk assessment of a fusion-reactor fuel-processing system

    International Nuclear Information System (INIS)

    Bruske, S.Z.; Holland, D.F.

    1983-07-01

    The probabilistic risk assessment (PRA) methodology provides a means to systematically examine the potential for accidents that may result in a release of hazardous materials. This report presents the PRA for a typical fusion reactor fuel processing system. The system used in the analysis is based on the Tritium Systems Test Assembly, which is being operated at the Los Alamos National Laboratory. The results of the evaluation are presented in a probability-consequence plot that describes the probability of various accidental tritium release magnitudes

  19. A Preliminary Assessment of the Adjuster Rod Depletion Effect in the CANDU Reactor

    International Nuclear Information System (INIS)

    Kim, Yonghee; Roh, Gyuhong; Kim, Won Young; Kim, Hak Sung; Park, Joo Hwan

    2008-01-01

    Lifetime of the Wolsong-1 CANDU reactor, which will be shutdown in April, 2009. Major reactor components such as the pressure tube are to be replaced and it is expected that the CANDU reactor can be operated for additional 25-30 years. Meanwhile, all the reactivity devices including the adjuster rods (ADJ) are supposed to be continuously used without any change. In the CANDU reactor, 21 stainless steel (SS) ADJs are used to control the core power distribution and compensate for some reactivity loss during several abnormal cases. The ADJs are normally fully inserted and the SS absorber should undergo a slow depletion through neutron irradiation for a long time. In April, 2009, the accumulated FPY (Full Power Day) of Wolsong-1 is about 23 years. Depletion of ADJs should result in a smaller ADJ worth and a higher fuel burnup and the core power distribution should also be affected by the ADJ depletion. In this work, the effects of the ADJ depletion have been assessed in terms of ADJ worth, time-average core characteristics

  20. 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)

  1. Towards harmonised self assessment of research reactor safety status in operating organisations

    International Nuclear Information System (INIS)

    Kirchsteiger, C.; Boeck, H.

    2006-01-01

    The objective of this paper is to describe the development of a methodology and corresponding web-based tool for mapping and cross-comparing the safety approaches in European and other Research Reactor (RR) facilities in order to detect the principal similarities and differences. As an example, the performance of a Probabilistic Safety Assessment (PSA) for RRs is mapped, as follows: is PSA performed at all? (Yes/No); if so, is PSA mandatory or just recommended? (Yes/No); what is the scope of PSA?, its objective? and practical use? (set of more detailed questions), etc. In this way, information on different types of safety verification practices and requirements for RRs from Europe, Argentina, Australia, Canada, South Africa and the USA has been collected in a systematic way and included in the web-based benchmarking tool DARES (DAtabase for REsearch Reactor Safety). DARES has been developed and filled with sample data by the European Commission's Joint Research Centre (JRC) together with members of the European Research Reactors Operator Group (RROG). A systematic mapping by using DARES in parallel to an international Working Group, consisting of both operators and authorities could be the starting point towards harmonisation of RR safety verification on an international level. In addition, the availability of a user-friendly Information System on the Internet such as DARES containing this information is considered a useful mechanism to exchange international experiences and practices in the area among qualified users. This approach is currently considered to be proposed to the International Atomic Energy Agency (IAES) as one possible application of the recently adopted IAEA Code of Conduct on the Safety of Research Reactors. The resulting process would be a self-assessment of the RR safety status in regulatory bodies and operating organisations relative to the guidance in the Code, practically realised and monitored by an Information System similar to DARES. (orig.)

  2. Reactor engineering department annual report

    International Nuclear Information System (INIS)

    1990-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1989 (April 1, 1989 - March 31, 1990). One of major Department's programs is the assessment of the high conversion light water reactor and the design activities of advanced reactor system. Development of a high energy proton linear accelerator for the nuclear engineering including is also TRU incineration promoted. Other major tasks of the Department are various basic researches on nuclear data and group constants, theoretical methods and code development, on reactor physics experiments and analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  3. Assessment of nucleonic methods and data for fusion reactors

    International Nuclear Information System (INIS)

    Dudziak, D.J.

    1976-01-01

    An assessment is provided of nucleonic methods, codes, and data necessary for a sound experimental fusion power reactor (EPR) technology base. Gaps in the base are identified and specific development recommendations are made in three areas: computational tools, nuclear data, and integral experiments. The current status of the first two areas is found to be sufficiently inadequate that viable engineering design of an EPR is precluded at this time. However, a program to provide the necessary data and computational capability is judged to be a low-risk effort

  4. Experience with safety assessment of digital upgrading of IandC in VVER type reactors

    International Nuclear Information System (INIS)

    Wach, D.; Mulka, B.; Schnuerer, G.

    1997-01-01

    The digital upgrading of IandC systems important to safety in WWER type reactors requires a broad expertise in various knowledge fields. The approach of the Institute for safety Technology to the qualification and categorization of safety-critical software systems is highlighted. The role of the Institute in the qualification of the Teleperm XS and the type testing of its components is described. The aspects of the safety assessment of digital IandC systems in WWER type reactors is discussed in some detail. (A.K.)

  5. Assessment of the Zaporizhya NPP unit 1 reactor pressure vessel safety

    International Nuclear Information System (INIS)

    Podkopaev, V.; Popov, V.; Zaritsky, N.

    1997-01-01

    This emergency situation had occurred at the ZNPP unit 1 while its being under ''hot shutdown'' in natural coolant circulation mode. The main difference between emergency situation and mode with improper setting of PPPD described in the ''Technical Safety Substantiation (TSS) is that this mode is being considered in the TSS under rated power of reactor with main circulation pumps (MCP) under operation. This difference is a substantial one. For this reason a necessity appeared to asses an integrity of referred reactor pressure vessel (RPV) under given emergency situation to judge whether results obtained meet the ND requirements (safety assessment). Under operation such RPV elements are being mostly affected as upper cooling, lower cowling, weld No. 3 weld No. 4 situated in front of core. These elements materials ageing process is the most intense one. Thus, this work was aimed at investigation of structure material behavior and RPV integrity assessment under thermal shock conditions while PPPD improper setting. At that time the most attention was drawn to above mentioned upper and lower cowlings along with welds No. 3 and 4. 5 refs, figs, 10 tabs

  6. Assessment of the Zaporizhya NPP unit 1 reactor pressure vessel safety

    Energy Technology Data Exchange (ETDEWEB)

    Podkopaev, V; Popov, V; Zaritsky, N [State Scientific and Technical Centre on Nuclear and Radiation Safety (SSTC NRS), Kiev (Ukraine)

    1997-09-01

    This emergency situation had occurred at the ZNPP unit 1 while its being under ``hot shutdown`` in natural coolant circulation mode. The main difference between emergency situation and mode with improper setting of PPPD described in the ``Technical Safety Substantiation (TSS) is that this mode is being considered in the TSS under rated power of reactor with main circulation pumps (MCP) under operation. This difference is a substantial one. For this reason a necessity appeared to asses an integrity of referred reactor pressure vessel (RPV) under given emergency situation to judge whether results obtained meet the ND requirements (safety assessment). Under operation such RPV elements are being mostly affected as upper cooling, lower cowling, weld No. 3 weld No. 4 situated in front of core. These elements materials ageing process is the most intense one. Thus, this work was aimed at investigation of structure material behavior and RPV integrity assessment under thermal shock conditions while PPPD improper setting. At that time the most attention was drawn to above mentioned upper and lower cowlings along with welds No. 3 and 4. 5 refs, figs, 10 tabs.

  7. Self Assessment for the Safety of Research Reactor in Indonesia

    International Nuclear Information System (INIS)

    Melani, Ai; Chang, Soon Heung

    2008-01-01

    At the present Indonesia has no nuclear power plant in operation yet, although it is expected that the first nuclear power plant will be operated and commercially available in around the year of 2016 to 2017 in Muria Peninsula. National Nuclear Energy Agency (BATAN) has three research reactor; which are: Reactor Triga Mark II at Bandung, Reactor Kartini at Yogyakarta and Reactor Serbaguna (Multi Purpose Reactor) at Serpong. The Code of Conduct on the Safety of Research Reactors establishes 'best practice' guidelines for the licensing, construction and operation of research reactors. In this paper the author use the requirement in code of conduct to review the safety of research reactor in Indonesia

  8. Structural assessments of plate type support system for APR1400 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nguyen, Anh Tung; Namgung, Ihn, E-mail: inamgung@kings.ac.kr

    2017-04-01

    Highlights: • This paper investigates plate-type support structure for the reactor vessel of the APR 1400. • The tall column supports of APR1400 reactor challenges in seismic and severe accident events. • A plate-type support of reactor vessel was proposed and evaluated based on ASME code. • The plate-type support was assessed to show its higher rigidity than column-type. - Abstract: This paper investigates an alternative form of support structure for the reactor vessel of the APR 1400. The current reactor vessel adopts a four-column support arrangement locating on the cold legs of the vessel. Although having been successfully designed, the tall column structure challenges in seismic events. In addition, for the mitigation of severe accident consequences, the columns inhibit ex-vessel coolant flow, hence the elimination of the support columns proposes extra safety advantages. A plate-type support was proposed and evaluated for the adequacy of meeting the structural stiffness by Finite Element Analysis (FEA) approach. ASME Boiler and Pressure Vessel Code was used to verify the design. The results, which cover thermal and static structural analysis, show stresses are within allowable limits in accordance with the design code. Even the heat conduction area is increased for the plate-type of support system, the results showed that the thermal stresses are within allowable limits. A comparison of natural frequencies and mode shapes for column support and plate-type support were presented as well which showed higher fundamental frequencies for the plate-type support system resulting in greater rigidity of the support system. From the outcome of this research, the plate-type support is proven to be an alternative to current APR column type support design.

  9. Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR)

    International Nuclear Information System (INIS)

    Gilbert, B.G.; Reece, W.J.; Gertman, D.I.; Gilmore, W.E.; Galyean, W.J.

    1990-12-01

    The Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) is an automated data base management system for processing and storing human error probability and hardware component failure data. The NUCLARR system software resides on an IBM (or compatible) personal computer. NUCLARR can furnish the end user with data inputs for both human and hardware reliability analysis in support of a variety of risk assessment activities. The NUCLARR system is documented in a five-volume series of reports. Volume 5: Data Manual provides a hard-copy representation of all data and related information available within the NUCLARR system software. This document is organized in three sections. Part 1 is the summary description, which presents an overview of the NUCLARR system and data processing procedures. Part 2 contains all data and information relevant to the human error probability (HEP) data side of NUCLARR. Data and information for the hardware component failure data (HCFD) side are presented in Part 3. 7 refs

  10. Initial Investigation into the Potential of CSP Industrial Process Heat for the Southwest United States

    Energy Technology Data Exchange (ETDEWEB)

    Kurup, Parthiv [National Renewable Energy Lab. (NREL), Golden, CO (United States); Turchi, Craig [National Renewable Energy Lab. (NREL), Golden, CO (United States)

    2015-11-01

    After significant interest in the 1970s, but relatively few deployments, the use of solar technologies for thermal applications, including enhanced oil recovery (EOR), desalination, and industrial process heat (IPH), is again receiving global interest. In particular, the European Union (EU) has been a leader in the use, development, deployment, and tracking of Solar Industrial Process Heat (SIPH) plants. The objective of this study is to ascertain U.S. market potential of IPH for concentrating collector technologies that have been developed and promoted through the U.S. Department of Energy's Concentrating Solar Power (CSP) Program. For this study, the solar-thermal collector technologies of interest are parabolic trough collectors (PTCs) and linear Fresnel (LF) systems.

  11. Good Practices for Water Quality Management in Research Reactors and Spent Fuel Storage Facilities

    International Nuclear Information System (INIS)

    2011-01-01

    for water chemistry control in research reactors; defines parameters recommended, techniques applicable, sampling procedures and sampling frequency to monitor water quality in RRs, and describes the importance of a quality assurance programme, and the implementation of a corrosion surveillance programme (CSP) as part of the water management programme. Whenever applicable, considerations are made for primary cooling system, spent fuel storage basins, secondary cooling system, emergency cooling systems, make-up systems and water reservoirs of RRs.

  12. Assessment and mitigation of power quality problems for PUSPATI TRIGA Reactor (RTP)

    Science.gov (United States)

    Zakaria, Mohd Fazli; Ramachandaramurthy, Vigna K.

    2017-01-01

    An electrical power systems are exposed to different types of power quality disturbances. Investigation and monitoring of power quality are necessary to maintain accurate operation of sensitive equipment especially for nuclear installations. This paper will discuss the power quality problems observed at the electrical sources of PUSPATI TRIGA Reactor (RTP). Assessment of power quality requires the identification of any anomalous behavior on a power system, which adversely affects the normal operation of electrical or electronic equipment. A power quality assessment involves gathering data resources; analyzing the data (with reference to power quality standards) then, if problems exist, recommendation of mitigation techniques must be considered. Field power quality data is collected by power quality recorder and analyzed with reference to power quality standards. Normally the electrical power is supplied to the RTP via two sources in order to keep a good reliability where each of them is designed to carry the full load. The assessment of power quality during reactor operation was performed for both electrical sources. There were several disturbances such as voltage harmonics and flicker that exceeded the thresholds. To reduce these disturbances, mitigation techniques have been proposed, such as to install passive harmonic filters to reduce harmonic distortion, dynamic voltage restorer (DVR) to reduce voltage disturbances and isolate all sensitive and critical loads.

  13. Solving the Container Stowage Problem (CSP) using Particle Swarm Optimization (PSO)

    Science.gov (United States)

    Matsaini; Santosa, Budi

    2018-04-01

    Container Stowage Problem (CSP) is a problem of containers arrangement into ships by considering rules such as: total weight, weight of one stack, destination, equilibrium, and placement of containers on vessel. Container stowage problem is combinatorial problem and hard to solve with enumeration technique. It is an NP-Hard Problem. Therefore, to find a solution, metaheuristics is preferred. The objective of solving the problem is to minimize the amount of shifting such that the unloading time is minimized. Particle Swarm Optimization (PSO) is proposed to solve the problem. The implementation of PSO is combined with some steps which are stack position change rules, stack changes based on destination, and stack changes based on the weight type of the stacks (light, medium, and heavy). The proposed method was applied on five different cases. The results were compared to Bee Swarm Optimization (BSO) and heuristics method. PSO provided mean of 0.87% gap and time gap of 60 second. While BSO provided mean of 2,98% gap and 459,6 second to the heuristcs.

  14. An assessment of methods of calculating Doppler effects in plutonium fuelled sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Butland, A.T.D.; Reddell, G.

    1979-01-01

    After a survey of the requirements, an assessment of UK methods and data is made on the basis of the following work. First, the analysis of the SEFOR Doppler experiments, carried out using the UK FGL5 fine group nuclear data library, the MURAL cell code and whole reactor diffusion theory calculations of the neutron flux. Second, the analysis of some Japanese FCA central sample perturbation measurements of structural material Doppler effects. Third, an assessment of the accuracy of Doppler predictions in a sodium voided core using results from Zebra 5 and BIZET, and theoretical studies of additional effects relevant to power reactors and not covered by the above analyses, including the following, the calculation of Doppler effects at high temperature, fuel cycle and burn-up effects, and the heterogeneity effects of large fuelled subassemblies in pin geometry. The importance of crystalline binding effects in the fuel are discussed as is the importance of reactor material boundaries in the calculation of resonance shielding effects. Some suggestions for further Doppler studies are made. (U.K.)

  15. Comparative life cycle assessment of real pilot reactors for microalgae cultivation in different seasons

    International Nuclear Information System (INIS)

    Pérez-López, Paula; De Vree, Jeroen H.; Feijoo, Gumersindo; Bosma, Rouke; Barbosa, Maria J.; Moreira, María Teresa; Wijffels, René H.; Van Boxtel, Anton J.B.; Kleinegris, Dorinde M.M.

    2017-01-01

    Highlights: •Life cycle assessment was used to compare 3 real pilot systems for algae cultivation. •The temperature control system was the main contributor to environmental impacts. •Tubular reactors had lower impacts per unit of biomass produced than open pond. •Meteorological conditions on the reactors played a critical role in LCA results. •Environmental impact reductions of 17–90% were estimated for optimized full-scale reactors. -- Abstract: Microalgae are promising natural resources for biofuels, chemical, food and feed products. Besides their economic potential, the environmental sustainability must be examined. Cultivation has a significant environmental impact that depends on reactor selection and operating conditions. To identify the main environmental bottlenecks for scale-up to industrial facilities this study provides a comparative life cycle assessment (LCA) of open raceway ponds and tubular photobioreactors at pilot scale. The results are based on experimental data from real pilot plants operated in summer, fall and winter at AlgaePARC (Wageningen, The Netherlands). The energy consumption for temperature regulation presented the highest environmental burden. The production of nutrients affected some categories. Despite limited differences compared to the vertical system, the horizontal PBR was found the most efficient in terms of productivity and environmental impact. The ORP was, given the Dutch climatic conditions, only feasible under summer operation. The results highlight the relevance of LCA as a tool for decision-making in process design. Weather conditions and availability of sources for temperature regulation were identified as essential factors for the selection of geographic locations and for microalgal cultivation systems based on environmental criteria. Simulation of large-scale reactors with optimized temperature regulation systems lead to environmental improvements and energy demand reductions ranging from 17% up to 90% for

  16. An assessment of the optimal timing and size of investments in concentrated solar power

    International Nuclear Information System (INIS)

    Massetti, Emanuele; Ricci, Elena Claire

    2013-01-01

    We extend the WITCH model to consider the possibility to produce and trade electricity generated by large-scale concentrated solar power plants (CSP) in highly productive areas that are connected to demand centers through High Voltage Direct Current cables. We test the attractiveness of the CSP option by imposing a global cap on Greenhouse gases concentration equal to 535 ppm CO 2 -eq in 2100, with and without constraints to the expansion of nuclear power and IGCC coal with carbon capture and storage (CCS). We find that it becomes optimal to produce with CSP from 2040 and to trade CSP electricity across the Mediterranean from 2050. Therefore projects like DESERTEC seem to be premature. After 2050, CSP electricity shares become significant. CSP has a high stabilization cost option value: depending on the constraints, it ranges between 2.1% and 4.1% of discounted GDP in the Middle East and North Africa (MENA), between 1.1. and 3.4 in China, between 0.2% and 1.2% in the USA, between 0.1 and 1.3% in Eastern Europe and between 0.1 and 0.4% in Western Europe. A moderate level of subsidy to invest more and earlier in CSP might increase welfare. However, large-scale deployment should occur after 2040. We also show that MENA countries have the incentive to form a cartel to sell electricity to Europe at a price higher than the marginal cost. This suggests that a hypothetical Mediterranean market for electricity should be carefully regulated. - Highlights: ► An extensive use of Concentrated Solar Power (CSP) will be optimal after 2050. ► Trade of CSP electricity between MENA and Europe will start in 2050. ► CSP reduces greatly the option value of nuclear power and coal with CCS. ► Learning externalities motivate moderate subsidies for earlier CSP investments. ► MENA countries have the incentive to form a cartel to sell electricity to Europe

  17. Performance Assessment of Turbulence Models for the Prediction of the Reactor Internal Flow in the Scale-down APR+

    International Nuclear Information System (INIS)

    Lee, Gonghee; Bang, Youngseok; Woo, Swengwoong; Kim, Dohyeong; Kang, Minku

    2013-01-01

    The types of errors in CFD simulation can be divided into the two main categories: numerical errors and model errors. Turbulence model is one of the important sources for model errors. In this study, in order to assess the prediction performance of Reynolds-averaged Navier-Stokes (RANS)-based two equations turbulence models for the analysis of flow distribution inside a 1/5 scale-down APR+, the simulation was conducted with the commercial CFD software, ANSYS CFX V. 14. In this study, in order to assess the prediction performance of turbulence models for the analysis of flow distribution inside a 1/5 scale-down APR+, the simulation was conducted with the commercial CFD software, ANSYS CFX V. 14. Both standard k-ε model and SST model predicted the similar flow pattern inside reactor. Therefore it was concluded that the prediction performance of both turbulence models was nearly same. Complex thermal-hydraulic characteristics exist inside reactor because the reactor internals consist of fuel assembly, control rod assembly, and the internal structures. Either flow distribution test for the scale-down reactor model or computational fluid dynamics (CFD) simulation have been conducted to understand these complex thermal-hydraulic features inside reactor

  18. Solar Thermoelectricity via Advanced Latent Heat Storage: A Cost-Effective Small-Scale CSP Application

    Energy Technology Data Exchange (ETDEWEB)

    Glatzmaier, Greg C.; Rea, J.; Olsen, Michele L.; Oshman, C.; Hardin, C.; Alleman, Jeff; Sharp, J.; Weigand, R.; Campo, D.; Hoeschele, G.; Parilla, Philip A.; Siegel, N. P.; Toberer, Eric S.; Ginley, David S.

    2017-06-27

    We are developing a novel concentrating solar electricity-generating technology that is both modular and dispatchable. Solar ThermoElectricity via Advanced Latent heat Storage (STEALS) uses concentrated solar flux to generate high-temperature thermal energy, which directly converts to electricity via thermoelectric generators (TEGs), stored within a phase-change material (PCM) for electricity generation at a later time, or both allowing for simultaneous charging of the PCM and electricity generation. STEALS has inherent features that drive its cost-competitive scale to be much smaller than current commercial concentrating solar power (CSP) plants. Most obvious is modularity of the solid-state TEG, which favors smaller scales in the kilowatt range as compared to CSP steam turbines, which are minimally 50 MWe for commercial power plants. Here, we present techno-economic and market analyses that show STEALS can be a cost-effective electricity-generating technology with particular appeal to small-scale microgrid applications. We evaluated levelized cost of energy (LCOE) for STEALS and for a comparable photovoltaic (PV) system with battery storage. For STEALS, we estimated capital costs and the LCOE as functions of the type of PCM including the use of recycled aluminum alloys, and evaluated the cost tradeoffs between plasma spray coatings and solution-based boron coatings that are applied to the wetted surfaces of the PCM subsystem. We developed a probabilistic cost model that accounts for uncertainties in the cost and performance inputs to the LCOE estimation. Our probabilistic model estimated LCOE for a 100-kWe STEALS system that had 5 hours of thermal storage and 8-10 hours of total daily power generation. For these cases, the solar multiple for the heliostat field varied between 1.12 and 1.5. We identified microgrids as a likely market for the STEALS system. We characterized microgrid markets in terms of nominal power, dispatchability, geographic location, and

  19. Solar thermoelectricity via advanced latent heat storage: A cost-effective small-scale CSP application

    Science.gov (United States)

    Glatzmaier, G. C.; Rea, J.; Olsen, M. L.; Oshman, C.; Hardin, C.; Alleman, J.; Sharp, J.; Weigand, R.; Campo, D.; Hoeschele, G.; Parilla, P. A.; Siegel, N. P.; Toberer, E. S.; Ginley, D. S.

    2017-06-01

    We are developing a novel concentrating solar electricity-generating technology that is both modular and dispatchable. Solar ThermoElectricity via Advanced Latent heat Storage (STEALS) uses concentrated solar flux to generate high-temperature thermal energy, which directly converts to electricity via thermoelectric generators (TEGs), stored within a phase-change material (PCM) for electricity generation at a later time, or both allowing for simultaneous charging of the PCM and electricity generation. STEALS has inherent features that drive its cost-competitive scale to be much smaller than current commercial concentrating solar power (CSP) plants. Most obvious is modularity of the solid-state TEG, which favors smaller scales in the kilowatt range as compared to CSP steam turbines, which are minimally 50 MWe for commercial power plants. Here, we present techno-economic and market analyses that show STEALS can be a cost-effective electricity-generating technology with particular appeal to small-scale microgrid applications. We evaluated levelized cost of energy (LCOE) for STEALS and for a comparable photovoltaic (PV) system with battery storage. For STEALS, we estimated capital costs and the LCOE as functions of the type of PCM including the use of recycled aluminum alloys, and evaluated the cost tradeoffs between plasma spray coatings and solution-based boron coatings that are applied to the wetted surfaces of the PCM subsystem. We developed a probabilistic cost model that accounts for uncertainties in the cost and performance inputs to the LCOE estimation. Our probabilistic model estimated LCOE for a 100-kWe STEALS system that had 5 hours of thermal storage and 8-10 hours of total daily power generation. For these cases, the solar multiple for the heliostat field varied between 1.12 and 1.5. We identified microgrids as a likely market for the STEALS system. We characterized microgrid markets in terms of nominal power, dispatchability, geographic location, and

  20. Assessment of the Dry Processed Oxide Fuel in Liquid Metal Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Gyu Hong; Choi, Hang Bok

    2005-09-15

    The neutronic feasibility of the dry process oxide fuel was assessed for the sodium-cooled and lead-cooled fast reactors (SFR and LFR, respectively), which were recommended as Generation-IV (Gen-IV) reactor systems by the Gen-IV international forum. The reactor analysis was performed for the equilibrium fuel cycle of two core configurations: Hybrid BN-600 benchmark core with an enlarged lattice pitch and a modified BN-600 core. The dry process technology assumed in this study is the molten-salt process, which was developed by Russian scientists for recycling oxide fuels. The core calculation was performed by the REBUS-3 code and the reactor characteristics such as the transuranic (TRU) enrichment, breeding ratio, peak linear power, burnup reactivity swing, etc. were calculated for the equilibrium core under a fixed fuel management scheme. The results showed that a fissile self-sustainable breakeven core was achievable without blanket fuels when the fuel volume fraction was {approx}50% and most of the fission products were removed. If the design criteria used in this study is proved to be acceptable through a detailed physics design and thermal hydraulic analysis in the future, it is practically possible to construct an equilibrium fuel cycle of the SFR and LFR systems based on the oxide fuel by utilizing the dry process technology.

  1. Assessment of the Dry Processed Oxide Fuel in Liquid Metal Fast Reactors

    International Nuclear Information System (INIS)

    Roh, Gyu Hong; Choi, Hang Bok

    2005-09-01

    The neutronic feasibility of the dry process oxide fuel was assessed for the sodium-cooled and lead-cooled fast reactors (SFR and LFR, respectively), which were recommended as Generation-IV (Gen-IV) reactor systems by the Gen-IV international forum. The reactor analysis was performed for the equilibrium fuel cycle of two core configurations: Hybrid BN-600 benchmark core with an enlarged lattice pitch and a modified BN-600 core. The dry process technology assumed in this study is the molten-salt process, which was developed by Russian scientists for recycling oxide fuels. The core calculation was performed by the REBUS-3 code and the reactor characteristics such as the transuranic (TRU) enrichment, breeding ratio, peak linear power, burnup reactivity swing, etc. were calculated for the equilibrium core under a fixed fuel management scheme. The results showed that a fissile self-sustainable breakeven core was achievable without blanket fuels when the fuel volume fraction was ∼50% and most of the fission products were removed. If the design criteria used in this study is proved to be acceptable through a detailed physics design and thermal hydraulic analysis in the future, it is practically possible to construct an equilibrium fuel cycle of the SFR and LFR systems based on the oxide fuel by utilizing the dry process technology

  2. The radiation safety assessment of the heating loop of district heating reactors

    International Nuclear Information System (INIS)

    Liu Yuanzhong

    1993-01-01

    The district heating reactors are used to supply heating to the houses in cities. The concerned problems are whether the radioactive materials reach the heated houses through heating loop, and whether the safety of the dwellers can be ensured. In order to prevent radioactive materials getting into the heated houses, the district heating reactors have three loops, namely, primary loop, intermediate loop, and heating loop. In the paper, the measures of preventing radioactive materials getting into the heating loop are presented, and the possible sources of the radioactivity in the water of the intermediate loop and the heating loop are given. The regulatory aim limit of radioactive concentration in the water of the intermediate loop is put forward, which is 18.5 Bq/l. Assuming that specific radioactivity of the water of contaminated intermediate loop is up to 18.5 Bq/l, the maximum concentration of radionuclides in water of the heating loop is calculated for the normal operation and the accident of district heating reactor. The results show that the maximum possible concentration is 5.7 x 10 -3 Bq/l. The radiation safety assessment of the heating loop is made out. The conclusions are that the district heating reactors do not bring any harmful impact to the dwellers, and the safety of the dwellers can be safeguarded completely

  3. Reactor safety study. An assessment of accident risks in U.S. commercial nuclear power plants. Executive summary: main report

    International Nuclear Information System (INIS)

    1975-10-01

    Information is presented concerning the objectives and organization of the reactor safety study; the basic concepts of risk; the nature of nuclear power plant accidents; risk assessment methodology; reactor accident risk; and comparison of nuclear risks to other societal risks

  4. Development of System Model for Level 1 Probabilistic Safety Assessment of TRIGA PUSPATI Reactor

    International Nuclear Information System (INIS)

    Tom, P.P; Mazleha Maskin; Ahmad Hassan Sallehudin Mohd Sarif; Faizal Mohamed; Mohd Fazli Zakaria; Shaharum Ramli; Muhamad Puad Abu

    2014-01-01

    Nuclear safety is a very big issue in the world. As a consequence of the accident at Fukushima, Japan, most of the reactors in the world have been reviewed their safety of the reactors including also research reactors. To develop Level 1 Probabilistic Safety Assessment (PSA) of TRIGA PUSPATI Reactor (RTP), three organizations are involved; Nuclear Malaysia, AELB and UKM. PSA methodology is a logical, deductive technique which specifies an undesired top event and uses fault trees and event trees to model the various parallel and sequential combinations of failures that might lead to an undesired event. Fault Trees (FT) methodology is use in developing of system models. At the lowest level, the Basic Events (BE) of the fault trees (components failure and human errors) are assigned probability distributions. In this study, Risk Spectrum software used to construct the fault trees and analyze the system models. The results of system models analysis such as core damage frequency (CDF), minimum cut set (MCS) and common cause failure (CCF) uses to support decision making for upgrading or modification of the RTP?s safety system. (author)

  5. An assessment of methods of calculating sodium voiding reactivity in plutonium fuelled fast reactors

    International Nuclear Information System (INIS)

    Butland, A.T.D.; Simmons, W.N.; Stevenson, J.M.

    1979-01-01

    After a survey of the requirements an assessment of the accuracy of calculations of the sodium void effect using UK methods and data is made on the basis of the following work. First, the analysis of small and large sodium voids in the MOZART and Zebra 13 small (300 MW(E)) fast reactor mock-ups and the BIZET large fast reactor mock-ups, all of conventional design. The analysis was carried out using the UK FGL5 fine group nuclear data library, the MURAL cell code, whole reactor diffusion theory calculations of the neutron flux and perturbation theory methods. Exact perturbation theory was used in many cases, otherwise first order perturbation theory calculations were adjusted to give results equivalent to exact perturbation theory. Second, theoretical studies of some effects, including, the effects of extrapolating to fuel operating temperatures, fuel cycle and burn-up effects, and the heterogeneity effects of large fuelled subassemblies in pin geometry. Third, theoretical studies of approximations in the calculational methods including, the importance in the whole reactor calculation of the energy group structure and the spatial mesh, the importance of reactor material boundaries in the calculation of resonance shielding effects, and the use of neutron fluxes calculated using neutron diffusion theory rather than transport theory. (U.K.)

  6. Application of probabilistic risk assessment in nuclear and environmental licensing processes of nuclear reactors in Brazil

    International Nuclear Information System (INIS)

    Mata, Jonatas F.C. da; Vasconcelos, Vanderley de; Mesquita, Amir Z.

    2015-01-01

    The nuclear accident at Fukushima Daiichi, occurred in Japan in 2011, brought reflections, worldwide, on the management of nuclear and environmental licensing processes of existing nuclear reactors. One of the key lessons learned in this matter, is that the studies of Probabilistic Safety Assessment and Severe Accidents are becoming essential, even in the early stage of a nuclear development project. In Brazil, Brazilian Nuclear Energy Commission, CNEN, conducts the nuclear licensing. The organism responsible for the environmental licensing is Brazilian Institute of Environment and Renewable Natural Resources, IBAMA. In the scope of the licensing processes of these two institutions, the safety analysis is essentially deterministic, complemented by probabilistic studies. The Probabilistic Safety Assessment (PSA) is the study performed to evaluate the behavior of the nuclear reactor in a sequence of events that may lead to the melting of its core. It includes both probability and consequence estimation of these events, which are called Severe Accidents, allowing to obtain the risk assessment of the plant. Thus, the possible shortcomings in the design of systems are identified, providing basis for safety assessment and improving safety. During the environmental licensing, a Quantitative Risk Analysis (QRA), including probabilistic evaluations, is required in order to support the development of the Risk Analysis Study, the Risk Management Program and the Emergency Plan. This article aims to provide an overview of probabilistic risk assessment methodologies and their applications in nuclear and environmental licensing processes of nuclear reactors in Brazil. (author)

  7. Application of probabilistic risk assessment in nuclear and environmental licensing processes of nuclear reactors in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Mata, Jonatas F.C. da; Vasconcelos, Vanderley de; Mesquita, Amir Z., E-mail: jonatasfmata@yahoo.com.br, E-mail: vasconv@cdtn.br, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    The nuclear accident at Fukushima Daiichi, occurred in Japan in 2011, brought reflections, worldwide, on the management of nuclear and environmental licensing processes of existing nuclear reactors. One of the key lessons learned in this matter, is that the studies of Probabilistic Safety Assessment and Severe Accidents are becoming essential, even in the early stage of a nuclear development project. In Brazil, Brazilian Nuclear Energy Commission, CNEN, conducts the nuclear licensing. The organism responsible for the environmental licensing is Brazilian Institute of Environment and Renewable Natural Resources, IBAMA. In the scope of the licensing processes of these two institutions, the safety analysis is essentially deterministic, complemented by probabilistic studies. The Probabilistic Safety Assessment (PSA) is the study performed to evaluate the behavior of the nuclear reactor in a sequence of events that may lead to the melting of its core. It includes both probability and consequence estimation of these events, which are called Severe Accidents, allowing to obtain the risk assessment of the plant. Thus, the possible shortcomings in the design of systems are identified, providing basis for safety assessment and improving safety. During the environmental licensing, a Quantitative Risk Analysis (QRA), including probabilistic evaluations, is required in order to support the development of the Risk Analysis Study, the Risk Management Program and the Emergency Plan. This article aims to provide an overview of probabilistic risk assessment methodologies and their applications in nuclear and environmental licensing processes of nuclear reactors in Brazil. (author)

  8. Basic concept of fuel safety design and assessment for sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Nakae, Nobuo; Baba, Toshikazu; Kamimura, Katsuichiro

    2013-03-01

    'Philosophy in Safety Evaluation of Fast Breeder Reactors' was published as a guideline for safety design and safety evaluation of Sodium-Cooled Fast Reactor in Japan. This guideline points out that cladding creep and swelling due to internal pressure should be taken into account since the fuel is used under high temperature and high burnup, and that fuel assembly deformation and the prevention from coolant channel blockage should be taken into account in viewpoints of nuclear and thermal hydraulic design. However, the requirements including their criteria and evaluation items are not described. Two other domestic guidelines related to core design are applied for fuel design of fast reactor, but the description is considered to not be enough to practically use. In addition, technical standard for nuclear fuel used in power reactors is also applied for fuel inspection. Therefore, the technical standard and guideline for fuel design and safety evaluation are considered to be very important issue for nuclear safety regulation. This document has been developed according to the following steps: The guidelines and the technical standards, which are prepared in foreign countries and international organization, were reviewed. The technical background concerning fuel design and safety evaluation for fast reactor was collected and summarized in the world wide scale. The basic concept of fuel safety design and assessment for sodium-cooled fast reactor was developed by considering a wide range of views of the specialists in Japan. In order to discuss the content with foreign specialists IAEA Consultancy Meetings have been held on January, 2011 and January, 2012. The participants of the meeting came from USA, UK, EC, India, China and South Korea. The specialists of IAEA and JNES were also joined. Although this document is prepared for application to 'Monju'(prototype LMFR), it may be applied to experimental, demonstration and commercial types of LMFR after revising it by taking

  9. Recommendations concerning models and parameters best suited to breeder reactor environmental radiological assessments

    International Nuclear Information System (INIS)

    Miller, C.W.; Baes, C.F. III; Dunning, D.E. Jr.

    1980-05-01

    Recommendations are presented concerning the models and parameters best suited for assessing the impact of radionuclide releases to the environment by breeder reactor facilities. These recommendations are based on the model and parameter evaluations performed during this project to date. Seven different areas are covered in separate sections

  10. Temperature fluctuations: an assessment of their use in the detection of fast reactor coolant blockages

    International Nuclear Information System (INIS)

    Greef, C.P.

    1979-01-01

    The temperature noise technique for the detection of local blockages in fast reactor subassemblies is discussed. The main factors involved in an assessment of the technique are outlined and the experimental and theoretical work that has been carried out at BNL on the various aspects of the problem is described. It is concluded that blockings appreciably smaller than those predicted to produce boiling should be detectable against a background noise level due to subassembly power tilts, on a time scale giving protection against rapidly developing incidents. Further work required to increase confidence in the application of the technique to the reactor is outlined, including measurements in fully representative geometries, data from sodium rigs and further information on reactor background noise levels. (Auth.)

  11. Special Purpose Nuclear Reactor (5 MW) for Reliable Power at Remote Sites Assessment Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division; Werner, James Elmer [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division; McKellar, Michael George [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division; Hummel, Andrew John [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division; Kennedy, John Charles [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division; Wright, Richard Neil [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division; Biersdorf, John Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division

    2017-04-01

    The Phenomena Identification and Ranking Table (PIRT) technique was conducted on the Special Purpose Reactor nuclear plant design. The PIRT is a structured process to identify safety-relevant/safety-significant phenomena and assess the importance and knowledge base by ranking the phenomena. The Special Purpose Reactor is currently in the conceptual design stage. The candidate reactor has a solid monolithic stainless steel core with an array of heat pipes and fuel pellets embedded in the monolith. The heat pipes are used to remove heat from the core using simple, reliable, and well-characterized physics (capillarity, boiling, and condensation). In the initial design, one heat exchanger is used for the working fluid that produces energy, and a second heat exchanger is used to remove decay heat in emergency or shutdown conditions. In addition, a power conversion cycle such as an open-air Brayton system is available as an option for power conversion and process heat. This report summarizes and documents the process and scope of the four PIRT reviews, noting the major activities and conclusions. The identified phenomena, analyses, rationales, and associated ratings are presented along with a summary of the findings from the four individual PIRTs, namely (1) Reactor Accident and Normal Operations, (2) Heat Pipes, (3) Materials, and (4) Power Conversion. The PIRT reports for these four major system areas evaluated are attached as appendixes to this report and provide considerably more detail about each assessment as well as a more complete listing of the phenomena that were evaluated.

  12. Assessment of radiological releases to the environment from a fusion reactor power plant

    International Nuclear Information System (INIS)

    Shank, K.E.; Oakes, T.W.; Easterly, C.E.

    1978-05-01

    This report summarizes the expected tritium and activation-product inventories and presents an assessment of the potential radiological releases from a fusion reactor power plant, hypothetically located at the Oak Ridge National Laboratory. Routine tritium releases and the resulting dose assessment are discussed. Uncertainties associated with the conversion of tritium gas to tritium oxide and the global tritium cycling are evaluated. The difficulties of estimating releases of activated materials and the subsequent dose commitment are reviewed

  13. Fracture assessment of the Oskarshamn 1 reactor pressure vessel under cold over-pressurization

    International Nuclear Information System (INIS)

    Sattari-Far, I.

    2001-03-01

    The major motivation of this study was to develop a methodology for fracture assessment of surface defects in the 01 reactor pressure vessel under cold loading scenarios, particularly the cold over-pressurization event. According to a previous study, the FENIX project, the cold over-pressurization of the O1 reactor is a limiting loading case, as the ductile/brittle transition temperature (RT NDT ) of certain welds in the O1 beltline region may be over 100 deg C at the-end-of-life condition. The FENIX project gave values of the acceptable and critical crack depth to be equal to the thickness of the cladding layer (about 6 mm) under this load case using the ASME K Ic reference curve methodology. This study is aimed to develop a methodology to give a more precise fracture assessment of the O1 reactor under cold loading scenarios. Some of the main objectives of this study have been as below: To prepare a material which can simulate the mechanical properties and RT NDT of the O1 reactor at the end-of-life conditions. To conduct a fracture mechanics test program to cover the essential influencing factors, such as crack geometry (shallow and deep cracks) and loading condition (uniaxial and biaxial) on the cleavage fracture toughness. To perform fracture mechanics analyses to identify a suitable methodology for assessment of the experimental results. To study the responses of engineering fracture assessment methods to the experimental results from the clad specimens. To propose a fracture assessment procedure for determination of the acceptable and critical flaw sizes in the 01 reactor under the cold loading events. A test program consisted of experiments on standard SEN(B) specimens and clad beams, containing surface cracks was conducted during the course of this project. A total of nine clad beams and clad cruciform specimens were tested under uniaxial and biaxial loading. The test material is reactor steel of type A 508 Grade B, which is specially heat-treated to

  14. Use of Master Curve technology for assessing shallow flaws in a reactor pressure vessel material

    International Nuclear Information System (INIS)

    Bass, Bennett Richard; Taylor, Nigel

    2006-01-01

    In the NESC-IV project an experimental/analytical program was performed to develop validated analysis methods for transferring fracture toughness data to shallow flaws in reactor pressure vessels subject to biaxial loading in the lower-transition temperature region. Within this scope an extensive range of fracture tests was performed on material removed from a production-quality reactor pressure vessel. The Master Curve analysis of this data is reported and its application to the assessment of the project feature tests on large beam test pieces.

  15. Probabilistic safety assessment of Tehran Research Reactor using systems analysis programs for hands-on integrated reliability evaluations

    International Nuclear Information System (INIS)

    Hosseini, M.H.; Nematollahi, M.R.; Sepanloo, K.

    2004-01-01

    Probabilistic safety assessment application is found to be a practical tool for research reactor safety due to intense involvement of human interactions in an experimental facility. In this document the application of the probabilistic safety assessment to the Tehran Research Reactor is presented. The level 1 practicabilities safety assessment application involved: Familiarization with the plant, selection of accident initiators, mitigating functions and system definitions, event tree constructions and quantifications, fault tree constructions and quantification, human reliability, component failure data base development and dependent failure analysis. Each of the steps of the analysis given above is discussed with highlights from the selected results. Quantification of the constructed models is done using systems analysis programs for hands-on integrated reliability evaluations software

  16. Reactor safety

    International Nuclear Information System (INIS)

    Butz, H.P.; Heuser, F.W.; May, H.

    1985-01-01

    The paper comprises an introduction into nuclear physics bases, the safety concept generally speaking, safety devices of pwr type reactors, accident analysis, external influences, probabilistic safety assessment and risk studies. It further describes operational experience, licensing procedures under the Atomic Energy Law, research in reactor safety and the nuclear fuel cycle. (DG) [de

  17. Assessment of the world market for small and medium reactors

    International Nuclear Information System (INIS)

    Csik, B.J.

    1998-01-01

    The market for SMRs until 2015 was assessed by individual countries, taking into account energy demand and supply patterns, growth rates, energy resources, economic and financial resources, electric grids, industrial and technical development, infrastructure availability, environmental and nuclear safety concerns and other policy issues. The market assessment includes all applications of these reactors, that is electricity generation as well as the supply of process head and district heating. It is expected that SMRs will be deployed primarily in countries which have already started nuclear projects, in particular in countries which have developed SMR designs themselves. Thus, projects would be supplied predominantly by domestic sources in the years ahead; later, the export market is expected to attain more importance. It is further expected that over two thirds of the SMR units would be in the medium size range, i.e. from 300 to 700 MW(e), the rest would be smaller. About one third of the SMRs to be implemented are expected to supply heat and/or electricity to integrated seawater desalination plants. More than half of these reactors would be below 300 MW(e) or 1000 MW(th). The overall market is estimated at about 60 to 100 SMR units to be implemented up to the year 2015. It is recognized that forecasts, just like national development plans, tend to err on the optimistic side. Therefore, an overall market estimate of 70 to 80 units seems reasonable. (author)

  18. Gamma Radiation Assessment In Kartini Reactor And Its Vicinity

    International Nuclear Information System (INIS)

    Yazid, M.; Supriyatni, E.; Maryono; Bastianudin, Aris

    2000-01-01

    Measurement to calculate dose assesment for gamma radiation in Kartini Reactor and its vicinity has been done whether on operated or un operated condition. Measurement was performed using height pressured ionization chamber, Reuther Stokes RS-112 production. Measurement location was determined based on distance variation inwardly and outwardly of reactor building and its vicinity. The result showed that the average dose rate in the reactor building when un operated is in the range of 11.4-38.6 mu rad/hour and when the reactor operated is 166.4-1910.9 mu rad/hour. While the vicinity of the reactor on operated condition the average dose rate is 34.4-38.6 mu rad/hour in un operated condition is 6.9-7.0 mu rad/hour. This result showed that the reactor operated did not rise the radiation exposure level in its vicinity. From the personnel assesment dose rate of gamma radiation is 28.54 mrem/week on operated condition, 0.90 mrem.week on un operated condition. While dose rate outside the reactor is 0.44 and 0.27 mrem/week for operated and un operated condition consecutively. This dose rate is still below maximum permissible dose than recommended by the national regulation of radiation protection from BAPETEN No. 01/Ka.BAPETEN/V-99

  19. Reactor units for power supply to the Russian Arctic regions: Priority assessment of nuclear energy sources

    Directory of Open Access Journals (Sweden)

    Mel'nikov N. N.

    2017-03-01

    Full Text Available Under conditions of competitiveness of small nuclear power plants (SNPP and feasibility of their use to supply power to remote and inaccessible regions the competition occurs between nuclear energy sources, which is caused by a wide range of proposals for solving the problem of power supply to different consumers in the decentralized area of the Russian Arctic power complex. The paper suggests a methodological approach for expert assessment of the priority of small power reactor units based on the application of the point system. The priority types of the reactor units have been determined based on evaluation of the unit's conformity to the following criteria: the level of referentiality and readiness degree of reactor units to implementation; duration of the fuel cycle, which largely determines an autonomy level of the nuclear energy source; the possibility of creating a modular block structure of SNPP; the maximum weight of a transported single equipment for the reactor unit; service life of the main equipment. Within the proposed methodological approach the authors have performed a preliminary ranking of the reactor units according to various criteria, which allows quantitatively determining relative difference and priority of the small nuclear power plants projects aimed at energy supply to the Russian Arctic. To assess the sensitivity of the ranking results to the parameters of the point system the authors have observed the five-point and ten-point scales under variations of importance (weights of different criteria. The paper presents the results of preliminary ranking, which have allowed distinguishing the following types of the reactor units in order of their priority: ABV-6E (ABV-6M, "Uniterm" and SVBR-10 in the energy range up to 20 MW; RITM-200 (RITM-200M, KLT-40S and SVBR-100 in the energy range above 20 MW.

  20. Nuclear Computerized Library for Assessing Reactor Risk (NUCLARR)

    International Nuclear Information System (INIS)

    Gilmore, W.E.; Blackman, H.S.; Ryan, T.G.

    1986-01-01

    The Nuclear Computerized Library for Assessing Reactor Risk (NUCLARR) program is a multiyear effort sponsored by the NRC and is being conducted at the Idaho National Engineering Laboratory (INEL). The goal of this program is to establish and operate computerized data base management tools for the human reliability data bank specification developed by Comer and Donovan. The NRC and the risk analysis community recognized that implementing a fully functional library would not be feasible, or practical, without the aid of computerized tools for management and manipulation of its data sources. The end users of the NUCLARR can be classified into three categories according to specific needs. The first category is those users interested in reviewing individual data sources for a given situation. The second category of users selects multiple data sources for a specific case, summarizing the information, and performing comparative studies. The last category of users interfaces the NUCLARR with other programming applications, such as other data banks, and simulation models of risk assessment. Project status is provided in the paper

  1. Qualitative assessment of the value of the Ohio State University TRIGA reactor

    International Nuclear Information System (INIS)

    Binney, S.E.; Johnson, A.G.

    1989-01-01

    The Oregon State University (OSU) TRIGA Reactor (OSTR) is a major regional research, training, and service facility. The OSTR supports a wide variety of organizations at the local, state, regional, national, and international levels. Examples of usage of the OSTR are given in this paper to serve as a basis for assessing the value of the OSTR to its user organizations. It is difficult to assess the value of a facility such as the OSTR quantitatively, primarily because a dollar value cannot be assigned to many of the services that the OSTR performs, e.g., forensic analysis to assist police agencies in criminal cases. Significant qualitative statements can be made, however, to demonstrate the fact that the value of a research reactor facility such as the OSTR substantially outweighs the capital and operating costs of such a facility. Analysis of the data presented above clearly indicates that the value of the OSTR facility is overwhelmingly positive, i.e., the benefits associated with the services provided by the OSTR facility outweigh the cost of providing such services by perhaps as much as an order of magnitude

  2. Loss of coolant accident (LOCA) analysis for McMaster Nuclear Reactor through probabilistic risk assessment (PRA)

    Energy Technology Data Exchange (ETDEWEB)

    Ha, T.; Garland, W.J. [McMaster Univ., Dept. of Engineering Physics, Hamilton, Ontario (Canada)]. E-mail: hats@mcmaster.ca

    2006-07-01

    A probabilistic risk assessment (PRA) was conducted for the loss of coolant accident (LOCA) sequence in the McMaster Nuclear Reactor (MNR). A level 1 PRA was completed including event sequence modeling, system modeling, and quantification. To support the quantification of the accident sequence identified, data analysis using the Bayesian method and human reliability analysis (HRA) using the ASEP approach were performed. Since human performance in research reactors is significantly different from that in power reactors, a different time-oriented HRA model was proposed and applied for the estimation of the human error probability (HEP) of core relocation. This HEP estimate was less than that by the ASEP approach by a factor of about 2. These two HEP estimates were used for sensitivity analysis, and modeling uncertainty in the PRA models was quantified. This showed the necessity of appropriate human reliability models in PRA for research reactors. This method could be implemented for the operators' actions which require extensive manual execution with little cognitive load, as might be the case for some maintenance operations in power reactors. (author)

  3. Integrity assessment of TAPS reactor pressure vessel at extended EOL using surveillance test results

    International Nuclear Information System (INIS)

    Chatterjee, S.; Shah, Priti Kotak

    2008-05-01

    Integrity assessment of pressure vessels of nuclear reactors (RPV) primarily concentrates on the prevention of brittle failure and conditions are defined under which brittle failure can be excluded. Accordingly, two approaches based on Transition Temperature Concept and Fracture Mechanics Concept were adopted using the impact test results of three credible surveillance data sets obtained from the surveillance specimens of Tarapur Atomic Power Station. RT NDT data towards end of life (EOL) were estimated from the impact test results in accordance with the procedures of USNRC Regulatory Guide 1.99, Rev. 2 and were used as primary input for assessment of the vessel integrity. SA302B (nickel modified) steel cladded with stainless steel is used as the pressure vessel material for the two 210 MWe boiling water reactors of the Tarapur Atomic Power Station (TAPS). The reactors were commissioned during the year 1969. The chemical compositions of SA302B (modified) steel used in fabricating the vessel and the specified tensile property and the Charpy impact property requirements of the steel broadly meet ASME specified requirements. Therefore, the pressure temperature limit curves prescribed by General Electric (G.E.) were compared with those as obtained using procedures of ASME Section XII, Appendix G. The tensile and the Charpy impact properties at 60 EFPY of vessel operation as derived from the surveillance specimens even fulfilled the specified requirements for the virgin material of ASME. Integrity assessment carried out using the two approaches indicated the safety of the vessel for continued operation up to 60 EFPY. (author)

  4. Life cycle assessment of greenhouse gas emissions, water and land use for concentrated solar power plants with different energy backup systems

    International Nuclear Information System (INIS)

    Klein, Sharon J.W.; Rubin, Edward S.

    2013-01-01

    Concentrated solar power (CSP) is unique among intermittent renewable energy options because for the past four years, utility-scale plants have been using an energy storage technology that could allow a CSP plant to operate as a baseload renewable energy generator in the future. No study to-date has directly compared the environmental implications of this technology with more conventional CSP backup energy options. This study compares the life cycle greenhouse gas (GHG) emissions, water consumption, and direct, onsite land use associated with one MW h of electricity production from CSP plants with wet and dry cooling and with three energy backup systems: (1) minimal backup (MB), (2) molten salt thermal energy storage (TES), and (3) a natural gas-fired heat transfer fluid heater (NG). Plants with NG had 4–9 times more life cycle GHG emissions than plants with TES. Plants with TES generally had twice as many life cycle GHG emissions as the MB plants. Dry cooling reduced life cycle water consumption by 71–78% compared to wet cooling. Plants with larger backup capacities had greater life cycle water consumption than plants with smaller backup capacities, and plants with NG had lower direct, onsite life cycle land use than plants with MB or TES. - highlights: • We assess life cycle environmental effects of concentrated solar power (CSP). • We compare CSP with three energy backup technologies and two cooling technologies. • We selected solar field area to minimize energy cost for plants with minimal backup and salt storage. • Life cycle greenhouse gas emissions were 4–9 times lower with thermal energy storage than with fossil fuel backup. • Dry cooling reduced life cycle water use by 71–78% compared to wet cooling

  5. Method to Assess the Radionuclide Inventory of Irradiated Graphite from Gas-Cooled Reactors - 13072

    Energy Technology Data Exchange (ETDEWEB)

    Poncet, Bernard [EDF-CIDEN, 154 Avenue Thiers, CS 60018, F-69458 LYON cedex 06 (France)

    2013-07-01

    About 17,000 t of irradiated graphite waste will be produced from the decommissioning of the six French gas-cooled nuclear reactors. Determining the radionuclide (RN) content of this waste is of relevant importance for safety reasons and in order to determine the best way to manage them. For many reasons the impurity content that gave rise to the RNs in irradiated graphite by neutron activation during operation is not always well known and sometimes actually unknown. So, assessing the RN content by the use of traditional calculation activation, starting from assumed impurity content, leads to a false assessment. Moreover, radiochemical measurements exhibit very wide discrepancies especially on RN corresponding to precursor at the trace level such as natural chlorine corresponding to chlorine 36. This wide discrepancy is unavoidable and is due to very simple reasons. The level of impurity is very low because the uranium fuel used at that very moment was not enriched, so it was a necessity to have very pure nuclear grade graphite and the very low size of radiochemical sample is a simple technical constraint because device size used to get mineralization product for measurement purpose is limited. The assessment of a radionuclide inventory only based on few number of radiochemical measurements lead in most cases, to a gross over or under-estimation that is detrimental for graphite waste management. A method using an identification calculation-measurement process is proposed in order to assess a radiological inventory for disposal sizing purpose as precise as possible while guaranteeing its upper character. This method present a closer approach to the reality of the main phenomenon at the origin of RNs in a reactor, while also incorporating the secondary effects that can alter this result such as RN (or its precursor) release during reactor operation. (authors)

  6. Preliminary Safeguards Assessment for the Pebble-Bed Fluoride High-Temperature Reactor (PB-FHR) Concept

    Energy Technology Data Exchange (ETDEWEB)

    Disser, Jay; Arthur, Edward; Lambert, Janine

    2016-09-01

    This report examines a preliminary design for a pebble bed fluoride salt-cooled high temperature reactor (PB-FHR) concept, assessing it from an international safeguards perspective. Safeguards features are defined, in a preliminary fashion, and suggestions are made for addressing further nuclear materials accountancy needs.

  7. Excerpts from the report: "BeyondTMY - Meteorological data sets for CSP/STE performance simulations"

    Science.gov (United States)

    Nielsen, Kristian Pagh; Vignola, Frank; Ramírez, Lourdes; Blanc, Philippe; Meyer, Richard; Blanco, Manuel

    2017-06-01

    In order to facilitate comprehensive economic modeling of CSP/STE power plants realistic long-term meteorological datasets with temporal resolution down to 1 minute is a main premise. Currently available standard datasets do not fulfil this premise. The datasets also need to combine the high quality of well-maintained ground-based irradiance measurements and the global coverage of satellite-derived data. Even with the best available data it is necessary to account for the uncertainty in this and the sampling uncertainty from finite time-series to enable the optimal statistical characterization. It is a general challenge that satellite-derived data lack the required temporal resolution, and also often does not cover periods with major volcanic eruptions. Here we see prospects in synthetically generated realistic datasets, although research and development work is required on how to optimally produce and quality assure these.

  8. Real time system design of motor imagery brain-computer interface based on multi band CSP and SVM

    Science.gov (United States)

    Zhao, Li; Li, Xiaoqin; Bian, Yan

    2018-04-01

    Motion imagery (MT) is an effective method to promote the recovery of limbs in patients after stroke. Though an online MT brain computer interface (BCT) system, which apply MT, can enhance the patient's participation and accelerate their recovery process. The traditional method deals with the electroencephalogram (EEG) induced by MT by common spatial pattern (CSP), which is used to extract information from a frequency band. Tn order to further improve the classification accuracy of the system, information of two characteristic frequency bands is extracted. The effectiveness of the proposed feature extraction method is verified by off-line analysis of competition data and the analysis of online system.

  9. Preconceptual design and assessment of a Tokamak Hybrid Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Teofilo, V.L.; Leonard, B.R. Jr.; Aase, D.T.

    1980-09-01

    The preconceptual design of a commercial Tokamak Hybrid Reactor (THR) power plant has been performed. The tokamak fusion driver for this hybrid is operated in the ignition mode. The D-T fusion plasma, which produces 1140 MW of power, has a major radius of 5.4 m and a minor radius of 1.0 m with an elongation of 2.0. Double null poloidal divertors are assumed for impurity control. The confining toroidal field is maintained by D-shaped Nb/sub 3/Sn superconducting magnets with a maximum field of 12T at the coil. Three blankets with four associated fuel cycle alternatives have been combined with the ignited tokamak fusion driver. The engineering, material, and balance of plant design requirements for the THR are briefly described. Estimates of the capital, operating and maintenance, and fuel cycle costs have been made for the various driver/blanket combinations and an assessment of the market penetrability of hybrid systems is presented. An analysis has been made of the nonproliferation aspects of the hybrid and its associated fuel cycles relative to fission reactors. The current and required level of technology for both the fusion and fission components of the hybrid system has been reviewed. Licensing hybrid systems is also considered.

  10. Preconceptual design and assessment of a Tokamak Hybrid Reactor

    International Nuclear Information System (INIS)

    Teofilo, V.L.; Leonard, B.R. Jr.; Aase, D.T.

    1980-09-01

    The preconceptual design of a commercial Tokamak Hybrid Reactor (THR) power plant has been performed. The tokamak fusion driver for this hybrid is operated in the ignition mode. The D-T fusion plasma, which produces 1140 MW of power, has a major radius of 5.4 m and a minor radius of 1.0 m with an elongation of 2.0. Double null poloidal divertors are assumed for impurity control. The confining toroidal field is maintained by D-shaped Nb 3 Sn superconducting magnets with a maximum field of 12T at the coil. Three blankets with four associated fuel cycle alternatives have been combined with the ignited tokamak fusion driver. The engineering, material, and balance of plant design requirements for the THR are briefly described. Estimates of the capital, operating and maintenance, and fuel cycle costs have been made for the various driver/blanket combinations and an assessment of the market penetrability of hybrid systems is presented. An analysis has been made of the nonproliferation aspects of the hybrid and its associated fuel cycles relative to fission reactors. The current and required level of technology for both the fusion and fission components of the hybrid system has been reviewed. Licensing hybrid systems is also considered

  11. Measurement control design and performance assessment in the Integral Fast Reactor fuel cycle

    International Nuclear Information System (INIS)

    Orechwa, Y.; Bucher, R.G.

    1994-01-01

    The Integral Fast Reactor (IFR)--consisting of a metal fueled and liquid metal cooled reactor together with an attendant fuel cycle facility (FCF)--is currently undergoing a phased demonstration of the closed fuel cycle at Argonne National Laboratory. The recycle technology is pyrometalurgical based with incomplete fission product separation and all transuranics following plutonium for recycle. The equipment operates in batch mode at 500 to 1,300 C. The materials are highly radioactive and pyrophoric, thus the FCF requires remote operation. Central to the material control and accounting system for the FCF are the balances for mass measurements. The remote operation of the balances limits direct adjustment. The radiation environment requires that removal and replacement of the balances be minimized. The uniqueness of the facility precludes historical data for design and performance assessment. To assure efficient operation of the facility, the design of the measurement control system has called for procedures which assess the performance of the balances in great detail and will support capabilities for the correction of systematic changes in the performance of the balances through software

  12. Self assessment of safety culture in HANARO using the code of conduct on the safety of research reactor by IAEA

    International Nuclear Information System (INIS)

    Lim, I.C.; Hwang, S.Y.; Woo, J.S.; Lee, M.; Jun, B.J.

    2003-01-01

    Full text: The safety culture in HANARO was self-assessed in accordance with the Code of Conduct on the Safety of Research Reactor drafted by IAEA. From 2002, IAEA has worked on the development of the Code of Conduct to achieve and maintain high level of nuclear safety in research reactors worldwide through the enhancement of national measures and international co-operation including, where appropriate, safety related technical cooperation. It defines the role of the state, the role of the regulatory body, the role of the operating organization and the role of the IAEA. As for the role of operating organization, the code specifies general requirements in assessment and verification of safety, financial and human resources, quality assurance, human factors, radiation protection and emergency preparedness. It also defines the role of operating organization for safety of research reactor in siting, design, operation, maintenance, modification and utilization as well. All of these items are the subjects for safety culture implementation, which means the Code could be a guideline for an operating organization to assess its safety culture. The self-assessment of safety culture in HANARO was made by using the sections of the Code describing the role of the operating organization for safety of research reactor. The major assessment items and the practices in HANARO for each items are as follow: The SAR of HANARO was reviewed by the regulatory body before the construction and the fuel loading of HANARO. Major design modifications and new installation of utilization facility needs the approval from regulatory body and safety assessment is a requirement for the approval. The Tech. Spec. for HANARO Operation specifies the analysis, surveillance, testing and inspection for HANARO operation. The reactor operation is mainly supported by the government and partly by nuclear R and D fund. The education and training of operation staff are one of major tasks of operating organization

  13. At-reactor storage concepts criteria for preliminary assessment

    International Nuclear Information System (INIS)

    Boydston, L.A.

    1981-12-01

    The licensing, safety, and environmental considerations of four wet and four dry at-reactor storage concepts are presented. Physical criteria for each concept are examined to determine the minimum site and facility requirements which must be met by a utility which desires to expand its at-reactor spent fuel storage capability

  14. Cost assessment of a generic magnetic fusion reactor

    International Nuclear Information System (INIS)

    Sheffield, J.; Dory, R.A.; Cohn, S.M.; Delene, J.G.; Parsly, L.F.; Ashby, D.E.T.F.; Reiersen, W.T.

    1986-03-01

    A generic reactor model is used to examine the economic viability of generating electricity by magnetic fusion. The simple model uses components that are representative of those used in previous reactor studies of deuterium-tritium-burning tokamaks, stellarators, bumpy tori, reversed-field pinches (RFPs), and tandem mirrors. Conservative costing assumptions are made. The generic reactor is not a tokamak; rather, it is intended to emphasize what is common to all magnetic fusion rectors. The reactor uses a superconducting toroidal coil set to produce the dominant magnetic field. To this extent, it is not as good an approximation to systems such as the RFP in which the main field is produced by a plasma current. The main output of the study is the cost of electricity as a function of the weight and size of the fusion core - blanket, shield, structure, and coils. The model shows that a 1200-MW(e) power plant with a fusion core weight of about 10,000 tonnes should be competitive in the future with fission and fossil plants. Studies of the sensitivity of the model to variations in the assumptions show that this result is not sensitively dependent on any given assumption. Of particular importance is the result that a fusion reactor of this scale may be realized with only moderate advances in physics and technology capabilities

  15. On the fundamentals of nuclear reactor safety assessment. Inherent threats and their implications

    Energy Technology Data Exchange (ETDEWEB)

    Hyvaerinen, J. [Finnish Centre for Radiation and Nuclear Safety, Helsinki (Finland). Nuclear Safety Dept.

    1996-12-01

    The thesis addresses some fundamental questions related to implementation and assessment of nuclear safety. The safety principles and assessment methods are described, followed by descriptions of selected novel technical challenges to nuclear safety. The novel challenges encompass a wide variety of technical issues, thus providing insights on the limitations of conventional safety assessment methods. Study of the limitations suggests means to improve nuclear reactor design criteria and safety assessment practices. The novel safety challenges discussed are (1) inherent boron dilution in PWRs, (2) metallic insulation performance with respect to total loss of emergency cooling systems in a loss-of-coolant accident, and (3) horizontal steam generator heat transfer performance at natural circulation conditions. (50 refs.).

  16. On the fundamentals of nuclear reactor safety assessment. Inherent threats and their implications

    International Nuclear Information System (INIS)

    Hyvaerinen, J.

    1996-12-01

    The thesis addresses some fundamental questions related to implementation and assessment of nuclear safety. The safety principles and assessment methods are described, followed by descriptions of selected novel technical challenges to nuclear safety. The novel challenges encompass a wide variety of technical issues, thus providing insights on the limitations of conventional safety assessment methods. Study of the limitations suggests means to improve nuclear reactor design criteria and safety assessment practices. The novel safety challenges discussed are (1) inherent boron dilution in PWRs, (2) metallic insulation performance with respect to total loss of emergency cooling systems in a loss-of-coolant accident, and (3) horizontal steam generator heat transfer performance at natural circulation conditions. (50 refs.)

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

    International Nuclear Information System (INIS)

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

    1981-09-01

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

  18. Redox-neutral rhodium-catalyzed C-H functionalization of arylamine N-oxides with diazo compounds: primary C(sp(3))-H/C(sp(2))-H activation and oxygen-atom transfer.

    Science.gov (United States)

    Zhou, Bing; Chen, Zhaoqiang; Yang, Yaxi; Ai, Wen; Tang, Huanyu; Wu, Yunxiang; Zhu, Weiliang; Li, Yuanchao

    2015-10-05

    An unprecedented rhodium(III)-catalyzed regioselective redox-neutral annulation reaction of 1-naphthylamine N-oxides with diazo compounds was developed to afford various biologically important 1H-benzo[g]indolines. This coupling reaction proceeds under mild reaction conditions and does not require external oxidants. The only by-products are dinitrogen and water. More significantly, this reaction represents the first example of dual functiaonalization of unactivated a primary C(sp(3) )H bond and C(sp(2) )H bond with diazocarbonyl compounds. DFT calculations revealed that an intermediate iminium is most likely involved in the catalytic cycle. Moreover, a rhodium(III)-catalyzed coupling of readily available tertiary aniline N-oxides with α-diazomalonates was also developed under external oxidant-free conditions to access various aminomandelic acid derivatives by an O-atom-transfer reaction. © 2015 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  19. International assessment of application of the Code of Conduct on the Safety of Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Shokr, A.M. [Atomic Energy Authority, Abouzabal (Egypt). Egypt Second Research Reactor

    2015-11-15

    The self-assessments performed by thirty-eight countries on application of the Code of Conduct on the Safety of Research Reactors were analyzed and discussed. The results of this analysis were used to identify areas of satisfactory application of the Code and area needing improvements, and therefore require more attention worldwide. The results showed improvement in application of the Code provisions; notably in aging management, regulatory supervision, and consideration of human factors. However, there is a continuing need for further improvement in these areas, as well as in operational radiation protection, emergency preparedness and decommissioning planning. Additionally, increased attention needs to be given to periodic safety reviews, evaluation of site-specific hazards, and assessment of extreme external events. The results showed consistency with the feedback from other sources of information on generic safety issues for research reactors.

  20. International assessment of application of the Code of Conduct on the Safety of Research Reactors

    International Nuclear Information System (INIS)

    Shokr, A.M.

    2015-01-01

    The self-assessments performed by thirty-eight countries on application of the Code of Conduct on the Safety of Research Reactors were analyzed and discussed. The results of this analysis were used to identify areas of satisfactory application of the Code and area needing improvements, and therefore require more attention worldwide. The results showed improvement in application of the Code provisions; notably in aging management, regulatory supervision, and consideration of human factors. However, there is a continuing need for further improvement in these areas, as well as in operational radiation protection, emergency preparedness and decommissioning planning. Additionally, increased attention needs to be given to periodic safety reviews, evaluation of site-specific hazards, and assessment of extreme external events. The results showed consistency with the feedback from other sources of information on generic safety issues for research reactors.

  1. Environmental impact assessment around TRIGA research reactor

    International Nuclear Information System (INIS)

    Lee, Jeong Ho; Lee, Hyun Duk; Lee, Young Bok; Cheong, Kyu Hoi; Ahn, Jong Sung; Kim, Kug Chan; You, Byung Sun; Kim, Byung Woo; Kim, Sang Bok; Han Moon Hee

    1985-01-01

    Population distribution, atmospheric change, X/Q, characteristics of terrestrial ecosystem around Seoul site were surveyed. Environmental radiation and radioactivities such as grossα, grossβ, Cs-137, Sr-90 and H-3 of various environmental samples were analyzed. The values of environmental radiation dose tended to increase gradually in the light of the recent five years' results of environmental radiation monitoring around the nuclear power plants from 1980 to 1984, however, the changes were not significant. In addition, continuous assessment of environmental radiation monitoring on the roofs of main building and life science building at KAERI showed that the environmental radiation dose tended to increase a little during the night time. Judging from the above results, it is concluded that environmental contamination level by radioactive materials could be ignored in the case of radioisotope production or experiment using radioisotopes except the release of gaseous radioactive materials such as Ar-41 of short half life by the operation of nuclear reactor. (Author)

  2. Assessment of core structural materials and surveillance programme of research reactors. Report of the consultants meeting. Working material

    International Nuclear Information System (INIS)

    2009-01-01

    A series of presentations on the assessment of core structural components and materials at their facilities were given by the experts. The different issues related to degradation mechanisms were discussed. The outputs include a more thorough understanding of the specific challenges related to Research Reactors (RRs) as well as proposals for activities which could assist RR organizations in their efforts to address the issues involved. The experts recommend that research reactor operators consider implementation of surveillance programs for materials of core structural components, as part of ageing management program (TECDOC-792 and DS-412). It is recognised by experts that adequate archived structural material data is not available for many RRs. Access to this data and extension of existing material databases could help many operating organisations extend the operation of their RRs. The experts agreed that an IAEA Technical Meeting (TM) on Assessment of Core Structural Materials should be organised in December 2009 (IAEA HQ Vienna). The proposed objectives of the TM are: (i) exchange of detailed technical information on the assessment and ageing management of core structural materials, (ii) identification of materials of interest for further investigation, (iii) proposal for a new IAEA CRP on Assessment of Core Structural Materials, and (iv) identification of RRs prepared to participate in proposed CRP. Based on the response to a questionnaire prepared for the 2008 meeting of the Technical Working Group for Research Reactors, the number of engineering capital projects related to core structural components is proportionally lower than those related to,for example, I and C or electrical power systems. This implies that many operating research reactors will be operating longer using their original core structural components and justifies the assessment and evaluation programmes and activities proposed in this report. (author)

  3. Assessment of specialized educational programs for licensed nuclear reactor operators

    International Nuclear Information System (INIS)

    Melber, B.D.; Saari, L.M.; White, A.S.; Geisendorfer, C.L.; Huenefeld, J.C.

    1986-02-01

    This report assesses the job-relatedness of specialized educational programs for licensed nuclear reactor operators. The approach used involved systematically comparing the curriculum of specialized educational programs for college credit, to academic knowledge identified as necessary for carrying out the jobs of licenses reactor operators. A sample of eight programs, including A.S. degree, B.S. degree, and coursework programs were studied. Subject matter experts in the field of nuclear operations curriculum and training determined the extent to which individual program curricula covered the identified job-related academic knowledge. The major conclusions of the report are: There is a great deal of variation among individual programs, ranging from coverage of 15% to 65% of the job-related academic knowledge. Four schools cover at least half, and four schools cover less than one-third of this knowledge content; There is no systematic difference in the job-relatedness of the different types of specialized educational programs, A.S. degree, B.S. degree, and coursework; and Traditional B.S. degree programs in nuclear engineering cover as much job-related knowledge (about one-half of this knowledge content) as most of the specialized educational programs

  4. Reliability assessment of emergency exhaust system in a pool-type research reactor

    International Nuclear Information System (INIS)

    Khan, S.A.

    1991-01-01

    The reliability of an extract system in a swimming-pool-type research reactor has been assessed. A global fault-tree analysis technique has been utilized. The basic event reliability data is based on both generic and reactor specific informations. The unavailability of the extract system is quantified in terms of the unavailability of the various functional requirements of the system. The unavailability is expressed as the probability of failure on demand. The computer system unavailability is determined from the minimal cutsets of the system. It is found that only three events have a major contribution to the top event, i.e., failures of compressed air supply, electric power supply and solenoid valve. A sensitivity analysis is performed to show the effects of variations in the data values of the dominant cutsets. An uncertainty analysis was also performend on the fault tree. The evaluations show that the reactor extract system lacks diversity and redundance in most of its components. It is tolerant of most minor degradations, as these are taken care of by the operating policies and procedures. However, it can not tolerate common cause failures, e.g. simultaneous compressed air and electric power supply failure. Based upon the results obtained, some recommendations are made. (orig.)

  5. Using SAFRAN Software to Assess Radiological Hazards from Dismantling of Tammuz-2 Reactor Core at Al-tuwaitha Nuclear Site

    Science.gov (United States)

    Abed Gatea, Mezher; Ahmed, Anwar A.; jundee kadhum, Saad; Ali, Hasan Mohammed; Hussein Muheisn, Abbas

    2018-05-01

    The Safety Assessment Framework (SAFRAN) software has implemented here for radiological safety analysis; to verify that the dose acceptance criteria and safety goals are met with a high degree of confidence for dismantling of Tammuz-2 reactor core at Al-tuwaitha nuclear site. The activities characterizing, dismantling and packaging were practiced to manage the generated radioactive waste. Dose to the worker was considered an endpoint-scenario while dose to the public has neglected due to that Tammuz-2 facility is located in a restricted zone and 30m berm surrounded Al-tuwaitha site. Safety assessment for dismantling worker endpoint-scenario based on maximum external dose at component position level in the reactor pool and internal dose via airborne activity while, for characterizing and packaging worker endpoints scenarios have been done via external dose only because no evidence for airborne radioactivity hazards outside the reactor pool. The in-situ measurements approved that reactor core components are radiologically activated by Co-60 radioisotope. SAFRAN results showed that the maximum received dose for workers are (1.85, 0.64 and 1.3mSv/y) for activities dismantling, characterizing and packaging of reactor core components respectively. Hence, the radiological hazards remain below the low level hazard and within the acceptable annual dose for workers in radiation field

  6. Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR): Summary description

    International Nuclear Information System (INIS)

    Gertman, D.I.; Gilmore, W.E.; Galyean, W.J.; Groh, M.R.; Gentillon, C.D.; Gilbert, B.G.; Reece, W.J.

    1990-05-01

    The Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) is an automate data base management system for storing and processing human error probability and hardware component failure rate data. The NUCLARR system software resides on an IBM (or compatible) personal microcomputer. NUCLARR can be accessed by the end user to furnish data suitable for input in human and/or hardware reliability analysis to support a variety of risk assessment activities. The NUCLARR system is documented in a five-volume series of reports. This document Volume 1, of this series is the Summary Description, which presents an overview of the data management system, including a description of data collection, data qualification, data structure, and taxonomies. Programming activities, procedures for processing data, a user's guide, and hard copy data manual are presented in Volumes 2 through 5, NUREG/CR-4639

  7. Fracture assessment of the Oskarshamn 1 reactor pressure vessel under cold over-pressurization

    Energy Technology Data Exchange (ETDEWEB)

    Sattari-Far, I. [DNV Technical Consulting AB, Stockholm (Sweden)

    2001-03-01

    The major motivation of this study was to develop a methodology for fracture assessment of surface defects in the 01 reactor pressure vessel under cold loading scenarios, particularly the cold over-pressurization event. According to a previous study, the FENIX project, the cold over-pressurization of the O1 reactor is a limiting loading case, as the ductile/brittle transition temperature (RT{sub NDT}) of certain welds in the O1 beltline region may be over 100 deg C at the-end-of-life condition. The FENIX project gave values of the acceptable and critical crack depth to be equal to the thickness of the cladding layer (about 6 mm) under this load case using the ASME K{sub Ic} reference curve methodology. This study is aimed to develop a methodology to give a more precise fracture assessment of the O1 reactor under cold loading scenarios. Some of the main objectives of this study have been as below: To prepare a material which can simulate the mechanical properties and RT{sub NDT} of the O1 reactor at the end-of-life conditions. To conduct a fracture mechanics test program to cover the essential influencing factors, such as crack geometry (shallow and deep cracks) and loading condition (uniaxial and biaxial) on the cleavage fracture toughness. To perform fracture mechanics analyses to identify a suitable methodology for assessment of the experimental results. To study the responses of engineering fracture assessment methods to the experimental results from the clad specimens. To propose a fracture assessment procedure for determination of the acceptable and critical flaw sizes in the 01 reactor under the cold loading events. A test program consisted of experiments on standard SEN(B) specimens and clad beams, containing surface cracks was conducted during the course of this project. A total of nine clad beams and clad cruciform specimens were tested under uniaxial and biaxial loading. The test material is reactor steel of type A 508 Grade B, which is specially heat

  8. The assessment of voce coefficient for WWR-c reactor

    International Nuclear Information System (INIS)

    Kochnov, O.Yu.; Rybkin, N.I.

    2006-01-01

    The air cavity effect in WWR-ts reactor core on the total reactivity is analyzed. The experimental data of void coefficient depending on the air cavity position inside the reactor core are obtained [ru

  9. Business interruption and loss of assets risk assessment in support of the design of an innovative concentrating solar power plant

    International Nuclear Information System (INIS)

    Amato, Andrea; Gallisto, Maurizio; Maccari, Augusto; Paganelli, Mauro; Compare, Michele; Zio, Enrico

    2011-01-01

    Concentrating Solar Power (CSP) plants are a promising technology of renewable energy production, as witnessed by the increasing public and private investments during the last decade. The assessment of the associated risks of business interruption (loss of production) and loss of assets due to the occurrence of undesired internal or external events, such as failures of components, unfavorable environmental conditions, etc., brings added values by informing design modifications and contributing to production assurance, for rational Company investments in these environmentally sustainable power plants. This work presents and applies a methodology for assessing the risks associated to a CSP of innovative design. The methodology is derived from traditional system risk analysis, specifically focused only on the economic consequences of the internal events of failure behavior of components. The innovation in the design considered is particularly aimed at augmenting the CSP intrinsic capability of being equipped with thermal storage systems by the introduction of a molten salt mixture as heat transfer fluid. This technology presents evident advantages in terms of system simplification and reduction of production costs but on the other hand introduces a risk factor with regards to the solidification of the salt mixture that occurs at about 240 C. (author)

  10. Creating a spatial multi-criteria decision support system for energy related integrated environmental impact assessment

    International Nuclear Information System (INIS)

    Wanderer, Thomas; Herle, Stefan

    2015-01-01

    By their spatially very distributed nature, profitability and impacts of renewable energy resources are highly correlated with the geographic locations of power plant deployments. A web-based Spatial Decision Support System (SDSS) based on a Multi-Criteria Decision Analysis (MCDA) approach has been implemented for identifying preferable locations for solar power plants based on user preferences. The designated areas found serve for the input scenario development for a subsequent integrated Environmental Impact Assessment. The capabilities of the SDSS service get showcased for Concentrated Solar Power (CSP) plants in the region of Andalusia, Spain. The resulting spatial patterns of possible power plant sites are an important input to the procedural chain of assessing impacts of renewable energies in an integrated effort. The applied methodology and the implemented SDSS are applicable for other renewable technologies as well. - Highlights: • The proposed tool facilitates well-founded CSP plant siting decisions. • Spatial MCDA methods are implemented in a WebGIS environment. • GIS-based SDSS can contribute to a modern integrated impact assessment workflow. • The conducted case study proves the suitability of the methodology

  11. Creating a spatial multi-criteria decision support system for energy related integrated environmental impact assessment

    Energy Technology Data Exchange (ETDEWEB)

    Wanderer, Thomas, E-mail: thomas.wanderer@dlr.de; Herle, Stefan, E-mail: stefan.herle@rwth-aachen.de

    2015-04-15

    By their spatially very distributed nature, profitability and impacts of renewable energy resources are highly correlated with the geographic locations of power plant deployments. A web-based Spatial Decision Support System (SDSS) based on a Multi-Criteria Decision Analysis (MCDA) approach has been implemented for identifying preferable locations for solar power plants based on user preferences. The designated areas found serve for the input scenario development for a subsequent integrated Environmental Impact Assessment. The capabilities of the SDSS service get showcased for Concentrated Solar Power (CSP) plants in the region of Andalusia, Spain. The resulting spatial patterns of possible power plant sites are an important input to the procedural chain of assessing impacts of renewable energies in an integrated effort. The applied methodology and the implemented SDSS are applicable for other renewable technologies as well. - Highlights: • The proposed tool facilitates well-founded CSP plant siting decisions. • Spatial MCDA methods are implemented in a WebGIS environment. • GIS-based SDSS can contribute to a modern integrated impact assessment workflow. • The conducted case study proves the suitability of the methodology.

  12. MAPLE research reactor safety uncertainty assessment methodology

    International Nuclear Information System (INIS)

    Sills, H.E.; Duffey, R.B.; Andres, T.H.

    1999-01-01

    The MAPLE (multipurpose Applied Physics Lattice Experiment) reactor is a low pressure, low temperature, open-tank-in pool type research reactor that operates at a power level of 5 to 35 MW. MAPLE is designed for ease of operation, maintenance, and to meet today's most demanding requirements for safety and licensing. The emphasis is on the use of passive safety systems and environmentally qualified components. Key safety features include two independent and diverse shutdown systems, two parallel and independent cooling loops, fail safe operation, and a building design that incorporates the concepts of primary containment supported by secondary confinement

  13. Preliminary risk assessment of the Integral Inherently-Safe Light Water Reactor

    International Nuclear Information System (INIS)

    McCarroll, Kellen R.; Lee, John C.; Manera, Annalisa; Memmott, Matthew J.; Ferroni, Paolo

    2017-01-01

    The Integral, Inherently Safe Light Water Reactor (I 2 S-LWR) concept seeks to significantly increase nuclear power plant safety. The project implements a safety-by-design philosophy, eliminating several initiating events and providing novel, passive safety systems at the conceptual phase. Pursuit of unparalleled safety employs an integrated development process linking design with deterministic and probabilistic safety analyses. Unique aspects of the I 2 S-LWR concept and design process present challenges to the probabilistic risk assessment (PRA), particularly regarding overall flexibility, auditability and resolution of results. Useful approaches to initiating events and conditional failures are presented. To exemplify the risk-informed design process using PRA, a trade-off study of two safety system configurations is presented. Although further optimization is required, preliminary results indicate that the I 2 S-LWR can achieve a core damage frequency (CDF) from internal events less than 1.01 × 10 −8 /ry, including reactor vessel ruptures. Containment bypass frequency due to primary heat exchanger rupture is found to be comparable to non-vessel rupture CDF.

  14. Environmental Assessment of Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    1994-04-01

    The Department of Energy has completed the Environmental Assessment (EA) of Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel and issued a Finding of No Significant Impact (FONSI) for the proposed action. The EA and FONSI are enclosed for your information. The Department has decided to accept a limited number of spent nuclear fuel elements (409 elements) containing uranium that was enriched in the United States from eight research reactors in Austria, Denmark, Germany, Greece, the Netherlands, Sweden, and Switzerland. This action is necessary to maintain the viability of a major US nuclear weapons nonproliferation program to limit or eliminate the use of highly enriched uranium in civil programs. The purpose of the EA is to maintain the cooperation of the foreign research reactor operators with the nonproliferation program while a more extensive Environmental Impact Statement (EIS) is prepared on a proposed broader policy involving the acceptance of up to 15,000 foreign research reactor spent fuel elements over a 10 to 15 year period. Based on an evaluation of transport by commercial container liner or chartered vessel, five eastern seaboard ports, and truck and train modes of transporting the spent fuel overland to the Savannah River Sits, the Department has concluded that no significant impact would result from any combination of port and made of transport. In addition, no significant impacts were found from interim storage of spent fuel at the Savannah River Site

  15. Environmental Assessment of Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    1994-04-01

    The Department of Energy has completed the Environmental Assessment (EA) of Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel and issued a Finding of No Significant Impact (FONSI) for the proposed action. The EA and FONSI are enclosed for your information. The Department has decided to accept a limited number of spent nuclear fuel elements (409 elements) containing uranium that was enriched in the United States from eight research reactors in Austria, Denmark, Germany, Greece, the Netherlands, Sweden, and Switzerland. This action is necessary to maintain the viability of a major US nuclear weapons nonproliferation program to limit or eliminate the use of highly enriched uranium in civil programs. The purpose of the EA is to maintain the cooperation of the foreign research reactor operators with the nonproliferation program while a more extensive Environmental Impact Statement (EIS) is prepared on a proposed broader policy involving the acceptance of up to 15,000 foreign research reactor spent fuel elements over a 10 to 15 year period. Based on an evaluation of transport by commercial container liner or chartered vessel, five eastern seaboard ports, and truck and train modes of transporting the spent fuel overland to the Savannah River Sits, the Department has concluded that no significant impact would result from any combination of port and made of transport. In addition, no significant impacts were found from interim storage of spent fuel at the Savannah River Site.

  16. 75 FR 62892 - Massachusetts Institute of Technology Research Reactor Environmental Assessment and Finding of No...

    Science.gov (United States)

    2010-10-13

    ... NUCLEAR REGULATORY COMMISSION [Docket No. 50-020; NRC-2010-0313] Massachusetts Institute of Technology Research Reactor Environmental Assessment and Finding of No Significant Impact Correction In notice document 2010-24809 beginning on page 61220 in the issue of Monday, October 4, 2010, make the...

  17. An assessment of methods of calculating sodium-voiding reactivity in plutonium-fuelled fast reactors

    International Nuclear Information System (INIS)

    Butland, A.T.D.; Simmons, W.N.; Stevenson, J.M.

    1980-01-01

    After a survey of the requirements an assessment of the accuracy of calculations of the sodium-void effect using UK methods and data is made on the basis of the following work: (a) The analysis of small and large sodium voids in the MOZART and Zebra 13 small (300 MW(e)) fast reactor mock-ups and the BIZET large fast reactor mock-ups, all of conventional design. The analysis was carried out using the UK FGL5 fine group nuclear data library, the MURAL cell code, whole reactor diffusion theory calculations of the neutron flux and perturbation theory methods. Exact perturbation theory was used in many cases, otherwise first-order perturbation theory calculations were adjusted to give results equivalent to exact perturbation theory. (b) Theoretical studies of some effects, including the following: (i) The effects of extrapolating to fuel operating temperature; (ii) Fuel-cycle and burnup effects, including the gradual replacement through a fuel cycle of control-rod absorption by fission product absorption, the loss of fissile material and the change in fuel nuclide relative composition; (iii) The heterogeneity effects of large fuelled subassemblies in pin geometry. (c) Theoretical studies of approximations in the calculational methods, including the following: (i) The importance in the whole reactor calculation of the energy group structure and the spatial mesh, including comparisons of calculations in two (RZ) and three-dimensional geometry; (ii) The importance of reactor material boundaries in the calculation of resonance shielding effects; (iii) The use of neutron fluxes calculated using neutron diffusion theory rather than transport theory. (author)

  18. On the automated assessment of nuclear reactor systems code accuracy

    International Nuclear Information System (INIS)

    Kunz, Robert F.; Kasmala, Gerald F.; Mahaffy, John H.; Murray, Christopher J.

    2002-01-01

    An automated code assessment program (ACAP) has been developed to provide quantitative comparisons between nuclear reactor systems (NRS) code results and experimental measurements. The tool provides a suite of metrics for quality of fit to specific data sets, and the means to produce one or more figures of merit (FOM) for a code, based on weighted averages of results from the batch execution of a large number of code-experiment and code-code data comparisons. Accordingly, this tool has the potential to significantly streamline the verification and validation (V and V) processes in NRS code development environments which are characterized by rapidly evolving software, many contributing developers and a large and growing body of validation data. In this paper, a survey of data conditioning and analysis techniques is summarized which focuses on their relevance to NRS code accuracy assessment. A number of methods are considered for their applicability to the automated assessment of the accuracy of NRS code simulations. A variety of data types and computational modeling methods are considered from a spectrum of mathematical and engineering disciplines. The goal of the survey was to identify needs, issues and techniques to be considered in the development of an automated code assessment procedure, to be used in United States Nuclear Regulatory Commission (NRC) advanced thermal-hydraulic T/H code consolidation efforts. The ACAP software was designed based in large measure on the findings of this survey. An overview of this tool is summarized and several NRS data applications are provided. The paper is organized as follows: The motivation for this work is first provided by background discussion that summarizes the relevance of this subject matter to the nuclear reactor industry. Next, the spectrum of NRS data types are classified into categories, in order to provide a basis for assessing individual comparison methods. Then, a summary of the survey is provided, where each

  19. Initiating Events for Multi-Reactor Plant Sites

    Energy Technology Data Exchange (ETDEWEB)

    Muhlheim, Michael David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Poore, III, Willis P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2014-09-01

    Inherent in the design of modular reactors is the increased likelihood of events that initiate at a single reactor affecting another reactor. Because of the increased level of interactions between reactors, it is apparent that the Probabilistic Risk Assessments (PRAs) for modular reactor designs need to specifically address the increased interactions and dependencies.

  20. Considerations in the development of safety requirements for innovative reactors: Application to modular high temperature gas cooled reactors

    International Nuclear Information System (INIS)

    2003-08-01

    Member States of the IAEA have frequently requested this organization to assess, at the conceptual stage, the safety of the design of nuclear reactors that rely on a variety of technologies and are of a high degree of innovation. However, to date, for advanced and innovative reactors and for reactors with characteristics that are different from those of existing light water reactors, widely accepted design standards and rules do not exist. This TECDOC is an outcome of the efforts deployed by the IAEA to develop a general approach for assessing the safety of the design of advanced and innovative reactors, and of all reactors in general including research reactors, with characteristics that differ from those of light water reactors. This publication puts forward a method for safety assessment that is based on the well established and accepted principle of defence in depth. The need to develop a general approach for assessing the safety of the design of reactors that applies to all kinds of advanced reactors was emphasized by the request to the IAEA by South Africa to review the safety of the South African pebble bed modular reactor. This reactor, as other modular high temperature gas cooled reactors (MHTGRs), adopts very specific design features such as the use of coated particle fuel. The characteristics of the fuel deeply affect the design and the safety of the plant, thereby posing several challenges to traditional safety assessment methods and to the application of existing safety requirements that have been developed primarily for water reactors. In this TECDOC, the MHTGR has been selected as a case study to demonstrate the viability of the method proposed. The approach presented is based on an extended interpretation of the concept of defence in depth and its link with the general safety objectives and fundamental safety functions as set out in 'Safety of Nuclear Power Plants: Design', IAEA Safety Standards No. NS-R.1, issued by the IAEA in 2000. The objective

  1. Assessment of martensitic steels for advanced fusion reactors

    International Nuclear Information System (INIS)

    Wareing, J.; Tavassoli, A.A.

    1995-01-01

    Martensitic steels are currently considered in Europe to be prime structural candidate materials for the first wall and breeding blanket of the DEMO fusion reactor. In this design, reactor power and wall loading will be significantly higher than those of an experimental reactor. ITER and will give rise to component operating temperatures in the range 250 to 550 0 C with neutron doses higher than 70 dpa. These conditions render austenitic stainless steel, which will be used in ITER, less favourable. Factors contributing to the promotion of martensitic steels are their excellent resistance to irradiation induced swelling, low thermal expansion and high thermal conductivity allied to advanced industrial maturity, compared to other candidate materials vanadium alloys. This paper described the development and optimisation of the steel and weld metal. Using data design rules generated on modified 9 Cr 1 Mo steel during its qualification as a steam generator material for the European Fast Reactor (EFR), interim design guidelines are formulated. Whilst the merits of the steel are validated, it is shown that irradiation embrittlement at low temperature, allied to the need for prolonged post-weld hat treatment and the long term creep response of welds remain areas of some concern. (author). 18 refs., 6 figs., 2 tabs

  2. An assessment of the low seismic risk of the inherently safe sodium advanced fast reactor (SAFR)

    International Nuclear Information System (INIS)

    Rutherford, P.D.

    1988-01-01

    A recent probabilistic risk assessment (PRA) of the sodium advanced fast reactor (SAFR) demonstrated the inherently low risk of advanced liquid-metal, pool-type fast reactors with inherent safety systems. As a result, it was recognized that external events, especially seismic events, may not only be a major contributor to risk (as shown in several LWR PRAs) but also may completely dominate the risk. Accordingly, a seismic risk assessment has been completed for SAFR, which resulted in a core damage frequency of 2 x 10 -7 /year and a large release frequency of 4 x 10 -9 /year. This paper reports that public health risk in terms of early fatality risk and latent fatality risk were also several orders of magnitude below the NRC safety goals and below recent LWR risks reported in NUREB/CR1150

  3. Assessment of Astrid reactor pit design options

    International Nuclear Information System (INIS)

    Verpoest, Thomas; Villedieu, Alexandre; Robin, Jean-Charles

    2014-01-01

    Answering the French Act of the 28. of June 2006 about nuclear materials and waste management, the ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) Project has the objectives to demonstrate the industrial feasibility based on identified domains (safety, operability, economy) of Sodium-cooled Fast Reactor and to perform transmutation demonstrations. The pre-conceptual design, started in 2010, considers several reactor pit design options. One of the objectives is to define a reference configuration for the ASTRID project which is able to answer safety and design requirements. The components addressed in this article are: the safety vessel and the Decay Heat Removal system through the main vessel. The core catcher associated to the different configurations studied in this article is an internal core catcher (inside the main vessel). This article deals with the different locations of the DHR through the main vessel and the type of the safety vessel (supported versus suspended vessel). These options are studied in order to establish the advantages and drawbacks of the different configurations in terms of economy, safety, In Service Inspection and Repair (ISIR), operability, robustness, and project risk (authors)

  4. Steam up over reactor policy

    International Nuclear Information System (INIS)

    Kovan, D.

    1976-01-01

    Britain is once more assessing its nuclear power programme in the light of recent forecasts that there is unlikely to be any growth in the demand for electricity for many years to come. This means that the extra costs of launching a commercially unproven reactor, the Steam Generating Heavy Water Reactor (SGHWR), will be an even greater burden than previously expected, because they would be spread over fewer reactors. Sir John Hill's reported assessment concludes that the present strategy would be the most expensive way of developing Britain's nuclear power programme; and under the circumstances, may not be the best option. The SGHWR programme will certainly be more expensive than either relaunching a programme of advanced gas-cooled reactors (AGRs), or building American designed pressurised water reactors (PWRs). Recent developments of the AGR and PWR's and their advantages in the present position are outlined. (U.K.)

  5. Evaluating the potential of concentrating solar power generation in Northwestern India

    International Nuclear Information System (INIS)

    Purohit, Ishan; Purohit, Pallav; Shekhar, Shashaank

    2013-01-01

    To accelerate the decarburization in the Indian power sector, concentrating solar power (CSP) needs to play an important role. CSP technologies have found significant space in the Jawaharlal Nehru National Solar Mission (JNNSM) of the Indian government in which 20,000 MW grid connected solar power projects have been targeted by 2022 with 50% capacity for CSP. In this study a preliminary attempt has been made to assess the potential of CSP generation in the Northwestern (NW) regions of India; which seems a high potential area as it has the highest annual solar radiation in India, favorable meteorological conditions for CSP and large amount of waste land. The potential of CSP systems in NW India is estimated on the basis of a detailed solar radiation and land resource assessment. The energy yield exercise has been carried out for the representative locations using System Advisor Model for four commercially available CSP technologies namely Parabolic Trough Collector (PTC), Central receiver system (CRS), Linear Fresnel Reflector (LFR) and Parabolic Dish System (PDS). The financial viability of CSP systems at different locations in NW India is also analyzed in this study. On the basis of a detailed solar radiation and land resource assessment, the maximum theoretical potential of CSP in NW India is estimated over 2000 GW taking into accounts the viability of different CSP technologies and land suitability criteria. The technical potential is estimated over 1700 GW at an annual direct normal incidence (DNI) over 1800 kW h/m 2 and finally, the economic potential is estimated over 700 GW at an annual DNI over 2000 kW h/m 2 in NW India. It is expected that in near future locations with lower DNI values could also become financially feasible with the development of new technologies, advancement of materials, economy of scale, manufacturing capability along with the enhanced policy measures etc. With an annual DNI over 1600 kW h/m 2 it is possible to exploit over 2000 GW CSP

  6. Assessment of nuclear reactor concepts for low power space applications

    Science.gov (United States)

    Klein, Andrew C.; Gedeon, Stephen R.; Morey, Dennis C.

    1988-01-01

    The results of a preliminary small reactor concepts feasibility and safety evaluation designed to provide a first order validation of the nuclear feasibility and safety of six small reactor concepts are given. These small reactor concepts have potential space applications for missions in the 1 to 20 kWe power output range. It was concluded that low power concepts are available from the U.S. nuclear industry that have the potential for meeting both the operational and launch safety space mission requirements. However, each design has its uncertainties, and further work is required. The reactor concepts must be mated to a power conversion technology that can offer safe and reliable operation.

  7. Probability safety assessment of the Kozloduy-5 and Kozloduy-6 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Boyadzhiev, A; Manchev, B [Risk Engineering Ltd., Sofia (Bulgaria)

    1996-12-31

    A probability safety assessment (PSA) of Level 1 (assessment of plant failures leading to the determination of core damage frequency) has been carried out for the NPP Kozloduy Units 5 and 6 (reactors WWER-1000). The scope of the study includes all significant accident initiators including seismic (earthquake) and fire initiators. Event trees for all initiators and fault trees for front line systems, support systems and major safety systems have been built. A distribution of the different initiators has been established as follows: internal initiators - 85%, seismic initiators - 5%, fire initiators- 10%. The loss of offsite power was identified as main contributor from the internal initiators with frequency 1,1.10{sup -4}/y. It is concluded that the safety functions of WWER-1000 are adequately covered by the safety systems. 4 refs., 2 tabs.

  8. Assessment of the high temperature fission chamber technology for the French fast reactor program

    Energy Technology Data Exchange (ETDEWEB)

    Jammes, C.; Filliatre, P.; Geslot, B.; Domenech, T.; Normand, S. [Commissariat a l' Energie Atomique, CEA (France)

    2011-07-01

    High temperature fission chambers are key instruments for the control and protection of the sodium-cooled fast reactor. First, the developments of those neutron detectors, which are carried out either in France or abroad are reviewed. Second, the French realizations are assessed with the use of the technology readiness levels in order to identify tracks of improvement. (authors)

  9. Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Executive summary: main report. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    1975-10-01

    Information is presented concerning the objectives and organization of the reactor safety study; the basic concepts of risk; the nature of nuclear power plant accidents; risk assessment methodology; reactor accident risk; and comparison of nuclear risks to other societal risks.

  10. South African safety assessment framework for the pebble bed modular reactor - HTR2008-58192

    International Nuclear Information System (INIS)

    Joubert, J.; Kohtz, N.; Coe, I.

    2008-01-01

    It is planned to construct a first of a kind Pebble Bed Modular Reactor (PBMR) in South Africa. A need has been recognized to accompany the licensing process for the PBMR with independent safety assessments to ensure that the safety case submitted by the applicant complies with the licensing requirements of the NNR. At the HTR 2006 Conference, the framework and major challenges on safety assessment that the South African National Nuclear Regulator (NNR) faces in developing and applying appropriate strategies and tools were presented. This paper discusses the current status of the various NNR assessment activities and describes how this will be considered in the NNR Final Report on the PBMR Safety Case. The traditional safety assessment process has been adapted to take into account the developmental nature of the project. By performing safety assessments, the designer and applicant must ensure that the design as proposed for construction and as-built meets the safety requirements defined by the regulatory framework. The regulator performs independent safety assessments, including independent analyses in areas deemed safety significant and potentially safety significant. The developmental nature of the project also led to the identification of a series of regulatory assessment activities preceding the formal assessment of the safety case. Besides an assessment of the resolution of Key Licensing Issues which have been defined in an early stage of the project and are discussed in /l/, these activities comprise the participation in an SAR Early Intervention Process, the execution of a regulatory HAZOP and the development of a regulatory assessment specification for the formal assessment of the safety case. This paper briefly describes these activities and their current status. During the last two years, significant progress was made with the development or adjustment of tools for the independent analysis by the regulator of the steady state core design, of the transient

  11. An assessment of the radiological consequences of accidents in research reactors

    International Nuclear Information System (INIS)

    Ferreira, N.L.D.

    1992-01-01

    This work analyses the radiological consequences of accidents in two types of research reactors: a 5 MWt open pool reactor and a 50 MWt PWR reactor. Two siting cases have been considered: the reactor located near to a large population center and sited in a rural area. The influence of several factors such as source term, meteorological conditions and population distribution have been considered in the present analysis. (author)

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-10-01

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

  14. Neutronic and thermal hydraulic assessment of fast reactor cooling by water of super critical parameters

    International Nuclear Information System (INIS)

    Baranaev, Yu. D.; Glebov, A. P.; Ukraintsev, V. F.; Kolesov, V. V.

    2007-01-01

    Necessity of essential improvement of competitiveness for reactors on light water determines development of new generation power reactors on water of super critical parameters. The main objective of these projects is reaching of high efficiency coefficients while decreasing of investment to NPP and simplification of thermal scheme and high safety level. International programme of IV generation in which super critical reactors present is already started. In the frame of this concept specific Super Critical Fast Reactor with tight lattice of pitch is developing by collaboration of the FEI and IATE. In present article neutronic and thermal hydraulic assessment of fast reactor with plutonium MOX fuel and a core with a double-path of super critical water cooling is presented (SCFR-2X). The scheme of double path of coolant via the core in which the core is divided by radius on central and periphery parts with approximately equal number of fuel assemblies is suggested. Periferia part is cooling while down coming coolant movement. At the down part of core into the mix chamber flows from the periphery assemblies joining and come to the inlet of the central part which is cooling by upcoming flow. Eight zone of different content of MOX fuel are used (4 in down coming and 4 in upcoming) sub zones. Calculation of fuel burn-up and approximate scheme of refueling is evaluated. Calculation results are presented and discussed

  15. Assessment of damage domains of the High-Temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    Flores, Alain; Izquierdo, José María; Tuček, Kamil; Gallego, Eduardo

    2014-01-01

    Highlights: • We developed an adequate model for the identification of damage domains of the HTTR. • We analysed an anticipated operational transient, using the HTTR5+/GASTEMP code. • We simulated several transients of the same sequence. • We identified the corresponding damage domains using two methods. • We calculated exceedance frequency using the two methods. - Abstract: This paper presents an assessment analysis of damage domains of the 30 MW th prototype High-Temperature Engineering Test Reactor (HTTR) operated by the Japan Atomic Energy Agency (JAEA). For this purpose, an in-house deterministic risk assessment computational tool was developed based on the Theory of Stimulated Dynamics (TSD). To illustrate the methodology and applicability of the developed modelling approach, assessment results of a control rod (CR) withdrawal accident during subcritical conditions are presented and compared with those obtained by the JAEA

  16. Probabilistic integrity assessment of pressure tubes in an operating pressurized heavy water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Young-Jin; Park, Heung-Bae [KEPCO E and C, 188 Gumi-dong, Bundang-gu, Seongnam-si, Gyeonggi-do 463-870 (Korea, Republic of); Lee, Jung-Min; Kim, Young-Jin [School of Mechanical Engineering, Sungkyunkwan University, 300 Chunchun-dong, Jangan-gu, Suwon-si, Gyeonggi-do 440-746 (Korea, Republic of); Ko, Han-Ok [Korea Institute of Nuclear Safety, 34 Gwahak-ro, Yuseong-gu, Daejeon-si 305-338 (Korea, Republic of); Chang, Yoon-Suk, E-mail: yschang@khu.ac.kr [Department of Nuclear Engineering, Kyung Hee University, 1 Seocheon-dong, Giheung-gu, Yongin-si, Gyeonggi-do 446-701 (Korea, Republic of)

    2012-02-15

    Even though pressure tubes are major components of a pressurized heavy water reactor (PHWR), only small proportions of pressure tubes are sampled for inspection due to limited inspection time and costs. Since the inspection scope and integrity evaluation have been treated by using a deterministic approach in general, a set of conservative data was used instead of all known information related to in-service degradation mechanisms because of inherent uncertainties in the examination. Recently, in order that pressure tube degradations identified in a sample of inspected pressure tubes are taken into account to address the balance of the uninspected ones in the reactor core, a probabilistic approach has been introduced. In the present paper, probabilistic integrity assessments of PHWR pressure tubes were carried out based on accumulated operating experiences and enhanced technology. Parametric analyses on key variables were conducted, which were periodically measured by in-service inspection program, such as deuterium uptake rate, dimensional change rate of pressure tube and flaw size distribution. Subsequently, a methodology to decide optimum statistical distribution by using a robust method adopting a genetic algorithm was proposed and applied to the most influential variable to verify the reliability of the proposed method. Finally, pros and cons of the alternative distributions comparing with corresponding ones derived from the traditional method as well as technical findings from the statistical assessment were discussed to show applicability to the probabilistic assessment of pressure tubes.

  17. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    1993-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1992 (April 1, 1992-March 31, 1993). The major Department's programs promoted in the year are the assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  18. High-temperature gas reactor (HTGR) market assessment, synthetic fuels analysis

    International Nuclear Information System (INIS)

    1980-08-01

    This study is an update of assessments made in TRW's October 1979 assessment of overall high-temperature gas-cooled reactor (HTGR) markets in the future synfuels industry (1985 to 2020). Three additional synfuels processes were assessed. Revised synfuel production forecasts were used. General environmental impacts were assessed. Additional market barriers, such as labor and materials, were researched. Market share estimates were used to consider the percent of markets applicable to the reference HTGR size plant. Eleven HTGR plants under nominal conditions and two under pessimistic assumptions are estimated for selection by 2020. No new HTGR markets were identified in the three additional synfuels processes studied. This reduction in TRW's earlier estimate is a result of later availability of HTGR's (commercial operation in 2008) and delayed build up in the total synfuels estimated markets. Also, a latest date for HTGR capture of a synfuels market could not be established because total markets continue to grow through 2020. If the nominal HTGR synfuels market is realized, just under one million tons of sulfur dioxide effluents and just over one million tons of nitrous oxide effluents will be avoided by 2020. Major barriers to a large synfuels industry discussed in this study include labor, materials, financing, siting, and licensing. Use of the HTGR intensifies these barriers

  19. Structural Integrity Assessment of Reactor Containment Subjected to Aircraft Crash

    International Nuclear Information System (INIS)

    Kim, Junyong; Chang, Yoonsuk

    2013-01-01

    When an accident occurs at the NPP, containment building which acts as the last barrier should be assessed and analyzed structural integrity by internal loading or external loading. On many occasions that can occur in the containment internal such as LOCA(Loss Of Coolant Accident) are already reflected to design. Likewise, there are several kinds of accidents that may occur from the outside of containment such as earthquakes, hurricanes and strong wind. However, aircraft crash that at outside of containment is not reflected yet in domestic because NPP sites have been selected based on the probabilistic method. After intentional aircraft crash such as World Trade Center and Pentagon accident in US, social awareness for safety of infrastructure like NPP was raised world widely and it is time for assessment of aircraft crash in domestic. The object of this paper is assessment of reactor containment subjected to aircraft crash by FEM(Finite Element Method). In this paper, assessment of structural integrity of containment building subjected to certain aircraft crash was carried out. Verification of structure integrity of containment by intentional severe accident. Maximum stress 61.21MPa of horizontal shell crash does not penetrate containment. Research for more realistic results needed by steel reinforced concrete model

  20. 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

  1. Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR): Data manual

    International Nuclear Information System (INIS)

    Gilbert, B.G.; Reece, W.J.; Gertman, D.I.; Gilmore, W.E.; Galyean, W.J.

    1990-12-01

    The Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) is an automated data base management system for processing and storing human error probability and hardware component failure data. The NUCLARR system software resides on an IBM (or compatible) personal computer. NUCLARR can furnish the end user with data inputs for both human and hardware reliability analysis in support of a variety of risk assessment activities. The NUCLARR system is documented is a five-volume series of reports. Volume V: Data Manual provides a hard-copy representation of all data and related information available within the NUCLARR system software. This document is organized in three sections. Part 1 is the summary description, which presents an overview of the NUCLARR system and data processing procedures. Part 2 contains all data and information relevant to the human error probability (HEP) data side of NUCLARR. Data and information for the hardware component failure data (HCFD) side are presented in Part 3. 7 refs

  2. Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR): Data manual

    International Nuclear Information System (INIS)

    Gilbert, B.G.; Reece, W.J.; Gertman, D.I.; Gilmore, W.E.; Galyean, W.J.

    1990-12-01

    The Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) is an automated data base management system for processing and storing human error probability and hardware component failure data. The NUCLARR system software resides on an IBM (or compatible) personal computer. NUCLARR can furnish the end user with data inputs for both human and hardware reliability analysis in support of a variety of risk assessment activities. The NUCLARR system is documented in a five-volume series of reports. Volume V: Data Manual provides a hard-copy representation of all data and related information available within the NUCLARR system software. This document is organized in three sections. Part 1 is the summary description, which presents an overview of the NUCLARR system and data processing procedures. Part 2 contains all data and information relevant to the human error probability (HEP) side of NUCLARR. Data and information for the hardware component failure data (HCFD) side are presented in Part 3. 7 refs., 1 fig

  3. Human performance analysis in the frame of probabilistic safety assessment of research reactors

    International Nuclear Information System (INIS)

    Farcasiu, Mita; Nitoi, Mirela; Apostol, Minodora; Turcu, I.; Florescu, Gh.

    2005-01-01

    Full text: The analysis of operating experience has identified the importance of human performance in reliability and safety of research reactors. In Probabilistic Safety Assessment (PSA) of nuclear facilities, human performance analysis (HPA) is used in order to estimate human error contribution to the failure of system components or functions. HPA is a qualitative and quantitative analysis of human actions identified for error-likely situations or accident-prone situations. Qualitative analysis is used to identify all man-machine interfaces that can lead to an accident, types of human interactions which may mitigate or exacerbate the accident, types of human errors and performance shaping factors. Quantitative analysis is used to develop estimates of human error probability as effects of human performance in reliability and safety. The goal of this paper is to accomplish a HPA in the PSA frame for research reactors. Human error probabilities estimated as results of human actions analysis could be included in system event tree and/or system fault tree. The achieved sensitivity analyses determine human performance sensibility at systematically variations both for dependencies level between human actions and for operator stress level. The necessary information was obtained from operating experience of research reactor TRIGA from INR Pitesti. The required data were obtained from generic data bases. (authors)

  4. Ligand-Controlled Chemoselective C(acyl)–O Bond vs C(aryl)–C Bond Activation of Aromatic Esters in Nickel Catalyzed C(sp2)–C(sp3) Cross-Couplings

    KAUST Repository

    Chatupheeraphat, Adisak

    2018-02-20

    A ligand-controlled and site-selective nickel catalyzed Suzuki-Miyaura cross-coupling reaction with aromatic esters and alkyl organoboron reagents as coupling partners was developed. This methodology provides a facile route for C(sp2)-C(sp3) bond formation in a straightforward fashion by successful suppression of the undesired β-hydride elimination process. By simply switching the phosphorus ligand, the ester substrates are converted into the alkylated arenes and ketone products, respectively. The utility of this newly developed protocol was demonstrated by its wide substrate scope, broad functional group tolerance and application in the synthesis of key intermediates for the synthesis of bioactive compounds. DFT studies on the oxidative addition step helped rationalizing this intriguing reaction chemoselectivity: whereas nickel complexes with bidentate ligands favor the C(aryl)-C bond cleavage in the oxidative addition step leading to the alkylated product via a decarbonylative process, nickel complexes with monodentate phosphorus ligands favor activation of the C(acyl)-O bond, which later generates the ketone product.

  5. Severe accident sequence assessment for boiling water reactors: program overview

    International Nuclear Information System (INIS)

    Fontana, M.H.

    1980-10-01

    The Severe Accident Sequence Assessment (SASA) Program was started at the Oak Ridge National Laboratory (ORNL) in June 1980. This report documents the initial planning, specification of objectives, potential uses of the results, plan of attack, and preliminary results. ORNL was assigned the Brown's Ferry Unit 1 Plant with the station blackout being the initial sequence set to be addressed. This set includes: (1) loss of offsite and onsite ac power with no coolant injection; and (2) loss of offsite and onsite ac power with high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) as long as dc power supply lasts. This report includes representative preliminary results for the former case

  6. Assessment and Application of the ROSE Code for Reactor Outage Thermal-Hydraulic and Safety Analysis

    International Nuclear Information System (INIS)

    Liang, Thomas K.S.; Ko, F.-K.; Dai, L.-C.

    2001-01-01

    The currently available tools, such as RELAP5, RETRAN, and others, cannot easily and correctly perform the task of analyzing the system behavior during plant outages. Therefore, a medium-sized program aiming at reactor outage simulation and evaluation, such as midloop operation (MLO) with loss of residual heat removal (RHR), has been developed. Important thermal-hydraulic processes involved during MLO with loss of RHR can be properly simulated by the newly developed reactor outage simulation and evaluation (ROSE) code. The two-region approach with a modified two-fluid model has been adopted to be the theoretical basis of the ROSE code.To verify the analytical model in the first step, posttest calculations against the integral midloop experiments with loss of RHR have been performed. The excellent simulation capacity of the ROSE code against the Institute of Nuclear Energy Research Integral System Test Facility test data is demonstrated. To further mature the ROSE code in simulating a full-sized pressurized water reactor, assessment against the WGOTHIC code and the Maanshan momentary-loss-of-RHR event has been undertaken. The successfully assessed ROSE code is then applied to evaluate the abnormal operation procedure (AOP) with loss of RHR during MLO (AOP 537.4) for the Maanshan plant. The ROSE code also has been successfully transplanted into the Maanshan training simulator to support operator training. How the simulator was upgraded by the ROSE code for MLO will be presented in the future

  7. Rhodium(III)-Catalyzed [3+2]/[5+2] Annulation of 4-Aryl 1,2,3-Triazoles with Internal Alkynes through Dual C(sp2)-H Functionalization.

    Science.gov (United States)

    Yang, Yuan; Zhou, Ming-Bo; Ouyang, Xuan-Hui; Pi, Rui; Song, Ren-Jie; Li, Jin-Heng

    2015-05-26

    A rhodium(III)-catalyzed [3+2]/[5+2] annulation of 4-aryl 1-tosyl-1,2,3-triazoles with internal alkynes is presented. This transformation provides straightforward access to indeno[1,7-cd]azepine architectures through a sequence involving the formation of a rhodium(III) azavinyl carbene, dual C(sp(2))-H functionalization, and [3+2]/[5+2] annulation. © 2015 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  8. Application of fuzzy set theory for safety culture and safety management assessment of Kartini research reactor

    International Nuclear Information System (INIS)

    Syarip; Hauptmanns, U.

    2000-01-01

    The safety culture status of nuclear power plant is usually assessed through interview and/or discussions with personnel and management in plant, and an assessment of the pertinent documentation. The approach for safety culture assessment described in IAEA Safety Series, make uses of a questionnaire composed of questions which require 'Yes' or 'No' as an answer. Hence, it is basically a check-list approach which is quite common for safety assessments in industry. Such a procedure ignores the fact that the expert answering the question usually has knowledge which goes far beyond a mere binary answer. Additionally, many situations cannot readily be described in such restricted terms. Therefore, it was developed a checklist consisting of questions which are formulated such that they require more than a simple 'yes' or 'no' as an answer. This allows one to exploit the expert knowledge of the analyst appropriately by asking him to qualify the degree of compliance of each of the topics examined. The method presented has proved useful in assessing the safety culture and quality of safety management of the research reactor. The safety culture status and the quality of safety management of Kartini research reactor is rated as 'average'. The method is also flexible and allows one to add questions to existing areas or to introduce new areas covering related topics

  9. 04 - Sodium cooled fast breeder fourth-generation reactors - The experimental reactor ALLEGRO, the other ways for fast breeder fourth-generation reactors

    International Nuclear Information System (INIS)

    2012-12-01

    The authors first present the technology of gas-cooled fast breeder reactors (basic principles, specific innovations, feasibility studies, fuel element, safety) and notably the ALLEGRO project (design options and expected performances, preliminary safety demonstration). Then, they present the lead-cooled fast-breeder reactor technology: interests and obstacles, return on experience, the issue of lead density, neutron assessment, transmutation potential, dosimetry, safety chemical properties and compatibility with the fuel, water, air and steels. The next part addresses the technology of molten-salt fast-breeder reactors: choice of the liquid fuel and geometry, reactor concept (difficulties, lack of past R and D), demonstration and demonstrators, international context

  10. Waste management for JAERI fusion reactors

    International Nuclear Information System (INIS)

    Tobita, K.; Nishio, S.; Konishi, S.; Jitsukawa, S.

    2004-01-01

    In the fusion reactor design study at Japan Atomic Energy Institute (JAERI), several waste management strategies were assessed. The assessed strategies are: (1) reinforced neutron shield to clear the massive ex-shielding components from regulatory control; (2) low aspect ratio tokamak to reduce the total waste; (3) reuse of liquid metal breeding material and neutron shield. Combining these strategies, the weight of disposal waste from a low aspect ratio reactor VECTOR is expected to be comparable with the metal radwaste from a light water reactor (∼4000 t)

  11. Assessment of the thorium and uranium fuel cycle in the fast breeder and the high temperature reactor

    International Nuclear Information System (INIS)

    Schikorr, W.M.

    1977-01-01

    This report assesses the fissile fuel economy of the uranium and thorium cycle in the advanced reactors currently under development, the fast breeder reactor (FBR) and the high temperature reactor (HTR). It is shown by means of detailed burnup calculations that replacing UO 2 with ThO 2 or Th-metal as the radial blanket breeding material will not have any significant imapct on the breeding and burnup properties of the FBR. A global, analytical investigation is performed to study the fissile fuel economy of the many fissile fuel cycles possible in the HTR. Here it is demonstrated that the optimum conversion ratio of CR 3 O 8 ) demands are evaluated for a country such as the FRG under the assumptions of different future reactor strategy scenarios. Here it is demonstrated that the employement of both HTRs and FBRs can lead to a practically resource independent energy supply system within the next 40 to 60 years. However only through the large scale employement of the fast breeder can the future nuclear resource requirements be assured. (orig.) [de

  12. Modular High Temperature Gas-Cooled Reactor Safety Basis and Approach

    Energy Technology Data Exchange (ETDEWEB)

    David Petti; Jim Kinsey; Dave Alberstein

    2014-01-01

    Various international efforts are underway to assess the safety of advanced nuclear reactor designs. For example, the International Atomic Energy Agency has recently held its first Consultancy Meeting on a new cooperative research program on high temperature gas-cooled reactor (HTGR) safety. Furthermore, the Generation IV International Forum Reactor Safety Working Group has recently developed a methodology, called the Integrated Safety Assessment Methodology, for use in Generation IV advanced reactor technology development, design, and design review. A risk and safety assessment white paper is under development with respect to the Very High Temperature Reactor to pilot the Integrated Safety Assessment Methodology and to demonstrate its validity and feasibility. To support such efforts, this information paper on the modular HTGR safety basis and approach has been prepared. The paper provides a summary level introduction to HTGR history, public safety objectives, inherent and passive safety features, radionuclide release barriers, functional safety approach, and risk-informed safety approach. The information in this paper is intended to further the understanding of the modular HTGR safety approach. The paper gives those involved in the assessment of advanced reactor designs an opportunity to assess an advanced design that has already received extensive review by regulatory authorities and to judge the utility of recently proposed new methods for advanced reactor safety assessment such as the Integrated Safety Assessment Methodology.

  13. A sensitivity analysis and assessment on the reactivity, economics and resorce implications of reactor systems and cycles with respect to uncertainity in nuclear data and other reactor parameters

    International Nuclear Information System (INIS)

    Quan, B.L.

    1980-01-01

    A general sensitivity analysis system for analyzing the effects of uncertainity in nuclear data and reactor parameters on fuel cycle economics, resources and physics has been developed. The sensitivity analysis has been performed on various reactor systems and cycles such as the thorium cycles, plutonium cycles, CANDU reactor fuel cycles and alternate once-through LWR cycles such as the 18 month cycle. Sensitivity coefficients were generated for a variety of materials pertinent to the LWR fuel cycle using a series of fast running codes developed for this purpose and running on a local PDP-15 computer. Their relative order of importance were assessed and the reasons explaining this difference were examined. This work is a result of EPRI project in determining the data needs for the LWR industry and should be valuable in identifying areas in which data improvements are worthwhile

  14. Assessment of management modes for graphite from reactor decommissioning

    International Nuclear Information System (INIS)

    White, I.F.; Smith, G.M.; Saunders, L.J.; Kaye, C.J.; Martin, T.J.; Clarke, G.H.; Wakerley, M.W.

    1984-01-01

    A technological and radiological assessment has been made of the management options for irradiated graphite wastes from the decommissioning of Magnox and advanced gas-cooled reactors. Detailed radionuclide inventories have been estimated, the main contribution being from activation of the graphite and its stable impurities. Three different packaging methods for graphite have been described; each could be used for either sea or land disposal, is logistically feasible and could be achieved at reasonable cost. Leaching tests have been carried out on small samples of irradiated graphite under a variety of conditions including those of the deep ocean bed; the different conditions had little effect on the observed leach rates of radiologically significant radionuclides. Radiological assessments were made of four generic options for disposal of packaged graphite: on the deep ocean bed, in deep geologic repositories at two different types of site, and by shallow land burial. Incineration of graphite was also considered, though this option presents logistical problems. With appropriate precautions during the lifetime of the Cobalt-60 content of the graphite, any of the options considered could give acceptably low doses to individuals, and all would merit further investigation in site-specific contexts

  15. Assessment of Feasibility of the Beneficial Use of Waste Heat from the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Donna P. Guillen

    2012-07-01

    This report investigates the feasibility of using waste heat from the Advanced Test Reactor (ATR). A proposed glycol waste heat recovery system was assessed for technical and economic feasibility. The system under consideration would use waste heat from the ATR secondary coolant system to preheat air for space heating of TRA-670. A tertiary coolant stream would be extracted from the secondary coolant system loop and pumped to a new plate and frame heat exchanger, where heat would be transferred to a glycol loop for preheating outdoor air in the heating and ventilation system. Historical data from Advanced Test Reactor operations over the past 10 years indicates that heat from the reactor coolant was available (when needed for heating) for 43.5% of the year on average. Potential energy cost savings by using the waste heat to preheat intake air is $242K/yr. Technical, safety, and logistics considerations of the glycol waste heat recovery system are outlined. Other opportunities for using waste heat and reducing water usage at ATR are considered.

  16. Assessment of different mechanisms of C-14 production in irradiated graphite of RBMK-1500 reactors

    International Nuclear Information System (INIS)

    Narkunas, Ernestas; Smaizys, Arturas; Poskas, Povilas; Kilda, Raimondas

    2010-01-01

    Two RBMK-1500 water-cooled graphite-moderated channel-type power reactors at the Ignalina Nuclear Power Plant (INPP) are under decommissioning now. The total mass of irradiated graphite in the cores of both units is more than 3600 tons. The main source of uncertainty in the numerical assessment of graphite activity is the uncertainty of the initial impurities content in graphite. Nitrogen is one of the most important impurities, having a large neutron capture cross-section. This impurity may become the dominant source of C-14 production. RBMK reactors graphite stacks operate in the cooling mixture of helium-nitrogen gases and this may additionally increase the quantity of the nitrogen impurity. In this paper the results of the numerical modelling of graphite activation for the INPP Unit I reactor are presented. In order to evaluate the C-14 activity dependence on the nitrogen impurity content, several cases with different nitrogen content were modelled taking into account initial nitrogen impurity quantities in the graphite matrix and possible nitrogen quantities entrapped in the graphite pores from cooling gases. (orig.)

  17. A feasibility assessment of nuclear reactor power system concepts for the NASA Growth Space Station

    Science.gov (United States)

    Bloomfield, H. S.; Heller, J. A.

    1986-01-01

    A preliminary feasibility assessment of the integration of reactor power system concepts with a projected growth Space Station architecture was conducted to address a variety of installation, operational, disposition and safety issues. A previous NASA sponsored study, which showed the advantages of Space Station - attached concepts, served as the basis for this study. A study methodology was defined and implemented to assess compatible combinations of reactor power installation concepts, disposal destinations, and propulsion methods. Three installation concepts that met a set of integration criteria were characterized from a configuration and operational viewpoint, with end-of-life disposal mass identified. Disposal destinations that met current aerospace nuclear safety criteria were identified and characterized from an operational and energy requirements viewpoint, with delta-V energy requirement as a key parameter. Chemical propulsion methods that met current and near-term application criteria were identified and payload mass and delta-V capabilities were characterized. These capabilities were matched against concept disposal mass and destination delta-V requirements to provide a feasibility of each combination.

  18. A feasibility assessment of nuclear reactor power system concepts for the NASA growth Space Station

    International Nuclear Information System (INIS)

    Bloomfield, H.S.; Heller, J.A.

    1986-01-01

    A preliminary feasibility assessment of the integration of reactor power system concepts with a projected growth space station architecture was conducted to address a variety of installation, operational, disposition and safety issues. A previous NASA sponsored study, which showed the advantages of space station related concepts, served as the basis for this study. A study methodology was defined and implemented to assess compatible combinations of reactor power installation concepts, disposal destinations, and propulsion methods. Three installation concepts that met a set of integration criteria were characterized from a configuration and operational viewpoint, with end-of-life disposal mass identified. Disposal destinations that met current aerospace nuclear safety criteria were identified and characterized from an operational and energy requirements viewpoint, with delta-V energy requirement as a key parameter. Chemical propulsion methods that met current and near-term application criteria were identified and payload mass and delta-V capabilities were characterized. These capabilities were matched against concept disposal mass and destination delta-V requirements to provide a feasibility of each combination

  19. Long-term assessment of contaminated articles from the Chernobyl reactor.

    Science.gov (United States)

    Alkhomashi, N; Monged, M H E

    2015-06-01

    The Chernobyl accident caused a release of radioactive materials from the reactor into the environment. This event contaminated people, their surroundings and their personal property, especially in the zone around the reactor. Among the affected individuals were British students who were studying in Minsk and Kiev at the time of the Chernobyl accident. These students were exposed to external and internal radiation, and the individuals' articles of clothing were contaminated. The primary objective of this study was to analyze a sample of this contaminated clothing 20 years after the accident using three different detectors, namely, a BP4/4C scintillation detector, a Min-Con Geiger-Müller tube detector and a high-purity germanium (HPGe) detector. The clothing articles were initially assessed and found not to be significantly contaminated. However, there were several hot spots of contamination in various regions of the articles. The net count rates for these hot spots were in the range of 10.00 ± 3.16 c/s to 41.00 ± 6.40 c/s when the BP4/4C scintillation detector was used. The HPGe detector was used to identify the radionuclides present in the clothing, and the results indicated that the only active radionuclide was (137)Cs because of this isotope's long half-life. Copyright © 2015 Elsevier Ltd. All rights reserved.

  20. Review of Thermal Materials for CSP Plants and LCOE Evaluation for Performance Improvement using Chilean Strategic Minerals: Lithium Salts and Copper Foams

    Directory of Open Access Journals (Sweden)

    Gustavo Cáceres

    2016-01-01

    Full Text Available The improvement of solar thermal technologies in emerging economies like Chile is particularly attractive because the country is endowed with one of the most consistently high solar potentials, lithium and copper reserves. In recent years, growing interests for lithium based salts and copper foams in application of thermal technologies could change the landscape of Chile transforming its lithium reserves and copper availability into competitive energy produced in the region. This study reviews the technical advantages of using lithium based salts—applied as heat storage media and heat transfer fluid—and copper foam/Phase Change Materials (PCM alternatives—applied as heat storage media—within tower and parabolic trough Concentrated Solar Power (CSP plants, and presents a first systematic evaluation of the costs of these alternatives based on real plant data. The methodology applied is based on material data base compilation of price and technical properties, selection of CSP plant and estimation of amount of required material, and analysis of Levelized Cost of Electricity (LCOE. Results confirm that some lithium based salts are effective in reducing the amount of required material and costs for the Thermal Energy Storage (TES systems for both plant cases, with savings of up to 68% and 4.14% in tons of salts and LCOE, respectively. Copper foam/PCM composites significantly increase thermal conductivity, decreasing the volume of the TES system, but costs of implementation are still higher than traditional options.

  1. Assessment of Current Inservice Inspection and Leak Monitoring Practices for Detecting Materials Degradation in Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, Michael T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Simonen, Fredric A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Muscara, Joseph [US Nuclear Regulatory Commission (NRC), Rockville, MD (United States); Doctor, Steven R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Kupperman, David S. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-09-01

    An assessment was performed to determine the effectiveness of existing inservice inspection (ISI) and leak monitoring techniques, and recommend improvements, as necessary, to the programs as currently performed for light water reactor (LWR) components. Information from nuclear power plant (NPP) aging studies and from the U. S. Nuclear Regulatory Commission’s Generic Aging Lessons Learned (GALL) report (NUREG-1801) was used to identify components that have already experienced, or are expected to experience, degradation. This report provides a discussion of the key aspects and parameters that constitute an effective ISI program and a discussion of the basis and background against which the effectiveness of the ISI and leak monitoring programs for timely detection of degradation was evaluated. Tables based on the GALL components were used to systematically guide the process, and table columns were included that contained the ISI requirements and effectiveness assessment. The information in the tables was analyzed using histograms to reduce the data and help identify any trends. The analysis shows that the overall effectiveness of the ISI programs is very similar for both boiling water reactors (BWRs) and pressurized water reactors (PWRs). The evaluations conducted as part of this research showed that many ISI programs are not effective at detecting degradation before its extent reached 75% of the component wall thickness. This work should be considered as an assessment of NDE practices at this time; however, industry and regulatory activities are currently underway that will impact future effectiveness assessments. A number of actions have been identified to improve the current ISI programs so that degradation can be more reliably detected.

  2. Conceptual innovations in hybrid reactors

    International Nuclear Information System (INIS)

    Greenspan, E.; Miley, G.H.

    1980-01-01

    A number of innovations in the conception of fusion-fission hybrid reactors, including the blanket, the fusion driver, the coupling of the fusion and the fission components as well as the application of hybrid reactors are described, and their feasibility assessed

  3. Uncertainty Quantification in the Reliability and Risk Assessment of Generation IV Reactors: Final Scientific/Technical Report

    Energy Technology Data Exchange (ETDEWEB)

    Vierow, Karen; Aldemir, Tunc

    2009-09-10

    The project entitled, “Uncertainty Quantification in the Reliability and Risk Assessment of Generation IV Reactors”, was conducted as a DOE NERI project collaboration between Texas A&M University and The Ohio State University between March 2006 and June 2009. The overall goal of the proposed project was to develop practical approaches and tools by which dynamic reliability and risk assessment techniques can be used to augment the uncertainty quantification process in probabilistic risk assessment (PRA) methods and PRA applications for Generation IV reactors. This report is the Final Scientific/Technical Report summarizing the project.

  4. Assessment of reactor vessel integrity (ARVI)

    Energy Technology Data Exchange (ETDEWEB)

    Sehgal, B.R. E-mail: sehgal@ne.kth.se; Theerthan, A.; Giri, A.; Karbojian, A.; Willschuetz, H.G.; Kymaelaeinen, O.; Vandroux, S.; Bonnet, J.M.; Seiler, J.M.; Ikkonen, K.; Sairanen, R.; Bhandari, S.; Buerger, M.; Buck, M.; Widmann, W.; Dienstbier, J.; Techy, Z.; Kostka, P.; Taubner, R.; Theofanous, T.; Dinh, T.N

    2003-04-01

    The cost-shared project ARVI (assessment of reactor vessel integrity) involves a total of nine organisations from Europe and USA. The objective of the ARVI Project is to resolve the safety issues that remain unresolved for the melt vessel interaction phase of the in-vessel progression of a severe accident. The work consists of experiments and analysis development. Four tests were performed in the EC-FOREVER Programme, in which failure was achieved in-vessels employing the French pressure vessel steel. The tests were analysed with the commercial code ANSYS-Multiphysics, and the codes SYSTUS+ and PASULA, and quite good agreement was achieved for the failure location. Natural convection experiments in stratified pools have been performed in the SIMECO and the COPO facilities, which showed that much greater heat is transferred downwards for immiscible layers or before layers mix. A model for gap cooling and a set of simplified models for the system codes have been developed. MVITA code calculations have been performed for the Czech and Hungarian VVERs, towards evaluation of the in-vessel melt retention accident management scheme. Tests have been performed at the ULPU facility with organised flow for vessel external cooling. Considerable enhancement of the critical heat flux (CHF) was obtained. The ARVI Project has reached the halfway stage. This paper presents the results obtained thus far from the project.

  5. Dragon project reference design assessment study for a 528 MW (E) thorium cycle high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Hosegood, S.B.

    1967-05-01

    The report presents an assessment of the feasibility, safety and cost of a large nuclear power station employing a high temperature gas-cooled reactor. A thermal output 1250 MW was chosen for the study, resulting in a net electrical output of 528.34 MW from a single reactor station, or 1056.7 MW from a twin reactor station. A reference design has been developed and is described. The reactor uses a U-235/Th-232/U-233 fuel cycle, on a feed and breed basis. It is believed that such a reactor could be built at an early date, requiring only a relatively modest development programme. Building costs are estimated to be Pound46.66/kW for a single unit station and Pound42.6/kW for a twin station, with power generation costs of 1.67p/kWh and 1.50p/kWh respectively. Optimisation studies have not been carried out and it should be possible to improve on the costs. The design has been made as flexible as possible to allow units of smaller or larger outputs to be designed with a minimum of change. (U.K.)

  6. Control of SHARON reactor for autotrophic nitrogen removal in two-reactor configuration

    DEFF Research Database (Denmark)

    Valverde Perez, Borja; Mauricio Iglesias, Miguel; Sin, Gürkan

    2012-01-01

    With the perspective of investigating a suitable control design for autotrophic nitrogen removal, this work explores the control design for a SHARON reactor. With this aim, a full model is developed, including the pH dependency, in order to simulate the reactor and determine the optimal operating...... conditions. Then, the screening of controlled variables and pairing is carried out by an assessment of the effect of the disturbances based on the closed loop disturbance gain plots. Two controlled structures are obtained and benchmarked by their capacity to reject the disturbances before the Anammox reactor....

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

    International Nuclear Information System (INIS)

    1990-01-01

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

  8. Development and methodology of level 1 probability safety assessment at PUSPATI TRIGA Reactor

    International Nuclear Information System (INIS)

    Maskin, Mazleha; Tom, Phongsakorn Prak; Lanyau, Tonny Anak; Saad, Mohamad Fauzi; Ismail, Ahmad Razali; Abu, Mohamad Puad Haji; Brayon, Fedrick Charlie Matthew; Mohamed, Faizal

    2014-01-01

    As a consequence of the accident at the Fukushima Dai-ichi Nuclear Power Plant in Japan, the safety aspects of the one and only research reactor (31 years old) in Malaysia need be reviewed. Based on this decision, Malaysian Nuclear Agency in collaboration with Atomic Energy Licensing Board and Universiti Kebangsaan Malaysia develop a Level-1 Probability Safety Assessment on this research reactor. This work is aimed to evaluate the potential risks of incidents in RTP and at the same time to identify internal and external hazard that may cause any extreme initiating events. This report documents the methodology in developing a Level 1 PSA performed for the RTP as a complementary approach to deterministic safety analysis both in neutronics and thermal hydraulics. This Level-1 PSA work has been performed according to the procedures suggested in relevant IAEA publications and at the same time numbers of procedures has been developed as part of an Integrated Management System programme implemented in Nuclear Malaysia

  9. A design assessment of tritium removal systems for the mirror advanced reactor study

    International Nuclear Information System (INIS)

    Sood, S.K.; Kveton, O.K.

    1983-01-01

    This study investigates the available processes for removing tritium from light water, and selects the most appropriate process for recovering tritium from the various tritiated water streams identified in the Mirror Advanced Reactor Study (MARS). A simplified flowsheet is shown for the process and the main process parameters are identified. Previous experience is utilized to predict direct capital costs and power requirement for the Tritiated Water Removal Unit (TWRU). A number of possibilities are discussed for lowering the cost of the TWRU. An estimate is made of the direct capital cost for the Air Detritiation System that has already been selected as the reference design by MARS personnel. The leakage from the MARS coolant loop is estimated, based on the experience obtained with Ontario Hydro's coolant systems. Design targets are identified for tritium levels in the reactor hall atmosphere and in water and air emissions. Tritium levels are predicted for these and are assessed against the previously identified targets

  10. Development and methodology of level 1 probability safety assessment at PUSPATI TRIGA Reactor

    International Nuclear Information System (INIS)

    Mazleha Maskin; Phongsakorn, P.T.; Tonny, A.L.; Fedrick, C.M.B.; Faizal Mohamed; Mohamad Fauzi Saad; Ahmad Razali Ismail; Mohamad Puad Haji Abu

    2013-01-01

    Full-text: As a consequence of the accident at the Fukushima Dai-ichi Nuclear Power Plant in Japan, the safety aspects of the one and only research reactor (31 years old) in Malaysia need be reviewed. Based on this decision, Malaysian Nuclear Agency in collaboration with Atomic Energy Licensing Board and Universiti Kebangsaan Malaysia develop a Level-1 Probability Safety Assessment on this research reactor. This work is aimed to evaluate the potential risks of incidents in RTP and at the same time to identify internal and external hazard that may cause any extreme initiating events. This report documents the methodology in developing a Level 1 PSA performed for the RTP as a complementary approach to deterministic safety analysis both in neutronics and thermal hydraulics. This Level-1 PSA work has been performed according to the procedures suggested in relevant IAEA publications and at the same time numbers of procedures has been developed as part of an Integrated Management System programme implemented in Nuclear Malaysia. (author)

  11. Use of dwell time concept in fission product inventory assessment for CANDU reactors

    International Nuclear Information System (INIS)

    Bae, C.J.; Choi, J.H.; Hwang, H.R.; Seo, J.T.

    2003-01-01

    A realistic approach in calculating the initial fission product inventory within the CANFLEX-NU fuel has been assessed for its applicability to the single channel event safety analysis for CANDU reactors. This approach is based on the dwell time concept in which the accident is assumed to occur at the dwell time when the summation of fission product inventory for all isotopes becomes largest. However, in the current conservative analysis, the maximum total inventory and the corresponding gap inventory for each isotope are used as the initial fission product inventories regardless of the accident initiation time. The fission product inventory analysis has been performed using ELESTRES code considering power histories and burnup of the fuel bundles in the limiting channel. The analysis results showed that the total fission product inventory is found to be largest at 20% dwell time. Therefore, the fission product inventory at 20% dwell time can be used as the initial condition for the single channel event for the CANDU 6 reactors. (author)

  12. Natural variation for responsiveness to flg22, flgII-28, and csp22 and Pseudomonas syringae pv. tomato in heirloom tomatoes.

    Directory of Open Access Journals (Sweden)

    Selvakumar Veluchamy

    Full Text Available Tomato (Solanum lycopersicum L. is susceptible to many diseases including bacterial speck caused by Pseudomonas syringae pv. tomato. Bacterial speck disease is a serious problem worldwide in tomato production areas where moist conditions and cool temperatures occur. To enhance breeding of speck resistant fresh-market tomato cultivars we identified a race 0 field isolate, NC-C3, of P. s. pv. tomato in North Carolina and used it to screen a collection of heirloom tomato lines for speck resistance in the field. We observed statistically significant variation among the heirloom tomatoes for their response to P. s. pv. tomato NC-C3 with two lines showing resistance approaching a cultivar that expresses the Pto resistance gene, although none of the heirloom lines have Pto. Using an assay that measures microbe-associated molecular pattern (MAMP-induced production of reactive oxygen species (ROS, we investigated whether the heirloom lines showed differential responsiveness to three bacterial-derived peptide MAMPs: flg22 and flgII-28 (from flagellin and csp22 (from cold shock protein. Significant differences were observed for MAMP responsiveness among the lines, although these differences did not correlate strongly with resistance or susceptibility to bacterial speck disease. The identification of natural variation for MAMP responsiveness opens up the possibility of using a genetic approach to identify the underlying loci and to facilitate breeding of cultivars with enhanced disease resistance. Towards this goal, we discovered that responsiveness to csp22 segregates as a single locus in an F2 population of tomato.

  13. Development of an accident consequence assessment code for evaluating site suitability of light- and heavy-water reactors based on the Korean Technical standards

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Won Tae; Jeong, Hae Sung; Jeong, Hyo Joon; Kil, A Reum; Kim, Eun Han; Han, Moon Hee [Nuclear Environment Safety Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-12-15

    Methodologies for a series of radiological consequence assessments show a distinctive difference according to the design principles of the original nuclear suppliers and their technical standards to be imposed. This is due to the uncertainties of the accidental source term, radionuclide behavior in the environment, and subsequent radiological dose. Both types of PWR and PHWR are operated in Korea. However, technical standards for evaluating atmospheric dispersion have been enacted based on the U.S. NRC's positions regardless of the reactor types. For this reason, it might cause a controversy between the licensor and licensee of a nuclear power plant. It was modelled under the framework of the NRC Regulatory Guide 1.145 for light-water reactors, reflecting the features of heavy-water reactors as specified in the Canadian National Standard and the modelling features in MACCS2, such as atmospheric diffusion coefficient, ground deposition, surface roughness, radioactive plume depletion, and exposure from ground deposition. An integrated accident consequence assessment code, ACCESS (Accident Consequence Assessment Code for Evaluating Site Suitability), was developed by taking into account the unique regulatory positions for reactor types under the framework of the current Korean technical standards. Field tracer experiments and hand calculations have been carried out for validation and verification of the models. The modelling approaches of ACCESS and its features are introduced, and its applicative results for a hypothetical accidental scenario are comprehensively discussed. In an applicative study, the predicted results by the light-water reactor assessment model were higher than those by other models in terms of total doses.

  14. Electro-Kinetic Pumping with Slip Irreversibility in Heat Exchange of CSP-Powered Bio-Digester Assemblies

    Directory of Open Access Journals (Sweden)

    Emmanuel O.B. Ogedengbe

    2012-12-01

    Full Text Available Parametric studies of the effects of slip irreversibility in concentrating solar power (CSP-powered bio-digester assemblies are investigated. Complexities regarding the identification of the appropriate electro-kinetic phenomena for certain electrolyte phases are reviewed. The application of exergy analysis to the design of energy conversion devices, like solar thermal collectors, for the required heat of formation in a downdraft waste food bio-digester, is discussed. Thermal management in the silicon-based substrate of the energy system is analyzed. The rectangular-shaped micro-channels are simulated with a finite-volume, staggered coupling of the pressure-velocity fields. Entropy generation transport within the energy system is determined and coupled with the solution procedure. Consequently, the effects of channel size perturbation, Reynolds number, and pressure ratios on the thermal performance and exergy destruction are presented. A comparative analysis of the axial heat conduction for thermal management in energy conversion devices is proposed.

  15. Assessment of some interfacial shear correlations in a model of ECC bypass flow in PWR reactor downcomer

    International Nuclear Information System (INIS)

    Popov, N.K.; Rohatgi, U.S.

    1987-01-01

    The bypass/refill process in the PWR reactor downcomer, following a large rupture of a cold leg coolant supply pipe, is a complicated thermo-hydraulic two-phase flow phenomenon. Mathematical modeling of such phenomena is always accompanied with a difficult task of selection of suitable constitutive correlations. In a typically hydrodynamic phenomenon, like ECC refill process of the reactor lower plenum is considered, the phasic interfacial friction is the most influential constitutive correlation. Therefore, assessment of the well-known widely-used interfacial friction constitutive correlations in the model of ECC bypass/refill process, is the subject of this paper

  16. Research reactor support

    International Nuclear Information System (INIS)

    2005-01-01

    spent research reactor fuel to the country of origin under the U.S. Spent Fuel Acceptance Program and the Russian Research Reactor Fuel Return program. This includes the provision of handbooks on technical and administrative preparations for shipping the fuel, as well as training courses. In addition the IAEA provides evaluation of the current status, progress and trends of research reactor spent fuel storage projects or national programmes in this field, present proven technologies and/or organizational/managerial practices that can serve as models to solve specific issues. It also assists in specific areas such as: assessment of infrastructure required to plan and implement research reactor spent fuel storage (wet or dry), improvement of management practices, implementation of water quality programmes, implementation of corrosion surveillance programmes and assessment of costs associated with research reactors spent fuel storage

  17. A safety assessment of the use of graphite in nuclear reactors licensed by the US NRC

    International Nuclear Information System (INIS)

    Schweitzer, D.G.; Gurinsky, D.H.; Kaplan, E.; Sastre, C.

    1987-09-01

    This report reviews existing literature and knowledge on graphite burning and on stored energy accumulation and releases in order to assess what role, if any, a stored energy release can have in initiating or contributing to hypothetical graphite burning scenarios in research reactors. It also addresses the question of graphite ignition and self-sustained combustion in the event of a loss-of-coolant accident (LOCA). The conditions necessary to initiate and maintain graphite burning are summarized and discussed. From analyses of existing information it is concluded that only stored energy accumulations and releases below the burning temperature (650 0 C) are pertinent. After reviewing the existing knowledge on stored energy it is possible to show that stored energy releases do not occur spontaneously, and that the maximum stored energy that can be released from any reactor containing graphite is a very small fraction of the energy produced during the first few minutes of a burning incident. The conclusions from these analyses are that the potential to initiate or maintain a graphite burning incident is essentially independent of the stored energy in the graphite, and depends on other factors that are unique for these reactors, research reactors, and for Fort St. Vrain. In order to have self-sustained rapid graphite oxidation in any of these reactors, certain necessary conditions of geometry, temperature, oxygen supply, reaction product removal, and a favorable heat balance must be maintained. There is no new evidence associated with either the Windscale Accident or the Chernobyl Accident that indicates a credible potential for a graphite burning accident in any of the reactors considered in this review

  18. Comparative assessment of thermophysical and thermohydraulic characteristics of lead, lead-bismuth and sodium coolants for fast reactors

    International Nuclear Information System (INIS)

    2002-06-01

    All prototype, demonstration and commercial liquid metal cooled fast reactors (LMFRs) have used liquid sodium as a coolant. Sodium cooled systems, operating at low pressure, are characterised by very large thermal margins relative to the coolant boiling temperature and a very low structural material corrosion rate. In spite of the negligible thermal energy stored in the liquid sodium available for release in case of leakage, there is some safety concern because of its chemical reactivity with respect to air and water. Lead, lead-bismuth or other alloys of lead, appear to eliminate these concerns because the chemical reactivity of these coolants with respect to air and water is very low. Some experts believe that conceptually, these systems could be attractive if high corrosion activity inherent in lead, long term materials compatibility and other problems will be resolved. Extensive research and development work is required to meet this goal. Preliminary studies on lead-bismuth and lead cooled reactors and ADS (accelerator driven systems) have been initiated in France, Japan, the United States of America, Italy, and other countries. Considerable experience has been gained in the Russian Federation in the course of development and operation of reactors cooled with lead-bismuth eutectic, in particular, propulsion reactors. Studies on lead cooled fast reactors are also under way in this country. The need to exchange information on alternative fast reactor coolants was a major consideration in the recommendation by the Technical Working Group on Fast Reactors (TWGFRs) to collect, review and document the information on lead and lead-bismuth alloy coolants: technology, thermohydraulics, physical and chemical properties, as well as to make an assessment and comparison with respective sodium characteristics

  19. Assessment studies on plutonium recycle in CANDU reactors

    International Nuclear Information System (INIS)

    1978-11-01

    This paper describes the CANDU reactor system in detail and goes on to explore the potential for using the system with plutonium recycle fuelling to improve fuel utilisation and to meet the long-term challenge of economic supplies of nuclear fuel. The paper includes comments on costs and non-proliferation aspects. It concludes that: recycle fuelling is feasible with little modification to the reactor design and no degradation of safety, and could offer over 50% savings in uranium requirements. However, recycle fuelling costs do not appear competitive with natural uranium in the CANDU system under current economic conditions

  20. Storage capacity assessment of liquid fuels production by solar gasification in a packed bed reactor using a dynamic process model

    International Nuclear Information System (INIS)

    Kaniyal, Ashok A.; Eyk, Philip J. van; Nathan, Graham J.

    2016-01-01

    Highlights: • First analysis to assess storage requirements of a stand-alone packed bed, batch process solar gasifier. • 35 days of storage required for stand-alone solar system, whereas 8 h of storage required for hybrid system. • Sensitivity of storage requirement to reactor operation, solar region and solar multiple evaluated. - Abstract: The first multi-day performance analysis of the feasibility of integrating a packed bed, indirectly irradiated solar gasification reactor with a downstream FT liquids production facility is reported. Two fuel-loading scenarios were assessed. In one, the residual unconverted fuel at the end of a day is reused, while in the second, the residual fuel is discarded. To estimate a full year time-series of operation, a simplified statistical model was developed from short-period simulations of the 1-D heat transfer, devolatilisation and gasification chemistry model of a 150 kW th packed bed reactor (based on the authors’ earlier work). The short time-series cover a variety of solar conditions to represent seasonal, diurnal and cloud-induced solar transience. Also assessed was the influence of increasing the solar flux incident at the emitter plate of the packed bed reactor on syngas production. The combination of the annual time-series and daily model of syngas production was found to represent reasonably the seasonal transience in syngas production. It was then used to estimate the minimum syngas storage volume required to maintain a stable flow-rate and composition of syngas to a FT reactor over a full year of operation. This found that, for an assumed heliostat field collection area of 1000 m 2 , at least 64 days of storage is required, under both the Residual Fuel Re-Use and Discard scenarios. This figure was not sensitive to the two solar sites assessed, Farmington, New Mexico or Tonopah Airport, Nevada. Increasing the heliostat field collection area from 1000 to 1500 m 2 , led to an increase in the calculated daily rate

  1. Evaporation Basin Test Reactor Area, Idaho National Engineering Laboratory: Environmental assessment

    International Nuclear Information System (INIS)

    1991-12-01

    The Department of Energy (DOE) has prepared an environmental assessment (EA), DOE/EA-0501, on the construction and operation of the proposed Evaporation Basin at the Test Reactor Area (TRA) at the Idaho National Engineering Laboratory (INEL) near Idaho Falls, Idaho. Based on the analyses in the EA, DOE has determined that the proposed action is not a major Federal action significantly affecting the quality of the human environment, within the meaning of the National Environmental Policy Act (NEPA) of 1969. Therefore, the preparation of an environmental impact statement (EIS) is not required, and the Department is issuing this Finding of No Significant Impact

  2. Aerobic Asymmetric Dehydrogenative Cross-Coupling between Two C(sp3)-H Groups Catalyzed by a Chiral-at-Metal Rhodium Complex.

    Science.gov (United States)

    Tan, Yuqi; Yuan, Wei; Gong, Lei; Meggers, Eric

    2015-10-26

    A sustainable C-C bond formation is merged with the catalytic asymmetric generation of one or two stereocenters. The introduced catalytic asymmetric cross-coupling of two C(sp3)-H groups with molecular oxygen as the oxidant profits from the oxidative robustness of a chiral-at-metal rhodium(III) catalyst and exploits an autoxidation mechanism or visible-light photosensitized oxidation. In the latter case, the catalyst serves a dual function, namely as a chiral Lewis acid for catalyzing enantioselective enolate chemistry and at the same time as a visible-light-driven photoredox catalyst. © 2015 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  3. Workshop on processing of nuclear data for use in power reactor pressure vessel lifetime assessment. Summary report

    International Nuclear Information System (INIS)

    Paviotti Corcuera, R.; Greenwood, L.R.; Muir, D.W.

    1999-02-01

    This document summarizes the contents of the workshop on processing of nuclear data for use in power reactor pressure vessel lifetime assessment. A short description of the main topics of the agenda, the list of participants and comments and recommendations are given. (author)

  4. Imaging Fukushima Daiichi reactors with muons

    Directory of Open Access Journals (Sweden)

    Haruo Miyadera

    2013-05-01

    Full Text Available A study of imaging the Fukushima Daiichi reactors with cosmic-ray muons to assess the damage to the reactors is presented. Muon scattering imaging has high sensitivity for detecting uranium fuel and debris even through thick concrete walls and a reactor pressure vessel. Technical demonstrations using a reactor mockup, detector radiation test at Fukushima Daiichi, and simulation studies have been carried out. These studies establish feasibility for the reactor imaging. A few months of measurement will reveal the spatial distribution of the reactor fuel. The muon scattering technique would be the best and probably the only way for Fukushima Daiichi to make this determination in the near future.

  5. Assessing the influence of reactor system design criteria on the performance of model colon fermentation units.

    Science.gov (United States)

    Moorthy, Arun S; Eberl, Hermann J

    2014-04-01

    Fermentation reactor systems are a key platform in studying intestinal microflora, specifically with respect to questions surrounding the effects of diet. In this study, we develop computational representations of colon fermentation reactor systems as a way to assess the influence of three design elements (number of reactors, emptying mechanism, and inclusion of microbial immobilization) on three performance measures (total biomass density, biomass composition, and fibre digestion efficiency) using a fractional-factorial experimental design. It was determined that the choice of emptying mechanism showed no effect on any of the performance measures. Additionally, it was determined that none of the design criteria had any measurable effect on reactor performance with respect to biomass composition. It is recommended that model fermentation systems used in the experimenting of dietary effects on intestinal biomass composition be streamlined to only include necessary system design complexities, as the measured performance is not benefited by the addition of microbial immobilization mechanisms or semi-continuous emptying scheme. Additionally, the added complexities significantly increase computational time during simulation experiments. It was also noted that the same factorial experiment could be directly adapted using in vitro colon fermentation systems. Copyright © 2013 The Society for Biotechnology, Japan. Published by Elsevier B.V. All rights reserved.

  6. Results of the mid-term assessment of the 'High Performance Light Water Reactor Phase 2' project

    International Nuclear Information System (INIS)

    Starflinger, J.; Schulenberg, T.; Marsault, P.

    2009-01-01

    The High Performance Light Water Reactor (HPLWR) is a Light Water Reactor (LWR) operating at supercritical pressure (p>22.1 MPa). In Europe, investigations on the HPLWR have been integrated into a joint research project, called High Performance Light Water Reactor Phase 2 (HPLWR Phase 2), which is co-funded by the European Commission. Within the second year of the project, the design of the reactor core, the pressure vessel and its internals have been analysed in detail by means of advanced codes and methods. The mechanical design has been assessed and shows that stresses inside components and possible deformations keep within acceptable limits. The neutronics and the flow inside the core have been investigated. The addition of a water layer in the reflector helps to flatten the radial power profile. The moderator flow path must be changed because of possible reverse flow in the gaps between the assemblies (downward flow). First calculations of transients showed an acceptable behaviour of the cladding temperatures. Material oxidation experiments were successfully performed. The auxiliary loop of the Supercritical Water Loop has been constructed. Heat transfer has been investigated numerically analysing heat transfer deterioration (HTD) and flow around fuel pins with wire wrap spacers. (author)

  7. Assessments of Water Ingress Accidents in a Modular High-Temperature Gas-Cooled Reactor

    International Nuclear Information System (INIS)

    Zhang Zuoyi; Dong Yujie; Scherer, Winfried

    2005-01-01

    Severe water ingress accidents in the 200-MW HTR-module were assessed to determine the safety margins of modular pebble-bed high-temperature gas-cooled reactors (HTR-module). The 200-MW HTR-module was designed by Siemens under the criteria that no active safety protection systems were necessary because of its inherent safe nature. For simulating the behavior of the HTR-module during severe water ingress accidents, a water, steam, and helium multiphase cavity model was developed and implemented in the dynamic simulator for nuclear power plants (DSNP) simulation system. Comparisons of the DSNP simulations incorporating these models with experiments and with calculations using the time-dependent neutronics and temperature dynamics code were made to validate the simulation. The analysis of the primary circuit showed that the maximum water concentration increase in the reactor core was 3 s). The water vaporization in the steam generator and characteristics of water transport from the steam generator to the reactor core would reduce the rate of water ingress into the reactor core. The analysis of a full cavitation of the feedwater pump showed that if the secondary circuit could be depressurized, the feedwater pump would be stopped by the full cavitation. This limits the water transported from the deaerator to the steam generator. A comprehensive simulation of the HTR-module power plant showed that the water inventory in the primary circuit was limited to ∼3000 kg. The nuclear reactivity increase caused by the water ingress would lead to a fast power excursion, which would be inherently counterbalanced by negative feedback effects. The integrity of the fuel elements, because the safety-relevant temperature limit of 1600 deg. C is not reached in any case, is not challenged

  8. Brookhaven leak reactor to close

    CERN Multimedia

    MacIlwain, C

    1999-01-01

    The DOE has announced that the High Flux Beam Reactor at Brookhaven is to close for good. Though the news was not unexpected researchers were angry the decision had been taken before the review to assess the impact of reopening the reactor had been concluded (1 page).

  9. Water desalination using different capacity reactors options

    International Nuclear Information System (INIS)

    Alonso, G.; Vargas, S.; Del Valle, E.; Ramirez, R.

    2010-01-01

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

  10. Gas reactor international cooperative program. HTR-synfuel application assessment

    Energy Technology Data Exchange (ETDEWEB)

    1979-09-01

    This study assesses the technical, environmental and economic factors affecting the application of the High Temperature Gas-Cooled Thermal Reactor (HTR) to: synthetic fuel production; and displacement of fossil fuels in other industrial and chemical processes. Synthetic fuel application considered include coal gasification, direct coal liquefaction, oil shale processing, and the upgrading of syncrude to motor fuel. A wide range of other industrial heat applications was also considered, with emphasis on the use of the closed-loop thermochemical energy pipeline to supply heat to dispersed industrial users. In this application syngas (H/sub 2/ +CO/sub 2/) is produced at the central station HTR by steam reforming and the gas is piped to individual methanators where typically 1000/sup 0/F steam is generated at the industrial user sites. The products of methanation (CH/sub 4/ + H/sub 2/O) are piped back to the reformer at the central station HTR.

  11. Gas reactor international cooperative program. HTR-synfuel application assessment

    International Nuclear Information System (INIS)

    1979-09-01

    This study assesses the technical, environmental and economic factors affecting the application of the High Temperature Gas-Cooled Thermal Reactor (HTR) to: synthetic fuel production; and displacement of fossil fuels in other industrial and chemical processes. Synthetic fuel application considered include coal gasification, direct coal liquefaction, oil shale processing, and the upgrading of syncrude to motor fuel. A wide range of other industrial heat applications was also considered, with emphasis on the use of the closed-loop thermochemical energy pipeline to supply heat to dispersed industrial users. In this application syngas (H 2 +CO 2 ) is produced at the central station HTR by steam reforming and the gas is piped to individual methanators where typically 1000 0 F steam is generated at the industrial user sites. The products of methanation (CH 4 + H 2 O) are piped back to the reformer at the central station HTR

  12. Prospects for the establishment of plutonium recycle in thermal reactors in the Foratom countries. Status and assessment

    International Nuclear Information System (INIS)

    Chamberlain, A.; Melches, C.

    1977-01-01

    The paper reviews the technical status of plutonium recycle in thermal reactors in the Foratom countries and assesses the prospect for it becoming established in the future with the implicit assumptions that uranium oxide reprocessing capacity will be installed commensurate with the projected programmes for thermal reactor installation and that there will be no insuperable environmental, security or safeguards obstacles to the use of plutonium as a fuel. It is argued that the feasibility of using plutonium as an alternative to 235 U as the fuel for thermal reactors, particularly LWRs, has been extensively demonstrated by a number of Foratom countries and the main problem areas are fuel fabrication and fuel reprocessing. Mixed-oxide fuel fabrication has been well established on the prototype plant scale using low-irradiation plutonium, but it is recognized that the future design of production-scale plants will need to cater for the significantly higher radiation levels from high burnup plutonium and meet stricter environmental requirements on operator dosage and waste arisings. The main constraint on the establishment of recycle up to now has been the lack of available plutonium owing to the absence of significant uranium-oxide fuel reprocessing capacity. An assessment of the plutonium arisings in Europe, based on the projected uranium-oxide reprocessing capacity, shows that by 1990 plutonium, surplus to FBR requirements, should be accumulating by about 10t/a, sufficient to fuel about 8000MW(e) of LWRs. A further constraint would then be the availability and technical problems of mixed-oxide reprocessing, which is one of the areas identified for international collaboration. It is concluded that whilst there is unlikely to be substantial recycle of plutonium in thermal reactors in the Foratom countries before the early 1990s, an incentive could possibly arise about that time. The strength of this incentive will depend on a number of factors including the status of

  13. Safety analysis for research reactors

    International Nuclear Information System (INIS)

    2008-01-01

    The aim of safety analysis for research reactors is to establish and confirm the design basis for items important to safety using appropriate analytical tools. The design, manufacture, construction and commissioning should be integrated with the safety analysis to ensure that the design intent has been incorporated into the as-built reactor. Safety analysis assesses the performance of the reactor against a broad range of operating conditions, postulated initiating events and other circumstances, in order to obtain a complete understanding of how the reactor is expected to perform in these situations. Safety analysis demonstrates that the reactor can be kept within the safety operating regimes established by the designer and approved by the regulatory body. This analysis can also be used as appropriate in the development of operating procedures, periodic testing and inspection programmes, proposals for modifications and experiments and emergency planning. The IAEA Safety Requirements publication on the Safety of Research Reactors states that the scope of safety analysis is required to include analysis of event sequences and evaluation of the consequences of the postulated initiating events and comparison of the results of the analysis with radiological acceptance criteria and design limits. This Safety Report elaborates on the requirements established in IAEA Safety Standards Series No. NS-R-4 on the Safety of Research Reactors, and the guidance given in IAEA Safety Series No. 35-G1, Safety Assessment of Research Reactors and Preparation of the Safety Analysis Report, providing detailed discussion and examples of related topics. Guidance is given in this report for carrying out safety analyses of research reactors, based on current international good practices. The report covers all the various steps required for a safety analysis; that is, selection of initiating events and acceptance criteria, rules and conventions, types of safety analysis, selection of

  14. Artificial intelligence in nuclear reactor operation

    International Nuclear Information System (INIS)

    Da Ruan; Benitez-Read, J.S.

    2005-01-01

    Assessment of four real fuzzy control applications at the MIT research reactor in the US, the FUGEN heavy water reactor in Japan, the BR1 research reactor in Belgium, and a TRIGA Mark III reactor in Mexico will be examined through a SWOT analysis (strengths, weakness, opportunities, and threats). Special attention will be paid to the current cooperation between the Belgian Nuclear Research Centre (SCK·CEN) and the Mexican Nuclear Centre (ININ) on AI-based intelligent control for nuclear reactor operation under the partial support of the National Council for Science and Technology of Mexico (CONACYT). (authors)

  15. Cavum Septum Pellucidum in Retired American Pro-Football Players

    OpenAIRE

    Gardner, Raquel C.; Hess, Christopher P.; Brus-Ramer, Marcel; Possin, Katherine L.; Cohn-Sheehy, Brendan I.; Kramer, Joel H.; Berger, Mitchel S.; Yaffe, Kristine; Miller, Bruce; Rabinovici, Gil D.

    2016-01-01

    Previous studies report that cavum septum pellucidum (CSP) is frequent among athletes with a history of repeated traumatic brain injury (TBI), such as boxers. Few studies of CSP in athletes, however, have assessed detailed features of the septum pellucidum in a case-control fashion. This is important because prevalence of CSP in the general population varies widely (2% to 85%) between studies. Further, rates of CSP among American pro-football players have not been described previously. We sou...

  16. Decision-support tool for assessing future nuclear reactor generation portfolios

    International Nuclear Information System (INIS)

    Jain, Shashi; Roelofs, Ferry; Oosterlee, Cornelis W.

    2014-01-01

    Capital costs, fuel, operation and maintenance (O and M) costs, and electricity prices play a key role in the economics of nuclear power plants. Often standardized reactor designs are required to be locally adapted, which often impacts the project plans and the supply chain. It then becomes difficult to ascertain how these changes will eventually reflect in costs, which makes the capital costs component of nuclear power plants uncertain. Different nuclear reactor types compete economically by having either lower and less uncertain construction costs, increased efficiencies, lower and less uncertain fuel cycles and O and M costs etc. The decision making process related to nuclear power plants requires a holistic approach that takes into account the key economic factors and their uncertainties. We here present a decision-support tool that satisfactorily takes into account the major uncertainties in the cost elements of a nuclear power plant, to provide an optimal portfolio of nuclear reactors. The portfolio so obtained, under our model assumptions and the constraints considered, maximizes the combined returns for a given level of risk or uncertainty. These decisions are made using a combination of real option theory and mean–variance portfolio optimization. - Highlights: • Decisions to continue or abandon the construction of NPPs • Mean–variance portfolio of nuclear reactors • Sensitivity study of mean–variance portfolio of nuclear reactors

  17. Radiological impact assessment of the shut-down Salaspils nuclear reactor

    International Nuclear Information System (INIS)

    Riekstina, D.; Berzins, J.; Veveris, O.; Alksnis, J.

    2004-01-01

    The aim of the present work is to gain an overview about the background level of radioactivity and gamma radiation in the 3x3 km area around the Salaspils (Latvia) nuclear reactor after its shutting down. The ultimate design of the project is to assess the impact environmental background level during its 37 years long working time. For this purpose we have carried out: 1) the determination of radioactivity in soils; 2) the determination of radioactivity in groundwater; 3) the measurement of gamma-ray background in the checkpoints. The net density for the collection of soil samples (5 cm thick layer was gathered) and the gamma background measuring was 500x500 m and the total number of checkpoints was 113. The gamma-spectrometric analysis of the groundwater taken from 34 places: in the reactor territory (4-10 m depth) and from the wells of surrounding farms (8-12 m depth) was performed. The soil samples were dried at the temperature 105 0 C until the constant weight, and sifted. The high-resolution gamma spectrometry was used for measurement within the energy range of 50-2000 keV; the time of measuring - 20 hours. The uncertainty of measurements is within a range of 3-10%, but the minimal detectable activity - from 0.3 up to 1 Bq/kg. Cs-137 and natural radionuclides Th-232, U-238, K-40 were detected in soils. The concentration of Cs-137 varies in the range 0.3-227 Bq/kg or 20-1940 Bq/m 2 . It was established that the concentration of Cs-137 in neighbouring checkpoints can differ significantly. It could be explained by the type of soil and the collection place (coniferous or leafy forest, grassland, plough land etc.). The differences of the U-238, Th-232, and K-40 content in samples taken from various places are due to the type of soil and the fertilizers used. The concentration of these radionuclides is significantly lower in the turf. In all water samples the concentration of Cs-137 was lower than the minimal detectable activity. The determined radionuclide

  18. A licence renewal approach for the NRU research reactor

    International Nuclear Information System (INIS)

    Natalizio, A.; Gumley, P.

    1991-01-01

    Licence Renewal is not only a subject that is being addressed for power reactors, but it is one of immediate interest for a number of research facilities, world-wide. In Canada, research reactors and power reactors are issued an operating licence for a limited term (typically two years), hence, licence renewal is done on a regular basis. Therefore, licence renewal in the Canadian context is different than in the context of this topical meeting. The NRU research reactor facility is being assessed for a licence renewal beyond its original design life. This paper describes the licence renewal approach, the assessments being performed to establish the condition of the facility, and the Safety Assessment Basis which defines the requirements for licence renewal. The current status of the assessments is also described. (author)

  19. Capabilities and limitations of fracture mechanics methods in the assessment of integrity of light water reactor components

    Energy Technology Data Exchange (ETDEWEB)

    Burdekin, F M

    1988-12-31

    This document deals with fracture mechanics methods used for the assessment of Light Water Reactor (LWR) components. The background to analysis methods using elastic plastic parameters is described. Several results obtained with these methods are presented as well as results of reliability analysis methods. (TEC). 27 refs.

  20. Assessment of uncertainty in full core reactor physics calculations using statistical methods

    International Nuclear Information System (INIS)

    McEwan, C.

    2012-01-01

    The best estimate method of safety analysis involves choosing a realistic set of input parameters for a proposed safety case and evaluating the uncertainty in the results. Determining the uncertainty in code outputs remains a challenge and is the subject of a benchmarking exercise proposed by the Organization for Economic Cooperation and Development. The work proposed in this paper will contribute to this benchmark by assessing the uncertainty in a depletion calculation of the final nuclide concentrations for an experiment performed in the Fukushima-2 reactor. This will be done using lattice transport code DRAGON and a tool known as DINOSAUR. (author)

  1. Assessment of uncertainty in full core reactor physics calculations using statistical methods

    Energy Technology Data Exchange (ETDEWEB)

    McEwan, C., E-mail: mcewac2@mcmaster.ca [McMaster Univ., Hamilton, Ontario (Canada)

    2012-07-01

    The best estimate method of safety analysis involves choosing a realistic set of input parameters for a proposed safety case and evaluating the uncertainty in the results. Determining the uncertainty in code outputs remains a challenge and is the subject of a benchmarking exercise proposed by the Organization for Economic Cooperation and Development. The work proposed in this paper will contribute to this benchmark by assessing the uncertainty in a depletion calculation of the final nuclide concentrations for an experiment performed in the Fukushima-2 reactor. This will be done using lattice transport code DRAGON and a tool known as DINOSAUR. (author)

  2. Performance Assessment of Passive Gaseous Provisions (PGAP). Report of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)

    International Nuclear Information System (INIS)

    2013-07-01

    The International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) was launched in 2000 on the basis of IAEA General Conference resolution GC(44)/RES/21. INPRO helps to ensure the availability of sustainable nuclear energy in the 21st century and seeks to bring together all interested Member States - both technology holders and technology users - to consider joint actions to achieve desired innovations. To contribute to an international consensus on the definition of the reliability of passive systems that involve natural circulation, and on a methodology to assess this reliability, INPRO initiated a collaborative project on Performance Assessment of Passive Gaseous Provisions (PGAP) in 2007. Advanced nuclear reactor designs incorporate several passive systems in addition to active ones, not only to enhance the operational safety of the reactors but also to mitigate the consequences of a severe accident should one occur. However, the reliability of passive safety systems is crucial and must be assessed before they are used extensively in future nuclear power plants. Several physical parameters affect the performance of a passive safety system, and their values at the time of operation are a priori unknown. The functions of many passive systems are based on thermohydraulic principles, which until recently were considered as not being subject to any kind of failure. Hence, large and consistent efforts are required to quantify the reliability of such systems. Three participants from three INPRO Member States were involved in this collaborative project. Reliability methods for passive systems (RMPS) and assessment of passive system reliability (APSRA) methodologies were used by the participants to assess the performance and reliability of the passive decay heat removal system of the French gas cooled fast reactor design for station blackout and a loss of coolant accident combined with loss of off-site power, respectively. This publication presents the

  3. Analysis of dynamic stability and safety of reactor system by reactor simulator; Analiza dinamicke stabilnosti i sigurnosti reaktorskog sistema pomocu reaktorskog simulatora

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N [Boris Kidric Institute of Nuclear Sciences Vinca, Belgrade (Yugoslavia)

    1963-11-15

    In order to enable qualitative analysis of dynamic properties of reactors RA and RB, mathematical models of these reactors were formulated and adapted for solution on analog computer. This report contains basic assessments for creating the model and complete equations for each reactor. Model was used to analyse three possible accidents at the RA reactor and possible hypothetical accidents at the RB reactor.

  4. Replacement research reactor for Australia

    International Nuclear Information System (INIS)

    Miller, Ross

    1998-01-01

    In 1992, the Australian Government commissioned a review into the need for a replacement research reactor. That review concluded that in about years, if certain conditions were met, the Government could make a decision in favour of a replacement reactor. A major milestone was achieved when, on 3 September 1997, the Australian Government announced the construction of a replacement research reactor at the site of Australia's existing research reactor HIFAR, subject to the satisfactory outcome of an environmental assessment process. The reactor will be have the dual purpose of providing a first class facility for neutron beam research as well as providing irradiation facilities for both medical isotope production and commercial irradiations. The project is scheduled for completion before the end of 2005. (author)

  5. Probability safety assessment activities in India for new and advanced reactors

    International Nuclear Information System (INIS)

    Guptan, R.; Ghagde, S.G.; Nama, R.; Varde, P.V.; Vinod, G.; Arul, J.; Solanki, R.B.

    2012-01-01

    This paper discusses, in brief, the salient features of the Level 1 PSA for New and Advanced reactors in India. The features of Level 1 PSA for new reactors are being discussed through a case study of 540 MWe twin unit (comprises of Unit 3 and 4) PHWRs at TAPS. The reactors uses Heavy water moderator and pressurized heavy water coolant, natural uranium fuel and horizontal pressure tubes. The major feature of PSA of advanced reactors is also discussed through the specific issues that were encountered during PSA modeling of AHWR (Advanced Heavy Water Reactor) and 700 MWe PHWR. The results of the PSA indicate that a fairly high level of redundancies exists in TAPS-3 and -4 design. It is recommended that staggered testing philosophy should be adopted especially for Emergency Core Cooling System, to reduce the probability of common cause failure among the motorized valves. It is also recommended to emphasize the importance of Small Break LOCA in general and their consequences in the licensing process of the plant operators

  6. Application of data analysis techniques to nuclear reactor systems code to accuracy assessment

    International Nuclear Information System (INIS)

    Kunz, R.F.; Kasmala, G.F.; Murray, C.J.; Mahaffy, J.H.

    2000-01-01

    An automated code assessment program (ACAP) has been developed by the authors to provide quantitative comparisons between nuclear reactor systems (NRS) code results and experimental measurements. This software was developed under subcontract to the United States Nuclear Regulatory Commission for use in its NRS code consolidation efforts. In this paper, background on the topic of NRS accuracy and uncertainty assessment is provided which motivates the development of and defines basic software requirements for ACAP. A survey of data analysis techniques was performed, focusing on the applicability of methods in the construction of NRS code-data comparison measures. The results of this review process, which further defined the scope, user interface and process for using ACAP are also summarized. A description of the software package and several sample applications to NRS data sets are provided. Its functionality and ability to provide objective accuracy assessment figures are demonstrated. (author)

  7. 75 FR 8412 - Office of New Reactors: Interim Staff Guidance on Assessing Ground Water Flow and Transport of...

    Science.gov (United States)

    2010-02-24

    ... NUCLEAR REGULATORY COMMISSION [NRC-2010-0047] Office of New Reactors: Interim Staff Guidance on Assessing Ground Water Flow and Transport of Accidental Radionuclide Releases; Solicitation of Public... ground water flow and transport of accidental radionuclide releases necessary to demonstrate compliance...

  8. Nuclear proliferation and civilian nuclear power. Report of the Nonproliferation Alternative Systems Assessment Program. Volume IX. Reactor and fuel cycle description

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-01

    The Nonproliferation Alterntive Systems Assessment Program (NASAP) has characterized and assessed various reactor/fuel-cycle systems. Volume IX provides, in summary form, the technical descriptions of the reactor/fuel-cycle systems studied. This includes the status of the system technology, as well as a discussion of the safety, environmental, and licensing needs from a technical perspective. This information was then used in developing the research, development, and demonstration (RD and D) program, including its cost and time frame, to advance the existing technology to the level needed for commercial use. Wherever possible, the cost data are given as ranges to reflect the uncertainties in the estimates.

  9. Nuclear proliferation and civilian nuclear power. Report of the Nonproliferation Alternative Systems Assessment Program. Volume IX. Reactor and fuel cycle description

    International Nuclear Information System (INIS)

    1980-06-01

    The Nonproliferation Alterntive Systems Assessment Program (NASAP) has characterized and assessed various reactor/fuel-cycle systems. Volume IX provides, in summary form, the technical descriptions of the reactor/fuel-cycle systems studied. This includes the status of the system technology, as well as a discussion of the safety, environmental, and licensing needs from a technical perspective. This information was then used in developing the research, development, and demonstration (RD and D) program, including its cost and time frame, to advance the existing technology to the level needed for commercial use. Wherever possible, the cost data are given as ranges to reflect the uncertainties in the estimates

  10. Solar Power Potential of Tanzania: Identifying CSP and PV Hot Spots through a GIS Multicriteria Decision Making Analysis

    DEFF Research Database (Denmark)

    Aly, Ahmed; Jensen, Steen Solvang; Pedersen, Anders Branth

    2017-01-01

    More than one billion people are still living without access to electricity today. More than half of them are living in Sub-Saharan Africa. There is a noticeable shortage of energy related information in Africa, especially for renewable energies. Due to lacking studies and researches on integrating...... renewable energy technologies, the Tanzanian official generation expansion plan till 2035 showed high dependency on fossil fuel and a negligible role of renewables other than large hydropower. This study investigates the spatial suitability for large-scale solar power installations in Tanzania through using...... technology-specific suitability map categorizes all the non-excluded areas into most suitable, suitable, moderately suitable, and least suitable areas. The study also suggests four hot spots (i.e. specific recommended locations) for Concentrated Solar Power (CSP) installations and four hot spots...

  11. Radiological controls and worker and public health and safety: An independent safety assessment of Department of Energy nuclear reactor facilities

    International Nuclear Information System (INIS)

    Tew, J.L.; Miles, M.E.; Knuth, D.; Boyd, R.

    1981-02-01

    DOE has formed a Nuclear Facilities Personnel Qualification and Training (NFPQT) Committee to assess the implications of the Report of the President's Commission on the Accident at Three Mile Island that are applicable to DOE's nuclear reactor operations. Thirteen DOE nuclear reactors were reviewed by the Committee. This report was prepared to provide a measure of how the radiological control and environmental practices at the 13 individual DOE reactor facilities measure up to (1) the recommendations contained in the Report of the President's Commission on the Accident at Three Mile Island, (2) the requirements and guidelines contained, and (3) the requirements of the applicable Title and Part of the Code of Federal Regulations

  12. Assessments of sheath strain and fission gas release data from 20 years of power reactor fuel irradiations

    International Nuclear Information System (INIS)

    Purdy, P.L.; Manzer, A.M.; Hu, R.H.; Gibb, R.A.; Kohn, E.

    1997-01-01

    Over the past 20 years, many fuel elements or bundles discharged from Canadian CANDU power reactors have been examined in the AECL hot cells. The post-irradiation examination (PIE) database covers a wide range of operating conditions, from which fuel performance characteristics can be assessed. In the present analysis, a PIE database was compiled representing elements from a total of 129 fuel bundles, of which 26% (34 bundles) were confirmed to have one or more defective elements. This comprehensive database was assessed in terms of measured sheath strain and fission gas release (FGR) for intact elements, in an attempt to identify any changes in these parameters over the history of CANDU reactor operation. Results from this assessment indicate that, for the data that are typical of normal CANDU operating conditions, tensile sheath strain and FGR have remained within 0.5% and 8%, respectively. Those data beyond these ranges are from fuel operated under abnormal conditions, not representative of normal operation, and thus do not indicate a trend toward unexpected fuel behaviour. The distributions of the PIE measurements indicate that maximum expected sheath strains and FGR for normally operated fuel are 0.7% and 13%, respectively. (author)

  13. Methodology for sodium fire vulnerability assessment of sodium cooled fast reactor based on the Monte-Carlo principle

    Energy Technology Data Exchange (ETDEWEB)

    Song, Wei [Nuclear and Radiation Safety Center, P. O. Box 8088, Beijing (China); Wu, Yuanyu [ITER Organization, Route de Vinon-sur-Verdon, 13115 Saint-Paul-lès-Durance (France); Hu, Wenjun [China Institute of Atomic Energy, P. O. Box 275(34), Beijing (China); Zuo, Jiaxu, E-mail: zuojiaxu@chinansc.cn [Nuclear and Radiation Safety Center, P. O. Box 8088, Beijing (China)

    2015-11-15

    Highlights: • Monte-Carlo principle coupling with fire dynamic code is adopted to perform sodium fire vulnerability assessment. • The method can be used to calculate the failure probability of sodium fire scenarios. • A calculation example and results are given to illustrate the feasibility of the methodology. • Some critical parameters and experience are shared. - Abstract: Sodium fire is a typical and distinctive hazard in sodium cooled fast reactors, which is significant for nuclear safety. In this paper, a method of sodium fire vulnerability assessment based on the Monte-Carlo principle was introduced, which could be used to calculate the probabilities of every failure mode in sodium fire scenarios. After that, the sodium fire scenario vulnerability assessment of primary cold trap room of China Experimental Fast Reactor was performed to illustrate the feasibility of the methodology. The calculation result of the example shows that the conditional failure probability of key cable is 23.6% in the sodium fire scenario which is caused by continuous sodium leakage because of the isolation device failure, but the wall temperature, the room pressure and the aerosol discharge mass are all lower than the safety limits.

  14. Methodology for sodium fire vulnerability assessment of sodium cooled fast reactor based on the Monte-Carlo principle

    International Nuclear Information System (INIS)

    Song, Wei; Wu, Yuanyu; Hu, Wenjun; Zuo, Jiaxu

    2015-01-01

    Highlights: • Monte-Carlo principle coupling with fire dynamic code is adopted to perform sodium fire vulnerability assessment. • The method can be used to calculate the failure probability of sodium fire scenarios. • A calculation example and results are given to illustrate the feasibility of the methodology. • Some critical parameters and experience are shared. - Abstract: Sodium fire is a typical and distinctive hazard in sodium cooled fast reactors, which is significant for nuclear safety. In this paper, a method of sodium fire vulnerability assessment based on the Monte-Carlo principle was introduced, which could be used to calculate the probabilities of every failure mode in sodium fire scenarios. After that, the sodium fire scenario vulnerability assessment of primary cold trap room of China Experimental Fast Reactor was performed to illustrate the feasibility of the methodology. The calculation result of the example shows that the conditional failure probability of key cable is 23.6% in the sodium fire scenario which is caused by continuous sodium leakage because of the isolation device failure, but the wall temperature, the room pressure and the aerosol discharge mass are all lower than the safety limits.

  15. Preliminary assessment of the effects of biaxial loading on reactor pressure vessel structural-integrity-assessment technology

    International Nuclear Information System (INIS)

    Pennell, W.E.; Bass, B.R.; Bryson, J.W.; Dickson, T.L.; McAfee, W.J.; Merkle, J.G.

    1996-01-01

    Effects of biaxial loading on shallow-flaw fracture toughness were studied to determine potential impact on structural integrity assessment of a reactor pressure vessel (RPV) under pressurized thermal shock (PTS) transient loading and pressure-temperature (PT) loading produced by reactor heatup and cooldown transients. Biaxial shallow-flaw fracture-toughness tests results were also used to determine the parameter controlling fracture in the transition temperature range, and to develop a related dual-parameter fracture-toughness correlation. Shallow-flaw and biaxial loading effects were found to reduce the conditional probability of crack initiation by a factor of nine when the shallow-flaw fracture-toughness K Jc data set, with biaxial-loading effects adjustments, was substituted in place of ASME Code K Ic data set in PTS analyses. Biaxial loading was found to reduce the shallow-flaw fracture toughness of RPV steel such that the lower-bound curve was located between ASME K Ic and K IR curves. This is relevant to future development of P-T curve analysis procedures. Fracture in shallow-flaw biaxial samples tested in the lower transition temperature range was shown to be strain controlled. A strain-based dual-parameter fracture-toughness correlation was developed and shown to be capable of predicting the effect of crack-tip constraint on fracture toughness for strain-controlled fracture

  16. A Global Assessment of Fast Reactors in the Future

    International Nuclear Information System (INIS)

    Devezeaux de Lavergne, J-G.; Mathonnière, G.

    2013-01-01

    Conclusions: • Fast reactors are the only way to fully achieve nuclear sustainability. • The SFR market cannot exist if a recycling market is not already present. • SFR has many other advantages that clearly outwheight the disadvantages (this trend is increasing). • Large data uncertainties (on uranium resources, world nuclear fleet deployment) return the little precise period at which economic competitiveness will be reached. Anyway, it is most likely to occur sometime in the second half of the century. • However, the market will start earlier, as it is splitted in two phases: before and after the economic competitiveness (this event is in fact country-dependant): – In the first phase 0-2 reactors will be built every year; – In the second phase up to 10-15 reactors will be built every year. • It is rather probable that there will be no more than two or three different Gen IV technologies in the world, because of the market size

  17. An assessment the severe accident equipment survivability for the Korean Next Generation Reactor

    International Nuclear Information System (INIS)

    Lee, B. C.; Moon, Y. T.; Park, J. W.; Kho, H. J.; Lee, S. W.

    1999-01-01

    One of the prominent design approaches to cope with the severe accident challenges in the Korean Next Generation Reactor is an assessment of equipment survivability in the severe accident environment at early design stage. In compliance with 10CFR50.34(f) and SECY-93-087, this work addresses that a reasonable level of assurance be provided to demonstrate that sufficient instrumentation and equipment will survive the consequences of a severe accident and will be available so that the operator may recover from and trend severe core damage sequences, including those scenarios which result in 100 percent oxidation of the active fuel cladding. An analytical and systematic approach was used to identify the equipment and instrumentation of safety-function and define severe accident environments including temperature, pressure, humidity, and radiation before and after the reactor vessel breach. As a result, it was concluded that with minor exceptions, existing design basis equipment qualification methods are sufficient to provide a reasonable level of assurance that this equipment will function during a severe accident. Furthermore, supplemental severe accident equipment and instrument procurement requirements were identified. (author)

  18. Technology assessment HTR. Part 4. Power upscaling of High Temperature Reactors

    International Nuclear Information System (INIS)

    Van Heek, A.I.

    1996-06-01

    Designs of nuclear reactors can be classified in evolutionary, revolutionary and innovative designs. An innovative design is the High Temperature Reactor (HTR). Introduction of innovative reactors has not been successful until now. Globally, three requirements for this reactors for successful market introduction can be identified: (1) Societal support for nuclear energy, or if separable, for this reactor type, should be repaired; (2) After market introduction the innovative plant must be able to operate economically competitive; and (3) The costs of market introduction of an innovative reactor design must be limited. Until now all reactor designs classified as innovative have not yet been realized. High temperature reactors exist in many different designs. Common features are: helium coolant, graphite moderator and coated particle fuel. The combination of these creates the potential to fulfill the first requirement (public support), and similarly a hurdle to the second requirement (economical operation). All three problems existing in the eyes of the public are addressed, while a high degree of transparency is reached, making the design understandable also by others than nuclear experts. A consequence of designing according to the social support requirement is a limitation of the unit power level. The usual method to make nuclear power plants economically competitive, i.e. just raising the power level (economy of scale) could not be applied anymore. Therefore other means of cost decreasing had to be used: modularization and simplification. These ideas are explained. Since all existing HTRs are currently out of operation, additional experience from two small HTRs under construction at this moment in the Far East will be essential. In the history of HTR designs, an evolutionary path can be identified. The early designs had a philosophy of safety and economics very similar to those of LWR. Modularization was introduced to attain economic viability and the design was

  19. Assessment of radiation fields from neutron irradiated structural components of the 40 MW research reactor CIRUS

    International Nuclear Information System (INIS)

    Sankaranarayanan, S.; Sharma, S.K.

    1993-01-01

    The paper summarizes the results of an assessment of the radiation fields from the long-lived neutron activation products (including the decay chain products) in the various structural components of the CIRUS reactor. Special attention is given for the analysis of neutron activation of impurity elements present in the materials of the structure. 16 refs, 4 figs, 4 tabs

  20. The TEX-I real-time expert system, applied to situation assessment for the SNR-300 reactor

    International Nuclear Information System (INIS)

    Schmal, N.; Leder, H.J.; Schade, H.J.

    1988-01-01

    Interatom, a subsidiary company of Siemens, is developing expert systems for the technical domain. These systems are operating in various industrial applications like flexible manufacturing or plant configuration, based on a domain-specific expert system shell, developed by Interatom. Additional projects are focusing on real-time diagnostics, e.g., for nuclear power plants. The authors report in this paper about a diagnosis expert system for the liquid-metal fast breeder reactor SNR-300, which uses new real-time tools, developed within the German TEX-I project (technical expert systems for data interpretation, diagnosis, and process control). The purpose of the system is to support the reactor operators in assessing plant status in real time, based on readings from many sensors. By on-line connection to the process control computer, it can monitor all incoming signal values, check the consistency of data, continuously diagnose the current plant status, detect unusual trends prior to accidents, localize faulty components, and recommended operator response in abnormal conditions. In the present knowledge acquisition and test phase, the expert system is connected to a real-time simulation of the reactor. The simulator is based on a thermohydraulic code for simulation of the transient behavior of temperatures and flow rates in the reactor core, plena, pipes, pumps, valves, intermediate heat exchangers, and cooling components. Additionally, the system's response to an asynchronous operator interaction can be simulated