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Sample records for bwr proliferation resistance

  1. Enhancing BWR proliferation resistance fuel with minor actinides

    Science.gov (United States)

    Chang, Gray S.

    2009-03-01

    To reduce spent fuel for storage and enhance the proliferation resistance for the intermediate-term, there are two major approaches (a) increase the discharged spent fuel burnup in the advanced light water reactor- LWR (Gen-III Plus), which not only can reduce the spent fuel for storage, but also increase the 238Pu isotopes ratio to enhance the proliferation resistance, and (b) use of transuranic nuclides ( 237Np and 241Am) in the high burnup fuel, which can drastically increase the proliferation resistance isotope ratio of 238Pu/Pu. For future advanced nuclear systems, minor actinides (MA) are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. As a result, MAs play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. In the study, a typical boiling water reactor (BWR) fuel unit lattice cell model with UO 2 fuel pins will be used to investigate the effectiveness of minor actinide reduction approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance in the intermediate-term goal for future nuclear energy systems. To account for the water coolant density variation from the bottom (0.76 g/cm 3) to the top (0.35 g/cm 3) of the core, the axial coolant channel and fuel pin were divided to 24 nodes. The MA transmutation characteristics at different elevations were compared and their impact on neutronics criticality discussed. The concept of MARA, which involves the use of transuranic nuclides ( 237Np and/or 241Am), significantly increases the 238Pu/Pu ratio for proliferation resistance, as well as serves as a burnable absorber to hold-down the initial excess reactivity. It is believed that MARA can play an important role in

  2. Proliferation resistance fuel cycle technology

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. S.; Ko, W. I

    1999-02-01

    The issues of dual use in nuclear technology are analysed for nuclear fuel cycle with special focus on uranium enrichment and spent fuel reprocessing which are considered as the most sensitive components in terms of vulnerability to diversion. Technical alternatives to mitigrate the vulnerability, as has been analysed in depth during the NASAP and INFCE era in the late seventies, are reviewed to characterize the DUPIC fuel cycle alternative. On the other hand, the new realities in nuclear energy including the disposition of weapon materials as a legacy of cold war are recast in an angle of nuclear proliferation resistance and safeguards with a discussion on the concept of spent fuel standard concept and its compliance with the DUPIC fuel cycle technology. (author)

  3. Proliferation resistance: issues, initiatives and evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Pilat, Joseph F [Los Alamos National Laboratory

    2009-01-01

    The vision of a nuclear renaissance has highlighted the issue of proliferation resistance. The prospects for a dramatic growth in nuclear power may depend on the effectiveness of, and the resources devoted to, plans to develop and implement technologies and approaches that strengthen proliferation resistance. The GenIV International Forum (GIF) and others have devoted attention and resources to proliferation resistance. However, the hope of finding a way to make the peaceful uses of nuclear energy resistant to proliferation has reappeared again and again in the history of nuclear power with little practical consequence. The concept of proliferation resistance has usually focused on intrinsic (technological) as opposed to extrinsic (institutional) factors. However, if there are benefits that may yet be realized from reactors and other facilities designed to minimize proliferation risks, it is their coupling with effective safeguards and other nonproliferation measures that likely will be critical. Proliferation resistance has also traditionally been applied only to state threats. Although there are no technologies that can wholly eliminate the risk of proliferation by a determined state, technology can play a limited role in reducing state threats and perhaps in eliminating many non-state threats. These and other issues are not academic. They affect efforts to evaluate proliferation resistance, including the methodology developed by GIF's Proliferation Resistance and Physical Protection (PR&PP) Working Group as well as the proliferation resistance initiatives that are being pursued or may be developed in the future. This paper will offer a new framework for thinking about proliferation resistance issues, including the ways the output of the methodology could be developed to inform the decisions that states, the International Atomic Energy (IAEA) and others will have to make in order to fully realize the promise of a nuclear renaissance.

  4. Proliferation resistance design of a plutonium cycle (Proliferation Resistance Engineering Program: PREP)

    Energy Technology Data Exchange (ETDEWEB)

    Sorenson, R.J.; Roberts, F.P.; Clark, R.G.

    1979-01-19

    This document describes the proliferation resistance engineering concepts developed to counter the threat of proliferation of nuclear weapons in an International Fuel Service Center (IFSC). The basic elements of an International Fuel Service Center are described. Possible methods for resisting proliferation such as processing alternatives, close-coupling of facilities, process equipment layout, maintenance philosophy, process control, and process monitoring are discussed. Political and institutional issues in providing proliferation resistance for an International Fuel Service Center are analyzed. The conclusions drawn are (1) use-denial can provide time for international response in the event of a host nation takeover. Passive use-denial is more acceptable than active use-denial, and acceptability of active-denial concepts is highly dependent on sovereignty, energy dependence and economic considerations; (2) multinational presence can enhance proliferation resistance; and (3) use-denial must be nonprejudicial with balanced interests for governments and/or private corporations being served. Comparisons between an IFSC as a national facility, an IFSC with minimum multinational effect, and an IFSC with maximum multinational effect show incremental design costs to be less than 2% of total cost of the baseline non-PRE concept facility. The total equipment acquisition cost increment is estimated to be less than 2% of total baseline facility costs. Personnel costs are estimated to increase by less than 10% due to maximum international presence. 46 figures, 9 tables.

  5. Analysis of nuclear proliferation resistance reprocessing and recycling technologies

    Energy Technology Data Exchange (ETDEWEB)

    Patricia Paviet-Hartmann; Gary Cerefice; Marcela Stacey; Steven Bakhtiar

    2011-05-01

    The PUREX process has been progressively and continuously improved during the past three decades, and these improvements account for successful commercialization of reprocessing in a few countries. The renewed interest in nuclear energy and the international growth of nuclear electricity generation do not equate – and should not be equated -with increasing proliferation risks. Indeed, the nuclear renaissance presents a unique opportunity to enhance the culture of non-proliferation. With the recent revival of interest in nuclear technology, technical methods for prevention of nuclear proliferation are being revisited. Robust strategies to develop new advanced separation technologies are emerging worldwide for sustainability and advancement of nuclear energy with enhanced proliferation resistance. On the other hand, at this moment, there are no proliferation resistance advanced technologies. . Until now proliferation resistance as it applies to reprocessing has been focused on not separating a pure stream of weapons-usable plutonium. France, as an example, has proposed a variant of the PUREX process, the COEX TM process, which does not result on a pure plutonium product stream. A further step is to implement a process based on group extraction of actinides and fission products associated with a homogeneous recycling strategy (UNEX process in the US, GANEX process in France). Such scheme will most likely not be deployable on an industrial scale before 2030 or so because it requires intensive R&D and robust flowsheets. Finally, future generation recycling schemes will handle the used nuclear fuel in fast neutron reactors. This means that the plutonium throughput of the recycling process may increase. The need is obvious for advanced aqueous recycling technologies that are intrinsically more proliferation resistant than the commercial PUREX process. In this paper, we review the actual PUREX process along with the advanced recycling technologies that will enhance

  6. 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)

  7. Key instrumentation in BWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Laendner, Alexander; Stellwag, Bernhard; Fandrich, Joerg [AREVA NP GmbH, Erlangen (Germany)

    2011-01-15

    This paper describes water chemistry surveillance practices at boiling water reactor (BWR) power plants. The key instrumentation in BWR plants consists of on-line as well as off-line instrumentation. The chemistry monitoring and control parameters are predominantly based on two guidelines, namely the VGB Water Chemistry Guidelines and the EPRI Water Chemistry Guidelines. Control parameters and action levels specified in the VGB guideline are described. Typical sampling locations in BWR plants, chemistry analysis methods and water chemistry data of European BWR plants are summarized. Measurement data confirm the high quality of reactor water of the BWRs in Europe. (orig.)

  8. PROLIFERATION RESISTANCE AND PHYSICAL PROTECTION WORKING GROUP: METHODOLOGY AND APPLICATIONS

    Energy Technology Data Exchange (ETDEWEB)

    Bari R. A.; Whitlock, J.; Therios, I.U.; Peterson, P.F.

    2012-11-14

    We summarize the technical progress and accomplishments on the evaluation methodology for proliferation resistance and physical protection (PR and PP) of Generation IV nuclear energy systems. We intend the results of the evaluations performed with the methodology for three types of users: system designers, program policy makers, and external stakeholders. The PR and PP Working Group developed the methodology through a series of demonstration and case studies. Over the past few years various national and international groups have applied the methodology to nuclear energy system designs as well as to developing approaches to advanced safeguards.

  9. A Comparison of Proliferation Resistance Measures of Misuse Scenarios Using a Markov Approach

    Energy Technology Data Exchange (ETDEWEB)

    Yue,M.; Cheng, L.-Y.; Bari, R.

    2008-05-11

    Misuse of declared nuclear facilities is one of the important proliferation threats. The robustness of a facility against these threats is characterized by a number of proliferation resistance (PR) measures. This paper evaluates and compares PR measures for several misuse scenarios using a Markov model approach to implement the pathway analysis methodology being developed by the PR&PP (Proliferation Resistance and Physical Protection) Expert Group. Different misue strategies can be adopted by a proliferator and each strategy is expected to have different impacts on the proliferator's success. Selected as the probabilistic measure to represent proliferation resistance, the probabilities of the proliferator's success of misusing a hypothetical ESFR (Example Sodium Fast Reactor) facility system are calculated using the Markov model based on the pathways constructed for individual misuse scenarios. Insights from a comparison of strategies that are likely to be adopted by the proliferator are discussed in this paper.

  10. Model development for quantitative evaluation of proliferation resistance of nuclear fuel cycles

    Energy Technology Data Exchange (ETDEWEB)

    Ko, Won Il; Kim, Ho Dong; Yang, Myung Seung

    2000-07-01

    This study addresses the quantitative evaluation of the proliferation resistance which is important factor of the alternative nuclear fuel cycle system. In this study, model was developed to quantitatively evaluate the proliferation resistance of the nuclear fuel cycles. The proposed models were then applied to Korean environment as a sample study to provide better references for the determination of future nuclear fuel cycle system in Korea. In order to quantify the proliferation resistance of the nuclear fuel cycle, the proliferation resistance index was defined in imitation of an electrical circuit with an electromotive force and various electrical resistance components. The analysis on the proliferation resistance of nuclear fuel cycles has shown that the resistance index as defined herein can be used as an international measure of the relative risk of the nuclear proliferation if the motivation index is appropriately defined. It has also shown that the proposed model can include political issues as well as technical ones relevant to the proliferation resistance, and consider all facilities and activities in a specific nuclear fuel cycle (from mining to disposal). In addition, sensitivity analyses on the sample study indicate that the direct disposal option in a country with high nuclear propensity may give rise to a high risk of the nuclear proliferation than the reprocessing option in a country with low nuclear propensity.

  11. High Fidelity BWR Fuel Simulations

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Su Jong [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-08-01

    This report describes the Consortium for Advanced Simulation of Light Water Reactors (CASL) work conducted for completion of the Thermal Hydraulics Methods (THM) Level 3 milestone THM.CFD.P13.03: High Fidelity BWR Fuel Simulation. High fidelity computational fluid dynamics (CFD) simulation for Boiling Water Reactor (BWR) was conducted to investigate the applicability and robustness performance of BWR closures. As a preliminary study, a CFD model with simplified Ferrule spacer grid geometry of NUPEC BWR Full-size Fine-mesh Bundle Test (BFBT) benchmark has been implemented. Performance of multiphase segregated solver with baseline boiling closures has been evaluated. Although the mean values of void fraction and exit quality of CFD result for BFBT case 4101-61 agreed with experimental data, the local void distribution was not predicted accurately. The mesh quality was one of the critical factors to obtain converged result. The stability and robustness of the simulation was mainly affected by the mesh quality, combination of BWR closure models. In addition, the CFD modeling of fully-detailed spacer grid geometry with mixing vane is necessary for improving the accuracy of CFD simulation.

  12. Nuclear proliferation and civilian nuclear power: report of the Nonproliferation Alternative Systems Assessment Program. Volume II. Proliferation resistance

    Energy Technology Data Exchange (ETDEWEB)

    1979-12-01

    Volume II assesses proliferation resistance. Chapters are devoted to: assessment of civilian nuclear systems (once-through fuel-cycle systems, closed fuel cycle systems, research reactors and critical facilities); assessment of associated sensitive materials and facilities (enrichment, problems with storage of spent fuel and plutonium content, and reprocessing and refabrication facilities); and safeguards for alternative fuel cycles.

  13. Establishment of Assessment Methodology Improvement of IAEA INPRO Proliferation Resistance

    Energy Technology Data Exchange (ETDEWEB)

    Park, J. H.; Yang, M. S.; Song, K. C.; Ko, W. I.; Kim, H. D.; Kim, Y. I.; Rhee, B. W.; Kim, H. T.

    2008-03-15

    For the development of assessment methodology of acquisition and diversion pathway of nuclear material, the PR assessment methodology which has been developed by GEN-IV PR and PP group was reviewed regarding the acquisition and diversion pathway of nuclear material and we proposed the research areas to develop the model of acquisition and diversion pathway of nuclear material including misuse of fuel cycle facilities and one of the IAEA INPRO CRPs which is aiming to develop its model. From the present study, its preliminary model for acquisition and diversion pathway of nuclear material was obtained. For preliminary evaluation of DUPIC system using methodology of acquisition/diversion pathway of nm and review of pyro-processing system characteristics, the research direction and work procedure was established to develop the assessment methodology of User Requirement 4 of INPRO PR by; 1) selection of the possible pathway to acquire and divert the nuclear material of DUPIC system, 2) the analysis of selected pathway, 3) the development of the assessment methodology of robustness and multiplicity of an INS. And, the PR characteristics and process/material flow analysis of the Pyro-processing system were preliminarily studied. For establishment of R and D direction for an INS and supporting international cooperation research, the collaborative research project titled as 'Acquisition and Diversion pathway analysis of Proliferation Resistance' as one of activities of IAEA INPRO was proposed, since Korean Government decided to actively support the IAEA INPRO. In order to review and clarify the Terms of Reference (TOR) of a Korean Proposed Collaborative Project (ROK1), two INPRO Consultancy Meetings were held. Its results were presented at two INPRO Steering Committees and the finalized TOR of Korean Proposal submit the 12-th INPRO Steering Committee Meeting which was held Dec. 3-5 2007. Four participants including USA, Canada, China and European Community (EC

  14. Proliferation resistance assessment of various methods of spent nuclear fuel storage and disposal

    Science.gov (United States)

    Kollar, Lenka

    Many countries are planning to build or already are building new nuclear power plants to match their growing energy needs. Since all nuclear power plants handle nuclear materials that could potentially be converted and used for nuclear weapons, they each present a nuclear proliferation risk. Spent nuclear fuel presents the largest build-up of nuclear material at a power plant. This is a proliferation risk because spent fuel contains plutonium that can be chemically separated and used for a nuclear weapon. The International Atomic Energy Agency (IAEA) safeguards spent fuel in all non-nuclear weapons states that are party to the Non-Proliferation Treaty. Various safeguards methods are in use at nuclear power plants and research is underway to develop safeguards methods for spent fuel in centralized storage or underground storage and disposal. Each method of spent fuel storage presents different proliferation risks due to the nature of the storage method and the safeguards techniques that are utilized. Previous proliferation resistance and proliferation risk assessments have mainly compared nuclear material through the whole fuel cycle and not specifically focused on spent fuel storage. This project evaluates the proliferation resistance of the three main types of spent fuel storage: spent fuel pool, dry cask storage, and geological repository. The proliferation resistance assessment methodology that is used in this project is adopted from previous work and altered to be applicable to spent fuel storage. The assessment methodology utilizes various intrinsic and extrinsic proliferation-resistant attributes for each spent fuel storage type. These attributes are used to calculate a total proliferation resistant (PR) value. The maximum PR value is 1.00 and a greater number means that the facility is more proliferation resistant. Current data for spent fuel storage in the United States and around the world was collected. The PR values obtained from this data are 0.49 for

  15. BWR Steam Dryer Alternating Stress Assessment Procedures

    Energy Technology Data Exchange (ETDEWEB)

    Morante, R. J. [Brookhaven National Lab. (BNL), Upton, NY (United States); Hambric, S. A. [Brookhaven National Lab. (BNL), Upton, NY (United States); Ziada, S. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2016-12-01

    This report presents an overview of Boiling Water Reactor (BWR) steam dryer design; the fatigue cracking failures that occurred at the Quad Cities (QC) plants and their root causes; a history of BWR Extended Power Uprates (EPUs) in the USA; and a discussion of steam dryer modifications/replacements, alternating stress mechanisms on steam dryers, and structural integrity evaluations (static and alternating stress).

  16. Proliferation-Resistant Nuclear Power Systems: A Workshop on New Ideas

    Energy Technology Data Exchange (ETDEWEB)

    Schock, R.

    2000-03-01

    The workshop addressed a number of major questions and challenges surrounding the relationship between the future of nuclear power and the broader issue of proliferation of nuclear materials for weapons or other means of nuclear terrorism. This is but one of at least four issues facing the civilian nuclear power industry, the others of note being safety, economics, and environmental impacts including the final disposition of waste. Various authorities attach different levels of significance to these issues, at least some maintaining that proliferation is the greatest, but all agree that they must be examined in parallel. Workshop participants were asked to consider several questions: What do we mean by nuclear proliferation and proliferation resistance? What metrics are useful for assessing proliferation resistance? What are meaningful goals and solutions? Can nuclear power systems and/or sub-systems be developed that are more resistant to proliferation than those in existence or being planned today? What are the barriers to the implementation of such systems? Can these solutions be applied to research, test, and isotope-production reactors?

  17. New concept of proliferation resistant sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Eliseev, V.A.; Krivitski, I.Y.; Matveev, V.I.; Popov, E.P.; Savitski, V.I.; Tsikunov, A.G. [Institute for Physics and Power Engineering, Obninsk (Russian Federation)

    2001-07-01

    The full text follows. It is proposed the concept of BN-800 sodium cooled fast reactor operating in the closed fuel cycle with special reprocessing technology. The use of nitride fuel allows improving the parameters of reactor safety (internal breeding {approx}1, zero value of sodium void reactivity effect), economy (one refueling per year), ecology (use of nitride enriched by nitrogen-15) and non-proliferation (use of reprocessing without separating the plutonium from uranium). The main difficulty of this type reactor development is that the technical project of BN-800 reactor with MOX fuel was developed. When using the nitride fuel it is necessary to serve (in max extent) the mail technical decisions of this project. This report presents first results on development and justification of the BN-800 reactor with nitride fuel core. (authors)

  18. Proliferation resistance of advanced sustainable nuclear fuel cycles

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, H.E.; Lineberry, M.J.; Aumeier, S.E.; McFarlane, H.F. [Argonne National Lab.-West (United States)

    2001-07-01

    Intrinsic and extrinsic proliferation barriers of a pyro-process-based nuclear fuel cycle are discussed. While technical characteristics of the process raise new challenges for safeguards, others naturally facilitate the implementation of more integrated schemes for unattended continuous monitoring. In particular, the concept of operations accountability and model-assisted methods are revisited. While traditional safeguards constructs, such as material control and accountability, place greater emphasis on input/output characterization of nuclear processes, a model- based discrete event accountability approach could explicitly verify not only facility use but also internal operational dynamics. Under the proposed remote integral safeguards approach, transparency can be achieved efficiently, without divulging competitive or national security sensitive information. (author)

  19. Proliferation resistances of Generation IV recycling facilities for nuclear fuel

    OpenAIRE

    Åberg Lindell, Matilda

    2013-01-01

    The effects of global warming raise demands for reduced CO2 emissions, whereas at the same time the world’s need for energy increases. With the aim to resolve some of the difficulties facing today’s nuclear power, striving for safety, sustainability and waste minimization, a new generation of nuclear energy systems is being pursued: Generation IV. New reactor concepts and new nuclear facilities should be at least as resistant to diversion of nuclear material for weapons production, as were th...

  20. Stanniocalcin 2 promotes cell proliferation and cisplatin resistance in cervical cancer

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Yuxia; Gao, Ying; Cheng, Hairong; Yang, Guichun [Department of Gynecology, Harbin Medical University Cancer Hospital, Harbin, Heilongjiang, 150081 (China); Tan, Wenhua, E-mail: tanwenhua1962@163.com [Department of Gynecology, The Second Affiliated Hospital of Harbin Medical University, Harbin, Heilongjiang, 150086 (China)

    2015-10-23

    Cervical cancer is one of the most common carcinomas in the female reproductive system. Treatment of cervical cancer involves surgical removal and chemotherapy. Resistance to platinum-based chemotherapy drugs including cisplatin has increasingly become an important problem in the treatment of cervical cancer patients. We found in this study that stanniocalcin 2 (STC2) expression was upregulated in both cervical cancer tissues and cell lines. The levels of STC2 expression in cervical cancer cell lines were positively correlated with the rate of cell proliferation. Furthermore, in cisplatin resistant cervical cancer cells, the levels of STC2 expression were significantly elevated. Modulation of STC2 expression by siRNA or overexpression in cisplatin resistant cells resulted in altered cell survival, apoptosis, and cisplatin resistance. Finally, we found that there was significant difference in the activity of the MAPK signaling pathway between cisplatin sensitive and resistant cervical cancer cells, and that STC2 could regulate the activity of the MAPK signaling pathway. - Highlights: • STC2 was upregulated in cervical cancer and promoted cervical cancer cell proliferation. • Cisplatin resistant cells had elevated STC2 levels and enhanced proliferation. • STC2 regulated cisplatin chemosensitivity in cervical cancer cells. • STC2 regulated the activity of the MAPK signaling pathway.

  1. Nuclear proliferation and civilian nuclear power. Report of the Nonproliferation Alternative Systems Assessment Program. Volume II. Proliferation resistance

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-01

    The purpose of this volume is limited to an assessment of the relative effects that particular choices of nuclear-power systems, for whatever reasons, may have on the possible spread of nuclear-weapons capabilities. This volume addresses the concern that non-nuclear-weapons states may be able to initiate efforts to acquire or to improve nuclear-weapons capabilities through civilian nuclear-power programs; it also addresses the concern that subnational groups may obtain and abuse the nuclear materials or facilities of such programs, whether in nuclear-weapons states (NWS's) or nonnuclear-weapons states (NNW's). Accordingly, this volume emphasizes one important factor in such decisions, the resistance of nuclear-power systems to the proliferation of nuclear-weapons capabilities.

  2. BWR Assembly Optimization for Minor Actinide Recycling

    Energy Technology Data Exchange (ETDEWEB)

    G. Ivan Maldonado; John M. Christenson; J.P. Renier; T.F. Marcille; J. Casal

    2010-03-22

    The Primary objective of the proposed project is to apply and extend the latest advancements in LWR fuel management optimization to the design of advanced boiling water reactor (BWR) fuel assemblies specifically for the recycling of minor actinides (MAs).

  3. Proliferation resistance for fast reactors and related fuel cycles: issues and impacts

    Energy Technology Data Exchange (ETDEWEB)

    Pilat, Joseph F [Los Alamos National Laboratory

    2010-01-01

    The prospects for a dramatic growth in nuclear power may depend to a significant degree on the effectiveness of, and the resources devoted to, plans to develop and implement technologies and approaches that strengthen proliferation resistance and nuclear materials accountability. The challenges for fast reactors and related fuel cycles are especially critical. They are being explored in the Generation IV Tnternational Forum (GIF) and the Tnternational Atomic Energy Agency's (IAEA's) International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) initiative, as well as by many states that are looking to these systems for the efficient lise of uranium resources and long-term energy security. How do any proliferation risks they may pose compare to other reactors, both existing and under development, and their fuel cycles? Can they be designed with intrinsic (technological) features to make these systems more proliferation resistant? What roles can extrinsic (institutional) features play in proliferation resistance? What are the anticipated safeguards requirements, and will new technologies and approaches need to be developed? How can safeguards be facilitated by the design process? These and other questions require a rethinking of proliferation resistance and the prospects for new technologies and other intrinsic and extrinsic features being developed that are responsive to specific issues for fast reactors and related fuel cycles and to the broader threat environment in which these systems will have to operate. There are no technologies that can wholly eliminate the risk of proliferation by a determined state, but technology and design can playa role in reducing state threats and perhaps in eliminating non-state threats. There will be a significant role for extrinsic factors, especially the various measures - from safeguards and physical protection to export controls - embodied in the international nuclear nonproliferation regime. This paper

  4. An Expert Elicitation of the Proliferation Resistance of Using Small Modular Reactors (SMR) for the Expansion of Civilian Nuclear Systems.

    Science.gov (United States)

    Siegel, Jonas; Gilmore, Elisabeth A; Gallagher, Nancy; Fetter, Steve

    2018-02-01

    To facilitate the use of nuclear energy globally, small modular reactors (SMRs) may represent a viable alternative or complement to large reactor designs. One potential benefit is that SMRs could allow for more proliferation resistant designs, manufacturing arrangements, and fuel-cycle practices at widespread deployment. However, there is limited work evaluating the proliferation resistance of SMRs, and existing proliferation assessment approaches are not well suited for these novel arrangements. Here, we conduct an expert elicitation of the relative proliferation resistance of scenarios for future nuclear energy deployment driven by Generation III+ light-water reactors, fast reactors, or SMRs. Specifically, we construct the scenarios to investigate relevant technical and institutional features that are postulated to enhance the proliferation resistance of SMRs. The experts do not consistently judge the scenario with SMRs to have greater overall proliferation resistance than scenarios that rely on conventional nuclear energy generation options. Further, the experts disagreed on whether incorporating a long-lifetime sealed core into an SMR design would strengthen or weaken proliferation resistance. However, regardless of the type of reactor, the experts judged that proliferation resistance would be enhanced by improving international safeguards and operating several multinational fuel-cycle facilities rather than supporting many more national facilities. © 2017 Society for Risk Analysis.

  5. Nuclear PIM1 confers resistance to rapamycin-impaired endothelial proliferation

    Energy Technology Data Exchange (ETDEWEB)

    Walpen, Thomas; Kalus, Ina [Research Unit, Division Internal Medicine, University Hospital Zuerich, 8091 Zuerich (Switzerland); Schwaller, Juerg [Department of Biomedicine, University of Basel, 4031 Basel (Switzerland); Peier, Martin A. [Research Unit, Division Internal Medicine, University Hospital Zuerich, 8091 Zuerich (Switzerland); Battegay, Edouard J. [Research Unit, Division Internal Medicine, University Hospital Zuerich, 8091 Zuerich (Switzerland); Zurich Center for Integrative Human Physiology (ZIHP), 8057 Zuerich (Switzerland); Humar, Rok, E-mail: Rok.Humar@usz.ch [Research Unit, Division Internal Medicine, University Hospital Zuerich, 8091 Zuerich (Switzerland); Zurich Center for Integrative Human Physiology (ZIHP), 8057 Zuerich (Switzerland)

    2012-12-07

    Highlights: Black-Right-Pointing-Pointer Pim1{sup -/-} endothelial cell proliferation displays increased sensitivity to rapamycin. Black-Right-Pointing-Pointer mTOR inhibition by rapamycin enhances PIM1 cytosolic and nuclear protein levels. Black-Right-Pointing-Pointer Truncation of Pim1 beyond serine 276 results in nuclear localization of the kinase. Black-Right-Pointing-Pointer Nuclear PIM1 increases endothelial proliferation independent of rapamycin. -- Abstract: The PIM serine/threonine kinases and the mTOR/AKT pathway integrate growth factor signaling and promote cell proliferation and survival. They both share phosphorylation targets and have overlapping functions, which can partially substitute for each other. In cancer cells PIM kinases have been reported to produce resistance to mTOR inhibition by rapamycin. Tumor growth depends highly on blood vessel infiltration into the malignant tissue and therefore on endothelial cell proliferation. We therefore investigated how the PIM1 kinase modulates growth inhibitory effects of rapamycin in mouse aortic endothelial cells (MAEC). We found that proliferation of MAEC lacking Pim1 was significantly more sensitive to rapamycin inhibition, compared to wildtype cells. Inhibition of mTOR and AKT in normal MAEC resulted in significantly elevated PIM1 protein levels in the cytosol and in the nucleus. We observed that truncation of the C-terminal part of Pim1 beyond Ser 276 resulted in almost exclusive nuclear localization of the protein. Re-expression of this Pim1 deletion mutant significantly increased the proliferation of Pim1{sup -/-} cells when compared to expression of the wildtype Pim1 cDNA. Finally, overexpression of the nuclear localization mutant and the wildtype Pim1 resulted in complete resistance to growth inhibition by rapamycin. Thus, mTOR inhibition-induced nuclear accumulation of PIM1 or expression of a nuclear C-terminal PIM1 truncation mutant is sufficient to increase endothelial cell proliferation

  6. Framework for Proliferation Resistance and Physical Protection for Nonproliferation Impact Assessments.

    Energy Technology Data Exchange (ETDEWEB)

    Bari,R.

    2008-03-01

    This report describes a framework for proliferation resistance and physical protection evaluation for the fuel cycle systems envisioned in the expansion of nuclear power for electricity generation. The methodology is based on an approach developed as part of the Generation IV technical evaluation framework and on a qualitative evaluation approach to policy factors similar to those that were introduced in previous Nonproliferation Impact Assessments performed by DOE.

  7. Neutronic simulation calculations to assess the proliferation resistance of nuclear technologies; Neutronenphysikalische Simulationsrechnungen zur Proliferationsresistenz nuklearer Technologien

    Energy Technology Data Exchange (ETDEWEB)

    Englert, Matthias

    2009-07-13

    This thesis investigates the proliferation resistance of nuclear technologies on the basis of three case studies. After a brief description of the concept of proliferation resistance the utilized computer codes and methods are presented. The first case study investigates the potential of monolithic fuel for the conversion of one-fuel-element high-flux research reactors from highly enriched to low enriched uranium using the example of the german research reactor FRM-II. The second case study assesses the proliferation potential of future tokamak based fusion reactors by using neutronic simulations of a possible plutonium production. The third example investigates the proliferation potential of spallation neutron sources to produce nuclear weapon relevant material and the proliferation resistance of such facilities. (orig.)

  8. Rage induces hepatocellular carcinoma proliferation and sorafenib resistance by modulating autophagy.

    Science.gov (United States)

    Li, Jun; Wu, Peng-Wen; Zhou, Yuan; Dai, Bo; Zhang, Peng-Fei; Zhang, Yu-Hen; Liu, Yang; Shi, Xiao-Lei

    2018-02-14

    The receptor for advanced glycation end products (Rage) is involved in the development of various tumors and acts as an oncogenic protein. Rage is overexpressed in tumors including hepatocellular carcinoma (HCC). However, the molecular mechanism of Rage in HCC progression and sorafenib resistance remains unclear. In this study, enhanced Rage expression is highly associated proliferation and contributes to sorafenib resistance. Rage deficiency contributed to autophagy induction through activating AMPK/mTOR signaling pathway, which is important for sorafenib response. Moreover, the interactions between Rage and Rage ligands such as high mobility group box 1 (HMGB1) and s100a4 positively increased Rage expression. Our data indicate that Rage may be a potential target for therapeutic intervention in HCC and biomarker for sorafenib resistance.

  9. Very High-Temperature Reactor (VHTR) Proliferation Resistance and Physical Protection (PR&PP)

    Energy Technology Data Exchange (ETDEWEB)

    Moses, David Lewis [ORNL

    2011-10-01

    This report documents the detailed background information that has been compiled to support the preparation of a much shorter white paper on the design features and fuel cycles of Very High-Temperature Reactors (VHTRs), including the proposed Next-Generation Nuclear Plant (NGNP), to identify the important proliferation resistance and physical protection (PR&PP) aspects of the proposed concepts. The shorter white paper derived from the information in this report was prepared for the Department of Energy Office of Nuclear Science and Technology for the Generation IV International Forum (GIF) VHTR Systems Steering Committee (SSC) as input to the GIF Proliferation Resistance and Physical Protection Working Group (PR&PPWG) (http://www.gen-4.org/Technology/horizontal/proliferation.htm). The short white paper was edited by the GIF VHTR SCC to address their concerns and thus may differ from the information presented in this supporting report. The GIF PR&PPWG will use the derived white paper based on this report along with other white papers on the six alternative Generation IV design concepts (http://www.gen-4.org/Technology/systems/index.htm) to employ an evaluation methodology that can be applied and will evolve from the earliest stages of design. This methodology will guide system designers, program policy makers, and external stakeholders in evaluating the response of each system, to determine each system's resistance to proliferation threats and robustness against sabotage and terrorism threats, and thereby guide future international cooperation on ensuring safeguards in the deployment of the Generation IV systems. The format and content of this report is that specified in a template prepared by the GIF PR&PPWG. Other than the level of detail, the key exception to the specified template format is the addition of Appendix C to document the history and status of coated-particle fuel reprocessing technologies, which fuel reprocessing technologies have yet to be

  10. PPARγ1 phosphorylation enhances proliferation and drug resistance in human fibrosarcoma cells

    Energy Technology Data Exchange (ETDEWEB)

    Pang, Xiaojuan; Shu, Yuxin; Niu, Zhiyuan; Zheng, Wei; Wu, Haochen [State Key Laboratory of Pharmaceutical Biotechnology, Nanjing University, Nanjing (China); Lu, Yan, E-mail: luyan@nju.edu.cn [State Key Laboratory of Pharmaceutical Biotechnology, Nanjing University, Nanjing (China); Shen, Pingping, E-mail: ppshen@nju.edu.cn [State Key Laboratory of Pharmaceutical Biotechnology, Nanjing University, Nanjing (China); Model Animal Research Center (MARC), Nanjing University, Nanjing (China)

    2014-03-10

    Post-translational regulation plays a critical role in the control of cell growth and proliferation. The phosphorylation of peroxisome proliferator-activated receptor γ (PPARγ) is the most important post-translational modification. The function of PPARγ phosphorylation has been studied extensively in the past. However, the relationship between phosphorylated PPARγ1 and tumors remains unclear. Here we investigated the role of PPARγ1 phosphorylation in human fibrosarcoma HT1080 cell line. Using the nonphosphorylation (Ser84 to alanine, S84A) and phosphorylation (Ser84 to aspartic acid, S84D) mutant of PPARγ1, the results suggested that phosphorylation attenuated PPARγ1 transcriptional activity. Meanwhile, we demonstrated that phosphorylated PPARγ1 promoted HT1080 cell proliferation and this effect was dependent on the regulation of cell cycle arrest. The mRNA levels of cyclin-dependent kinase inhibitor (CKI) p21{sup Waf1/Cip1} and p27{sup Kip1} descended in PPARγ1{sup S84D} stable HT1080 cell, whereas the expression of p18{sup INK4C} was not changed. Moreover, compared to the PPARγ1{sup S84A}, PPARγ1{sup S84D} up-regulated the expression levels of cyclin D1 and cyclin A. Finally, PPARγ1 phosphorylation reduced sensitivity to agonist rosiglitazone and increased resistance to anticancer drug 5-fluorouracil (5-FU) in HT1080 cell. Our findings establish PPARγ1 phosphorylation as a critical event in human fibrosarcoma growth. These findings raise the possibility that chemical compounds that prevent the phosphorylation of PPARγ1 could act as anticancer drugs. - Highlights: • Phosphorylation attenuates PPARγ1 transcriptional activity. • Phosphorylated PPARγ1 promotes HT1080 cells proliferation. • PPARγ1 phosphorylation regulates cell cycle by mediating expression of cell cycle regulators. • PPARγ1 phosphorylation reduces sensitivity to agonist and anticancer drug. • Our findings establish PPARγ1 phosphorylation as a critical event in HT1080

  11. 44-BWR WASTE PACKAGE LOADING CURVE EVALUATION

    Energy Technology Data Exchange (ETDEWEB)

    J.M. Scaglione

    2004-08-25

    The objective of this calculation is to evaluate the required minimum burnup as a function of initial boiling water reactor (BWR) assembly enrichment that would permit loading of spent nuclear fuel into the 44 BWR waste package configuration as provided in Attachment IV. This calculation is an application of the methodology presented in ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2003). The scope of this calculation covers a range of enrichments from 0 through 5.0 weight percent (wt%) U-235, and a burnup range of 0 through 40 GWd/MTU. This activity supports the validation of the use of burnup credit for commercial spent nuclear fuel applications. The intended use of these results will be in establishing BWR waste package configuration loading specifications. Limitations of this evaluation are as follows: (1) The results are based on burnup credit for actinides and selected fission products as proposed in YMP (2003, Table 3-1) and referred to as the ''Principal Isotopes''. Any change to the isotope listing will have a direct impact on the results of this report. (2) The results of 100 percent of the current BWR projected waste stream being able to be disposed of in the 44-BWR waste package with Ni-Gd Alloy absorber plates is contingent upon the referenced waste stream being sufficiently similar to the waste stream received for disposal. (3) The results are based on 1.5 wt% Gd in the Ni-Gd Alloy material and having no tuff inside the waste package. If the Gd loading is reduced or a process to introduce tuff inside the waste package is defined, then this report would need to be reevaluated based on the alternative materials.

  12. Water chemistry control and decontamination experience with TEPCO BWR`s and the measures planned for the future

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, N.; Miyamaru, K. [Tokyo Electric Power Co. (Japan)

    1995-03-01

    The new TEPCO BWR`s are capable of having the occupational radiation exposure controlled successfully at a low level by selecting low cobalt steel, using corrosion-resistant steel, employing dual condensate polishing systems, and controlling Ni/Fe ratio during operation. The occupational radiation exposure of the old BWR`s, on the other hand, remains high though reduced substantially through the use of low cobalt replacement steel and the partial addition of a filter in the condensate polishing system. Currently under review is the overall decontamination procedure for the old BWR`s to find out to measures needed to reduce the amount of crud that is and has been carried over into the nuclear reactor. The current status of decontamination is reported below.

  13. Prevalence and proliferation of antibiotic resistance genes in two municipal wastewater treatment plants.

    Science.gov (United States)

    Mao, Daqing; Yu, Shuai; Rysz, Michal; Luo, Yi; Yang, Fengxia; Li, Fengxiang; Hou, Jie; Mu, Quanhua; Alvarez, P J J

    2015-11-15

    The propagation of antibiotic resistance genes (ARGs) is an emerging health concern worldwide. Thus, it is important to understand and mitigate their occurrence in different systems. In this study, 30 ARGs that confer resistance to tetracyclines, sulfonamides, quinolones or macrolides were detected in two activated sludge wastewater treatment plants (WWTPs) in northern China. Bacteria harboring ARGs persisted through all treatment units, and survived disinfection by chlorination in greater percentages than total Bacteria (assessed by 16S rRNA genes). Although the absolute abundances of ARGs were reduced from the raw influent to the effluent by 89.0%-99.8%, considerable ARG levels [(1.0 ± 0.2) × 10(3) to (9.5 ± 1.8) × 10(5) copies/mL)] were found in WWTP effluent samples. ARGs were concentrated in the waste sludge (through settling of bacteria and sludge dewatering) at (1.5 ± 2.3) × 10(9) to (2.2 ± 2.8) × 10(11) copies/g dry weight. Twelve ARGs (tetA, tetB, tetE, tetG, tetH, tetS, tetT, tetX, sul1, sul2, qnrB, ermC) were discharged through the dewatered sludge and plant effluent at higher rates than influent values, indicating overall proliferation of resistant bacteria. Significant antibiotic concentrations (2%-50% of raw influent concentrations) remained throughout all treatment units. This apparently contributed selective pressure for ARG replication since the relative abundance of resistant bacteria (assessed by ARG/16S rRNA gene ratios) was significantly correlated to the corresponding effluent antibiotic concentrations. Similarly, the concentrations of various heavy metals (which induce a similar bacterial resistance mechanism as antibiotics - efflux pumps) were also correlated to the enrichment of some ARGs. Thus, curtailing the release of antibiotics and heavy metals to sewage systems (or enhancing their removal in pre-treatment units) may alleviate their selective pressure and mitigate ARG proliferation in WWTPs. Copyright © 2015 Elsevier Ltd. All

  14. Anti-proliferation effect of blue light-emitting diodes against antibiotic-resistant Helicobacter pylori.

    Science.gov (United States)

    Ma, Jianwei; Hiratsuka, Takahiro; Etoh, Tsuyoshi; Akada, Junko; Fujishima, Hajime; Shiraishi, Norio; Yamaoka, Yoshio; Inomata, Masafumi

    2017-12-07

    Infection by Helicobacter pylori is implicated in a wide range of upper gastrointestinal diseases. Owing to the rapid emergence of antibiotic-resistant strains of H. pylori, the development of novel treatment modalities for antibiotic-resistant H. pylori infection is a key priority. Blue light-emitting diodes (LED) may represent a unique option owing to their antimicrobial effect. In this study, we aimed to evaluate the anti-proliferative effect of blue LED against antibiotic-resistant H. pylori. Ten antibiotic-resistant strains and one sensitive H. pylori strain were used in this study. After irradiation by blue LED along time course, the viability of H. pylori was evaluated by enumerating colony forming units. Morphological changes in H. pylori were observed using a scanning electron microscope. Reductase activity was measured as an indicator of bacterial cellular activity. Total reactive oxygen species was monitored using fluorescence intensity and fluorescence microscope imaging. After irradiation by blue LED, the numbers of H. pylori in all the strains were significantly reduced compared to control group. The H. pylori exhibited a short rod-shaped morphology after irradiation; no such change was observed in H. pylori not exposed to blue LED. Re-irradiation of surviving strain after the initial irradiation also exhibited the same anti-proliferation effect. After blue LED irradiation, bacterial cellular activity was lower and total reactive oxygen species production was significantly higher in blue LED group, compared to that in control. Blue LED could be a new treatment to eradicate infection with antibiotic-resistant H. pylori. This article is protected by copyright. All rights reserved.

  15. PPARγ1 phosphorylation enhances proliferation and drug resistance in human fibrosarcoma cells.

    Science.gov (United States)

    Pang, Xiaojuan; Shu, Yuxin; Niu, Zhiyuan; Zheng, Wei; Wu, Haochen; Lu, Yan; Shen, Pingping

    2014-03-10

    Post-translational regulation plays a critical role in the control of cell growth and proliferation. The phosphorylation of peroxisome proliferator-activated receptor γ (PPARγ) is the most important post-translational modification. The function of PPARγ phosphorylation has been studied extensively in the past. However, the relationship between phosphorylated PPARγ1 and tumors remains unclear. Here we investigated the role of PPARγ1 phosphorylation in human fibrosarcoma HT1080 cell line. Using the nonphosphorylation (Ser84 to alanine, S84A) and phosphorylation (Ser84 to aspartic acid, S84D) mutant of PPARγ1, the results suggested that phosphorylation attenuated PPARγ1 transcriptional activity. Meanwhile, we demonstrated that phosphorylated PPARγ1 promoted HT1080 cell proliferation and this effect was dependent on the regulation of cell cycle arrest. The mRNA levels of cyclin-dependent kinase inhibitor (CKI) p21(Waf1/Cip1) and p27(Kip1) descended in PPARγ1(S84D) stable HT1080 cell, whereas the expression of p18(INK4C) was not changed. Moreover, compared to the PPARγ1(S84A), PPARγ1(S84D) up-regulated the expression levels of cyclin D1 and cyclin A. Finally, PPARγ1 phosphorylation reduced sensitivity to agonist rosiglitazone and increased resistance to anticancer drug 5-fluorouracil (5-FU) in HT1080 cell. Our findings establish PPARγ1 phosphorylation as a critical event in human fibrosarcoma growth. These findings raise the possibility that chemical compounds that prevent the phosphorylation of PPARγ1 could act as anticancer drugs. Copyright © 2014 Elsevier Inc. All rights reserved.

  16. BWR MOX core monitoring at Kernkraftwerk Gundremmingen

    Energy Technology Data Exchange (ETDEWEB)

    Noel, Alejandro [Studsvik Scandpower (Suisse) GmbH, Nussbaumen AG (Switzerland); Holzer, Robert [NIS Ingenieurgesellschaft GmbH, Alzenau (Germany); Anton, Gerd [Studsvik Scandpower GmbH, Norderstedt (Germany); Smith, Kord [Studsvik Scandpower Inc., Idaho Falls (United States)

    2008-07-01

    The replacement of the core monitoring system for twin KWU Boiling Water Reactors (BWR) is presented. The reactors, Kernkraftwerk Gundremmingen B and C (KGG), are located in Germany. Core monitoring for KGG is more challenging than for most BWR reactors due to its core composition with about 30% MOX fuel assemblies. The objectives of this paper are to discuss the specific MOX modelling aspects in CASMO-4/Simulate-3, the impact of the MOX fuel on several core monitoring aspects like the LPRM detector modelling and to present some core monitoring results since the beginning of GARDEL's operation. The available core monitoring results confirm the accuracy of the underlying physical methods. The core monitoring system replacement att KGG was a common project of Studsvik Scandpower and NIS Ingenieurgesellschaft GmbH, where Studsvik Scandpower supplied its standard core monitoring system GARDEL and NIS was responsible for the computer hardware, system integration and plant specific add-ons. (authors)

  17. Water chemistry practice at German BWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Stellwag, B. [Framatome ANP GmbH, Erlangen (Germany); Staudt, U. [VGB PowerTech e.V., Essen (Germany)

    2005-02-01

    As visual examinations carried out in 1994 detected cracks in a German boiling water reactor (BWR) plant due to intergranular stress corrosion cracking in core shroud components manufactured from Nb-stabilized CrNi steel 1.4550, safety-related assessments and in-service inspections were subsequently performed for the other six German BWRs. No cracks were found in the core shrouds of these plants. The second major event in the early 1990s was the detection of cracks at various German BWRs in piping systems made of Ti-stabilized CrNi steel 1.4541 caused by thermal sensitization in the heat-affected zone of welds. Comprehensive investigations resulted in a number of remedial measures (repair, replacement) implemented at piping in contact with reactor coolant of temperatures above 200 C. Thanks to the remedial measures and according to the analyses performed, cracking in the components in question due to the considered damage mechanisms need not be expected. German operators have therefore continued operating their BWR plants on normal water chemistry with an oxidizing environment. As a precaution, more stringent reactor coolant quality requirements have been specified and the limiting values of VGB Guideline R 401 J revised. This paper gives an overview of the trends in chemistry parameters at German BWR plants in the past 10 years. In addition, other relevant experience gained from the German BWR plants operating under normal water chemistry conditions is outlined: dose rates and collective doses, fuel performance, and results of periodic in-service inspections of major components of the reactor system. In the nearly 10 years of plant operation since implementation of the remedial measures, no cracks or other indications have been detected in any of the systems and components concerned. (orig.)

  18. Role of bottom-fermenting brewer's yeast KEX2 in high temperature resistance and poor proliferation at low temperatures.

    Science.gov (United States)

    Yamagishi, Hiromi; Ohnuki, Shinsuke; Nogami, Satoru; Ogata, Tomoo; Ohya, Yoshikazu

    2010-08-01

    Variants of bottom-fermenting brewer's yeast that grew at high temperatures and showed poor proliferation and fermentation at low temperatures were isolated. Similar variants of laboratory yeast were also isolated and found to be incapable of mating. The KEX2 gene was cloned by complementation. It was shown to be responsible for these traits, because a KEX2 disruptant of Saccharomyces cerevisiae (S. cerevisiae) laboratory yeast grew poorly at low temperatures and was resistant to high temperatures. In addition, a Saccharomyces bayanus (S. bayanus)-type KEX2 (Sb-KEX2) disruptant of bottom-fermenting brewer's yeast grew poorly at low temperatures and was resistant to high temperatures. The KEX2 gene product plays an important role in proliferation of yeast at low temperatures, which is an important trait of bottom-fermenting brewer's yeast. These findings advance our understanding of the proliferation of yeast at low temperatures, especially that of bottom-fermenting brewer's yeast.

  19. Assessment of Proliferation Resistance of Closed Nuclear Fuel Cycle System with Sodium Cooled Fast Reactors Using INPRO Evaluation Methodology

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young In; Hahn, Do Hee; Won, Byung Chool; Lee, Dong Uk

    2007-11-15

    Using the INPRO methodology, the proliferation resistance of an innovative nuclear energy system(INS) defined as a closed nuclear fuel cycle system consisting of KALIMER and pyroprocessing, has been assessed. Considering a very early development stage of the INS concept, the PR assessment is carried out based on intrinsic features, if required information and data are not available. The PR assessment of KALIMER and JSFR using the INPRO methodology affirmed that an adequate proliferation resistance has been achieved in both INSs CNFC-SFR, considering the assessor's progress and maturity of design development. KALIMER and JSFR are developed or being developed conforming to the targets and criteria defined for developing Gen IV nuclear reactor system. Based on these assessment results, proliferation resistance and physical protection(PR and PP) of KALIMER and JSFR are evaluated from the viewpoint of requirements for future nuclear fuel cycle system. The envisioned INSs CNFC-SFR rely on active plutonium management based on a closed fuel cycle, in which a fissile material is recycled in an integrated fuel cycle facility within proper safeguards. There is no isolated plutonium in the closed fuel cycle. The material remains continuously in a sequence of highly radioactive matrices within inaccessible facilities. The proliferation resistance assessment should be an ongoing analysis that keeps up with the progress and maturity of the design of Gen IV SFR.

  20. VGB guideline for the water in nuclear power plants with light water reactors (BWR). VGB-R 401 J; VGB-Richtlinie fuer das Wasser in Kernkraftwerken mit Leichtwasserreaktoren (SWR). VGB-R 401 J

    Energy Technology Data Exchange (ETDEWEB)

    Rosskamp, M. [Vattenvall Europe Nuclear Energy, Kernkraftwerk Brunsbuettel GmbH und Co OHG (Germany); Albrecht, N. [Vattenvall Europe Nuclear Energy, Kernkraftwerk Kruemmel GmbH und Co OHG, Geesthacht (Germany); Ilg, U. [EnBW Kraftwerke AG, Kernkraftwerk Philippsburg (Germany); Neder, H. [E.ON Kernkraftwerk GmbH, Kernkraftwerk Isar, Essenbach (Germany); Reitzner, U. [AREVA NP GmbH, Erlangen (Germany); Riedmueller, B. [Kernkraftwerk Gundremmingen GmbH (Germany); Rutschow, D. [VGB-Geschaeftsstelle, Essen (Germany)

    2007-07-01

    The new VGB guideline for light water reactor (BWR) is presented. The guideline specifies the classical, oxidative water chemistry for BWR. Principal strategies of water chemistry are described. Recommendations are made for deviations of normal operation. The guideline specifies - based on latest findings - especially chloride and sulphate in reactor water in view to corrosion resistance of austenitic materials, normal operating values and action levels. (orig.)

  1. Utility of Social Modeling for Proliferation Assessment - Enhancing a Facility-Level Model for Proliferation Resistance Assessment of a Nuclear Enegry System

    Energy Technology Data Exchange (ETDEWEB)

    Coles, Garill A.; Brothers, Alan J.; Gastelum, Zoe N.; Olson, Jarrod; Thompson, Sandra E.

    2009-10-26

    The Utility of Social Modeling for Proliferation Assessment project (PL09-UtilSocial) investigates the use of social and cultural information to improve nuclear proliferation assessments, including nonproliferation assessments, Proliferation Resistance (PR) assessments, safeguards assessments, and other related studies. These assessments often use and create technical information about a host State and its posture towards proliferation, the vulnerability of a nuclear energy system (NES) to an undesired event, and the effectiveness of safeguards. This objective of this project is to find and integrate social and technical information by explicitly considering the role of cultural, social, and behavioral factors relevant to proliferation; and to describe and demonstrate if and how social science modeling has utility in proliferation assessment. This report describes a modeling approach and how it might be used to support a location-specific assessment of the PR assessment of a particular NES. The report demonstrates the use of social modeling to enhance an existing assessment process that relies on primarily technical factors. This effort builds on a literature review and preliminary assessment performed as the first stage of the project and compiled in PNNL-18438. [ T his report describes an effort to answer questions about whether it is possible to incorporate social modeling into a PR assessment in such a way that we can determine the effects of social factors on a primarily technical assessment. This report provides: 1. background information about relevant social factors literature; 2. background information about a particular PR assessment approach relevant to this particular demonstration; 3. a discussion of social modeling undertaken to find and characterize social factors that are relevant to the PR assessment of a nuclear facility in a specific location; 4. description of an enhancement concept that integrates social factors into an existing, technically

  2. Development status and potential program for development of proliferation-resistant molten-salt reactors

    Energy Technology Data Exchange (ETDEWEB)

    Engel, J.R.; Bauman, H.F.; Dearing, J.F.; Grimes, W.R.; McCoy, H.E. Jr.

    1979-03-01

    Preliminary studies of existing and conceptual molten-salt reactor (MSR) designs have led to the identification of conceptual systems that are technologically attractive when operated with denatured uranium as the principal fissile fuel. These denatured MSRs would also have favorable resource-utilization characteristics and substantial resistance to proliferation of weapons-usable nuclear materials. The report presents a summary of the current status of technology and a discussion of the major technical areas of a possible base program to develop commercial denatured MSRs. The general areas treated are (1) reactor design and development, (2) safety and safety related technology, (3) fuel-coolant behavior and fuel processing, and (4) reactor materials. A substantial development effort could lead to authorization for construction of a molten-salt test reactor about 5 years after the start of the program and operation of the unit about 10 years later. A prototype commercial denatured MSR could be expected to begin operating 25 years from the start of the program. The postulated base program would extend over 32 years and would cost about $700 million (1978 dollars, unescalated). Additional costs to construct the MSTR, $600 million, and the prototype commercial plant, $1470 million, would bring the total program cost to about $2.8 billion. Additional allowances probably should be made to cover contingencies and incidental technology areas not explicitly treated in this preliminary review.

  3. Residual stress analysis in BWR pressure vessel attachments

    Energy Technology Data Exchange (ETDEWEB)

    Dexter, R.J.; Leung, C.P. (Southwest Research Inst., San Antonio, TX (United States)); Pont, D. (FRAMASOFT+CSI, 69 - Lyon (France). Div. of Framatome)

    1992-06-01

    Residual stresses from welding processes can be the primary driving force for stress corrosion cracking (SCC) in BWR components. Thus, a better understanding of the causes and nature of these residual stresses can help assess and remedy SCC. Numerical welding simulation software, such as SYSWELD, and material property data have been used to quantify residual stresses for application to SCC assessments in BWR components. Furthermore, parametric studies using SYSWELD have revealed which variables significantly affect predicted residual stress. Overall, numerical modeling techniques can be used to evaluate residual stress for SCC assessments of BWR components and to identify and plan future SCC research.

  4. LAPUR5. BWR Core Stability Measurements

    Energy Technology Data Exchange (ETDEWEB)

    March-Leuba, J.; Otaduy, P.J. [Oak Ridge National Lab., TN (United States)

    1990-01-01

    LAPUR5 is a mathematical description of the core of a boiling water reactor (BWR). The program uses a point kinetics description of the neutron dynamics together with a distributed-parameter model of the core thermal hydrodynamics to produce a space-dependent representation of the dynamics of a BWR in the frequency domain for small perturbations about a steady-state condition. LAPUR5 consists of two autonomous modules, LAPUR5X, the steady-state module, and LAPUR5W the dynamics module, which are linked by use of an intermediate storage device. LAPUR5X solves the coolant and the fuel steady-state governing equations while LAPUR5W solves the dynamic equations for the coolant, fuel, and neutron field in the frequency domain. Considerable detail exists in the modeling of the thermohydrodynamics and the reactivity feedbacks. LAPUR5 can be run interactively (LAPUR) or in batch mode (BLAPUR); these are equivalent except that LAPUR allows the user to edit input parameters at run time. Both spawn the two autonomous modules consecutively and delete the intermediate storage files.

  5. LBB application in Swedish BWR design

    Energy Technology Data Exchange (ETDEWEB)

    Kornfeldt, H.; Bjoerk, K.O.; Ekstroem, P. [ABB Atom, Vaesteras (Sweden)

    1997-04-01

    The protection against dynamic effects in connection with potential pipe breaks has been implemented in different ways in the development of BWR reactor designs. First-generation plant designs reflect code requirements in effect at that time which means that no piping restraint systems were designed and built into those plants. Modern designs have, in contrast, implemented full protection against damage in connection with postulated pipe breaks, as required in current codes and regulations. Moderns standards and current regulatory demands can be met for the older plants by backfitting pipe whip restraint hardware. This could lead to several practical difficulties as these installations were not anticipated in the original plant design and layout. Meeting the new demands by analysis would in this situation have great advantages. Application of leak-before-break criteria gives an alternative opportunity of meeting modem standards in reactor safety design. Analysis takes into account data specific to BWR primary system operation, actual pipe material properties, piping loads and leak detection capability. Special attention must be given to ensure that the data used reflects actual plant conditions.

  6. A BWR licensing experience in the USA

    Energy Technology Data Exchange (ETDEWEB)

    Powers, J.; Ogura, C. [Toshiba America Nuclear Energy, Charlotte, North Carolina (United States); Arai, K. [Toshiba Corporation, Yokohama, Kanagawa (Japan); Thomas, S.; Mookhoek, B., E-mail: jim.powers@toshiba.com [Nuclear Innovation North America, Lake Jackson, Texas (United States)

    2015-09-15

    The US-Advanced Boiling Water Reactor (A BWR), certified by the United States Nuclear Regulatory Commission (US NRC), is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (STP3-4) Combined License Application (Cola). Nuclear Innovation North America (Nina) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. The STP3-4 project has finished the US NRC technical review of the Cola through the final meeting of the Advisory Committee on Reactor Safeguards (ACRS), and the Final Safety Evaluation Report (FSER) is scheduled to be issued by the US NRC in the middle of 2015. The next steps are to support the Mandatory Hearing process, and voting by the NRC commissioners on the motion to grant the Combined License, which is scheduled beginning of 2016 according to US NRC schedule as of March 30, 2015. This paper summarizes the history and progress of the US-A BWR licensing, including the experiences of the Licensee, Nina, and Toshiba as the Epc team worked through the Code of Federal Regulations Title 10 (10-Cfr) Part 52 process, and provides some perspectives on how the related licensing material would also be of value within a 10-Cfr Part 50, two-step process to minimize schedule and financial risks which could arise from ongoing technical developments and regulatory reviews. (Author)

  7. Endogenous growth factor stimulation of hemocyte proliferation induces resistance to Schistosoma mansoni challenge in the snail host.

    Science.gov (United States)

    Pila, Emmanuel A; Gordy, Michelle A; Phillips, Valerie K; Kabore, Alethe L; Rudko, Sydney P; Hanington, Patrick C

    2016-05-10

    Digenean trematodes are a large, complex group of parasitic flatworms that infect an incredible diversity of organisms, including humans. Larval development of most digeneans takes place within a snail (Gastropoda). Compatibility between snails and digeneans is often very specific, such that suitable snail hosts define the geographical ranges of diseases caused by these worms. The immune cells (hemocytes) of a snail are sentinels that act as a crucial barrier to infection by larval digeneans. Hemocytes coordinate a robust and specific immunological response, participating directly in parasite killing by encapsulating and clearing the infection. Hemocyte proliferation and differentiation are influenced by unknown digenean-specific exogenous factors. However, we know nothing about the endogenous control of hemocyte development in any gastropod model. Here, we identify and functionally characterize a progranulin [Biomphalaria glabrata granulin (BgGRN)] from the snail B. glabrata, a natural host for the human blood fluke Schistosoma mansoni Granulins are growth factors that drive proliferation of immune cells in organisms, spanning the animal kingdom. We demonstrate that BgGRN induces proliferation of B. glabrata hemocytes, and specifically drives the production of an adherent hemocyte subset that participates centrally in the anti-digenean defense response. Additionally, we demonstrate that susceptible B. glabrata snails can be made resistant to infection with S. mansoni by first inducing hemocyte proliferation with BgGRN. This marks the functional characterization of an endogenous growth factor of a gastropod mollusc, and provides direct evidence of gain of resistance in a snail-digenean infection model using a defined factor to induce snail resistance to infection.

  8. Calcium or resistant starch does not affect colonic epithelial cell proliferation throughout the colon in adenoma patients : A randomized controlled trial

    NARCIS (Netherlands)

    van Gorkom, Britta A P; Karrenbeld, Arend; van der Sluis, Tineke; Zwart, Nynke; van der Meer, Roelof; de Vries, Elisabeth G E; Kleibeuker, Jan H

    2002-01-01

    Patients with a history of sporadic adenomas have increased epithelial cell proliferative activity, an intermediate risk marker for colorectal cancer. Reduction of proliferation by dietary intervention may reflect a decreased colorectal cancer risk. To evaluate whether calcium or resistant starch

  9. Behaviour of the reactivity for BWR fuel cells; Comportamiento de la reactividad para celdas de combustible BWR

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez, J. A.; Alonso, G.; Delfin, A.; Vargas, S. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Del Valle G, E., E-mail: galonso@inin.gob.mx [IPN, Escuela Superior de Fisica y Matematicas, U. P. Adolfo Lopez Mateos, Col. Lindavista, 07738 Mexico D. F. (Mexico)

    2011-11-15

    In this work the behaviour of the reactivity of a fuel assembly type BWR was studied, the objective is to obtain some expressions that consider the average enrichment of U-235 and the gadolinium concentration like a function of the fuel cells burnt. Also, the applicability of the lineal reactivity model was analyzed for fuel cells type BWR. The analysis was carried out with the CASMO-4 code. (Author)

  10. Power oscillations in BWR reactors; Oscilaciones de potencia en reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa P, G. [Division de Ciencias Basicas e Ingenieria, Universidad Autonoma Metropolitana-Iztapalapa, 09340 Mexico D.F. (Mexico)]. E-mail: gepe@xanum.uam.mx

    2002-07-01

    One of the main problems in the operation of BWR type reactors is the instability in power that these could present. One type of oscillations and that is the objective of this work is the named density wave, which is attributed to the thermohydraulic processes that take place in the reactor core. From the beginnings of the development of BWR reactors, the stability of these has been an important aspect in their design, due to its possible consequences on the fuel integrity. The reactor core operates in two phase flow conditions and it is observed that under certain power and flow conditions, power instabilities appear. Studying this type of phenomena is complex, due to that a reactor core is constituted approximately by 27,000 fuel bars with different distributions of power and flow. The phenomena that cause the instability in BWR reactors continue being matter of scientific study. In the literature mainly in nuclear subject, it can be observed that exist different methods and approximations for studying this type of phenomena, nevertheless, their results are focused to establish safety limits in the reactor operation, instead of studying in depth of the knowledge about. Also in this line sense of the reactor data analysis, the oscillations characteristic frequencies are obtained for trying to establish if the power is growing or decreasing. In addition to that before mentioned in this paper it is presented a rigorous study applying the volumetric average method, for obtaining the vacuum waves propagation velocities and its possible connection with the power oscillations. (Author)

  11. Peroxisome Proliferator-Activated Receptors in Regulation of Cytochromes P450: New Way to Overcome Multidrug Resistance?

    Directory of Open Access Journals (Sweden)

    Katerina Cizkova

    2012-01-01

    Full Text Available Embryonic and tumour cells are able to protect themselves against various harmful compounds. In human pathology, this phenomenon exists in the form of multidrug resistance (MDR that significantly deteriorates success of anticancer treatment. Cytochromes P450 (CYPs play one of the key roles in the xenobiotic metabolism. CYP expression could contribute to resistance of cancer cells to chemotherapy. CYP epoxygenases (CYP2C and CYP2J metabolize about 20% of clinically important drugs. Besides of drug metabolism, CYP epoxygenases and their metabolites play important role in embryos, normal body function, and tumors. They participate in angiogenesis, mitogenesis, and cell signaling. It was found that CYP epoxygenases are affected by peroxisome proliferator-activated receptor α (PPARα. Based on the results of current studies, we assume that PPARs ligands may regulate CYP2C and CYP2J and in some extent they may contribute to overcoming of MDR in patients with different types of tumours.

  12. Piperlongumine inhibits the proliferation and survival of B-cell acute lymphoblastic leukemia cell lines irrespective of glucocorticoid resistance

    Energy Technology Data Exchange (ETDEWEB)

    Han, Seong-Su, E-mail: seong-su-han@uiowa.edu [Division of Pediatric Hematology-Oncology, University of Iowa Carver College of Medicine, Iowa City, IA (United States); Han, Sangwoo [Health and Human Physiology, University of Iowa Carver College of Medicine, Iowa City, IA (United States); Kamberos, Natalie L. [Division of Pediatric Hematology-Oncology, University of Iowa Carver College of Medicine, Iowa City, IA (United States)

    2014-09-26

    Highlights: • PL inhibits the proliferation of B-ALL cell lines irrespective of GC-resistance. • PL selectively kills B-ALL cells by increasing ROS, but not normal counterpart. • PL does not sensitize majority of B-ALL cells to DEX. • PL represses the network of constitutively activated TFs and modulates their target genes. • PL may serve as a new therapeutic molecule for GC-resistant B-ALL. - Abstract: Piperlongumine (PL), a pepper plant alkaloid from Piper longum, has anti-inflammatory and anti-cancer properties. PL selectively kills both solid and hematologic cancer cells, but not normal counterparts. Here we evaluated the effect of PL on the proliferation and survival of B-cell acute lymphoblastic leukemia (B-ALL), including glucocorticoid (GC)-resistant B-ALL. Regardless of GC-resistance, PL inhibited the proliferation of all B-ALL cell lines, but not normal B cells, in a dose- and time-dependent manner and induced apoptosis via elevation of ROS. Interestingly, PL did not sensitize most of B-ALL cell lines to dexamethasone (DEX). Only UoC-B1 exhibited a weak synergistic effect between PL and DEX. All B-ALL cell lines tested exhibited constitutive activation of multiple transcription factors (TFs), including AP-1, MYC, NF-κB, SP1, STAT1, STAT3, STAT6 and YY1. Treatment of the B-ALL cells with PL significantly downregulated these TFs and modulated their target genes. While activation of AURKB, BIRC5, E2F1, and MYB mRNA levels were significantly downregulated by PL, but SOX4 and XBP levels were increased by PL. Intriguingly, PL also increased the expression of p21 in B-ALL cells through a p53-independent mechanism. Given that these TFs and their target genes play critical roles in a variety of hematological malignancies, our findings provide a strong preclinical rationale for considering PL as a new therapeutic agent for the treatment of B-cell malignancies, including B-ALL and GC-resistant B-ALL.

  13. BWR ASSEMBLY SOURCE TERMS FOR WASTE PACKAGE DESIGN

    Energy Technology Data Exchange (ETDEWEB)

    T.L. Lotz

    1997-02-15

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide boiling water reactor (BWR) assembly radiation source term data for use during Waste Package (WP) design. The BWR assembly radiation source terms are to be used for evaluation of radiolysis effects at the WP surface, and for personnel shielding requirements during assembly or WP handling operations. The objectives of this evaluation are to generate BWR assembly radiation source terms that bound selected groupings of BWR assemblies, with regard to assembly average burnup and cooling time, which comprise the anticipated MGDS BWR commercial spent nuclear fuel (SNF) waste stream. The source term data is to be provided in a form which can easily be utilized in subsequent shielding/radiation dose calculations. Since these calculations may also be used for Total System Performance Assessment (TSPA), with appropriate justification provided by TSPA, or radionuclide release rate analysis, the grams of each element and additional cooling times out to 25 years will also be calculated and the data included in the output files.

  14. Status update of the BWR cask simulator

    Energy Technology Data Exchange (ETDEWEB)

    Lindgren, Eric R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Durbin, Samuel G. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-09-01

    The performance of commercial nuclear spent fuel dry storage casks are typically evaluated through detailed numerical analysis of the system's thermal performance. These modeling efforts are performed by the vendor to demonstrate the performance and regulatory compliance and are independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Numerous studies have been previously conducted. Recent advances in dry storage cask designs have moved the storage location from above ground to below ground and significantly increased the maximum thermal load allowed in a cask in part by increasing the canister helium pressure. Previous cask performance validation testing did not capture these parameters. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern dry casks. These modern cask designs utilize elevated helium pressure in the sealed canister or are intended for subsurface storage. The BWR cask simulator (BCS) has been designed in detail for both the above ground and below ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the canister. The symmetric single assembly geometry with well-controlled boundary conditions simplifies interpretation of results. Various configurations of outer concentric ducting will be used to mimic conditions for above and below ground storage configurations

  15. BWR Source Term Generation and Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    J.C. Ryman

    2003-07-31

    This calculation is a revision of a previous calculation (Ref. 7.5) that bears the same title and has the document identifier BBAC00000-01717-0210-00006 REV 01. The purpose of this revision is to remove TBV (to-be-verified) -41 10 associated with the output files of the previous version (Ref. 7.30). The purpose of this and the previous calculation is to generate source terms for a representative boiling water reactor (BWR) spent nuclear fuel (SNF) assembly for the first one million years after the SNF is discharged from the reactors. This calculation includes an examination of several ways to represent BWR assemblies and operating conditions in SAS2H in order to quantify the effects these representations may have on source terms. These source terms provide information characterizing the neutron and gamma spectra in particles per second, the decay heat in watts, and radionuclide inventories in curies. Source terms are generated for a range of burnups and enrichments (see Table 2) that are representative of the waste stream and stainless steel (SS) clad assemblies. During this revision, it was determined that the burnups used for the computer runs of the previous revision were actually about 1.7% less than the stated, or nominal, burnups. See Section 6.6 for a discussion of how to account for this effect before using any source terms from this calculation. The source term due to the activation of corrosion products deposited on the surfaces of the assembly from the coolant is also calculated. The results of this calculation support many areas of the Monitored Geologic Repository (MGR), which include thermal evaluation, radiation dose determination, radiological safety analyses, surface and subsurface facility designs, and total system performance assessment. This includes MGR items classified as Quality Level 1, for example, the Uncanistered Spent Nuclear Fuel Disposal Container (Ref. 7.27, page 7). Therefore, this calculation is subject to the requirements of the

  16. FOXD1 promotes breast cancer proliferation and chemotherapeutic drug resistance by targeting p27

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Yi-Fan; Zhao, Jing-Yu; Yue, Hong [Department of Anesthesiology, The First Affiliated Hospital of the General Hospital of CPLA, Beijing 100048 (China); Hu, Ke-Shi; Shen, Hao [Department of Anesthesiology, The General Hospital of CPLA, Beijing 100853 (China); Guo, Zheng-Gang, E-mail: gsgzg304@163.com [Department of Anesthesiology, The First Affiliated Hospital of the General Hospital of CPLA, Beijing 100048 (China); Su, Xiao-Jun, E-mail: lucusebibi@163.com [Department of Anesthesiology, The First Affiliated Hospital of the General Hospital of CPLA, Beijing 100048 (China)

    2015-01-02

    Highlights: • FOXD1 is up-regulated in breast cancer tissues. • FOXD1 promotes breast cancer cell proliferation and chemoresistance by inducing G1 to S transition. • FOXD1 transcriptionally suppresses p27 expression. - Abstract: Forkhead transcription factors are essential for diverse processes in early embryonic development and organogenesis. As a member of the forkhead family, FOXD1 is required during kidney development and its inactivation results in failure of nephron progenitor cells. However, the role of FOXD1 in carcinogenesis and progression is still limited. Here, we reported that FOXD1 is a potential oncogene in breast cancer. We found that FOXD1 is up-regulated in breast cancer tissues. Depletion of FOXD1 expression decreases the ability of cell proliferation and chemoresistance in MDA-MB-231 cells, whereas overexpression of FOXD1 increases the ability of cell proliferation and chemoresistance in MCF-7 cells. Furthermore, we observed that FOXD1 induces G1 to S phase transition by targeting p27 expression. Our results suggest that FOXD1 may be a potential therapy target for patients with breast cancer.

  17. Improved Insulin Resistance and Lipid Metabolism by Cinnamon Extract through Activation of Peroxisome Proliferator-Activated Receptors

    Directory of Open Access Journals (Sweden)

    Xiaoyan Sheng

    2008-01-01

    Full Text Available Peroxisome proliferator-activated receptors (PPARs are transcriptional factors involved in the regulation of insulin resistance and adipogenesis. Cinnamon, a widely used spice in food preparation and traditional antidiabetic remedy, is found to activate PPARγ and α, resulting in improved insulin resistance, reduced fasted glucose, FFA, LDL-c, and AST levels in high-caloric diet-induced obesity (DIO and db/db mice in its water extract form. In vitro studies demonstrate that cinnamon increases the expression of peroxisome proliferator-activated receptors γ and α (PPARγ/α and their target genes such as LPL, CD36, GLUT4, and ACO in 3T3-L1 adipocyte. The transactivities of both full length and ligand-binding domain (LBD of PPARγ and PPARα are activated by cinnamon as evidenced by reporter gene assays. These data suggest that cinnamon in its water extract form can act as a dual activator of PPARγ and α, and may be an alternative to PPARγ activator in managing obesity-related diabetes and hyperlipidemia.

  18. An overview of the BWR ECCS strainer blockage issues

    Energy Technology Data Exchange (ETDEWEB)

    Serkiz, A.W.; Marshall, M.L. Jr.; Elliott, R. [Nuclear Regulatory Commission, Washington, DC (United States)

    1996-03-01

    This Paper provides a brief overview of actions taken in the mid 1980s to resolve Unresolved Safety Issue (USI) A-43, {open_quotes}Containment Emergency Sump Performance,{close_quotes} and their relationship to the BWR strainer blockage issue; the importance of insights gained from the Barseback-2 (a Swedish BWR) incident in 1992 and from ECCS strainer testing and inspections at the Perry nuclear power plant in 1992 and 1993; an analysis of an US BWR/4 with a Mark I containment; an international community sharing of knowledge relevant to ECCS strainer blockage, additional experimental programs; and identification of actions needed to resolve the strainer blockage issue and the status of such efforts.

  19. OPTIMIZATION OF HETEROGENEOUS UTILIZATION OF THORIUM IN PWRS TO ENHANCE PROLIFERATION RESISTANCE AND REDUCE WASTE.

    Energy Technology Data Exchange (ETDEWEB)

    TODOSOW,M.; KAZIMI,M.

    2004-08-01

    Issues affecting the implementation, public perception and acceptance of nuclear power include: proliferation, radioactive waste, safety, and economics. The thorium cycle directly addresses the proliferation and waste issues, but optimization studies of core design and fuel management are needed to ensure that it fits within acceptable safety and economic margins. Typical pressurized water reactors, although loaded with uranium fuel, produce 225 to 275 kg of plutonium per gigawatt-year of operation. Although the spent fuel is highly radioactive, it nevertheless offers a potential proliferation pathway because the plutonium is relatively easy to separate, amounts to many critical masses, and does not present any significant intrinsic barrier to weapon assembly. Uranium 233, on the other hand, produced by the irradiation of thorium, although it too can be used in weapons, may be ''denatured'' by the addition of natural, depleted or low enriched uranium. Furthermore, it appears that the chemical behavior of thoria or thoria-urania fuel makes it a more stable medium for the geological disposal of the spent fuel. It is therefore particularly well suited for a once-through fuel cycle. The use of thorium as a fertile material in nuclear fuel has been of interest since the dawn of nuclear power technology due to its abundance and to potential neutronic advantages. Early projects include homogeneous mixtures of thorium and uranium oxides in the BORAX-IV, Indian Point I, and Elk River reactors, as well as heterogeneous mixtures in the Shippingport seed-blanket reactor. However these projects were developed under considerably different circumstances than those which prevail at present. The earlier applications preceded the current proscription, for non-proliferation purposes, of the use of uranium enriched to more than 20 w/o in {sup 235}U, and has in practice generally prohibited the use of uranium highly enriched in {sup 235}U. They were designed when

  20. Pirarubicin inhibits multidrug-resistant osteosarcoma cell proliferation through induction of G2/M phase cell cycle arrest

    Science.gov (United States)

    Zheng, Shui-er; Xiong, Sang; Lin, Feng; Qiao, Guang-lei; Feng, Tao; Shen, Zan; Min, Da-liu; Zhang, Chun-ling; Yao, Yang

    2012-01-01

    Aim: Pirarubicin (THP) is recently found to be effective in treating patients with advanced, relapsed or recurrent high-grade osteosarcoma. In this study, the effects of THP on the multidrug-resistant (MDR) osteosarcoma cells were assessed, and the underlying mechanisms for the disruption of cell cycle kinetics by THP were explored. Methods: Human osteosarcoma cell line MG63 and human MDR osteosarcoma cell line MG63/DOX were tested. The cytotoxicity of drugs was examined using a cell proliferation assay with the Cell Counting Kit-8 (CCK-8). The distribution of cells across the cell cycle was determined with flow cytometry. The expression of cell cycle-regulated genes cyclin B1 and Cdc2 (CDK1), and the phosphorylated Cdc2 and Cdc25C was examined using Western blot analyses. Results: MG63/DOX cells were highly resistant to doxorubicin (ADM) and gemcitabine (GEM), but were sensitive or lowly resistant to THP, methotrexate (MTX) and cisplatin (DDP). Treatment of MG63/DOX cells with THP (200–1000 ng/mL) inhibited the cell proliferation in time- and concentration-dependent manners. THP (50–500 ng/mL) induced MG63/DOX cell cycle arrest at the G2/M phase in time- and concentration-dependent manners. Furthermore, the treatment of MG63/DOX cells with THP (200–1000 ng/mL) downregulated cyclin B1 expression, and decreased the phosphorylated Cdc2 at Thr161. Conversely, the treatment increased the phosphorylated Cdc2 at Thr14/Tyr15 and Cdc25C at Ser216, which led to a decrease in Cdc2-cyclin B1 activity. Conclusion: The cytotoxicity of THP to MG63/DOX cells may be in part due to its ability to arrest cell cycle progression at the G2/M phase, which supports the use of THP for managing patients with MDR osteosarcoma. PMID:22580740

  1. Prostate cancer stem-like cells proliferate slowly and resist etoposide-induced cytotoxicity via enhancing DNA damage response

    Energy Technology Data Exchange (ETDEWEB)

    Yan, Judy [Division of Nephrology, Department of Medicine, McMaster University, Juravinski Innovation Tower, Room T3310, St. Joseph' s Hospital, 50 Charlton Ave East, Hamilton, Ontario, Canada L8S 4L8 (Canada); Father Sean O' Sullivan Research Institute, Hamilton, Ontario, Canada L8N 4A6 (Canada); The Hamilton Centre for Kidney Research (HCKR), St. Joseph' s Hamilton Healthcare, Hamilton, Ontario, Canada L8N 4A6 (Canada); Tang, Damu, E-mail: damut@mcmaster.ca [Division of Nephrology, Department of Medicine, McMaster University, Juravinski Innovation Tower, Room T3310, St. Joseph' s Hospital, 50 Charlton Ave East, Hamilton, Ontario, Canada L8S 4L8 (Canada); Father Sean O' Sullivan Research Institute, Hamilton, Ontario, Canada L8N 4A6 (Canada); The Hamilton Centre for Kidney Research (HCKR), St. Joseph' s Hamilton Healthcare, Hamilton, Ontario, Canada L8N 4A6 (Canada)

    2014-10-15

    Despite the development of chemoresistance as a major concern in prostate cancer therapy, the underlying mechanisms remain elusive. In this report, we demonstrate that DU145-derived prostate cancer stem cells (PCSCs) progress slowly with more cells accumulating in the G1 phase in comparison to DU145 non-PCSCs. Consistent with the important role of the AKT pathway in promoting G1 progression, DU145 PCSCs were less sensitive to growth factor-induced activation of AKT in comparison to non-PCSCs. In response to etoposide (one of the most commonly used chemotherapeutic drugs), DU145 PCSCs survived significantly better than non-PCSCs. In addition to etoposide, PCSCs demonstrated increased resistance to docetaxel, a taxane drug that is commonly used to treat castration-resistant prostate cancer. Etoposide produced elevated levels of γH2AX and triggered a robust G2/M arrest along with a coordinated reduction of the G1 population in PCSCs compared to non-PCSCs, suggesting that elevated γH2AX plays a role in the resistance of PCSCs to etoposide-induced cytotoxicity. We have generated xenograft tumors from DU145 PCSCs and non-PCSCs. Consistent with the knowledge that PCSCs produce xenograft tumors with more advanced features, we were able to demonstrate that PCSC-derived xenograft tumors displayed higher levels of γH2AX and p-CHK1 compared to non-PCSC-produced xenograft tumors. Collectively, our research suggests that the elevation of DNA damage response contributes to PCSC-associated resistance to genotoxic reagents. - Highlights: • Increased survival in DU145 PCSCs following etoposide-induced cytotoxicity. • PCSCs exhibit increased sensitivity to etoposide-induced DDR. • Resistance to cytotoxicity may be due to slower proliferation in PCSCs. • Reduced kinetics to growth factor induced activation of AKT in PCSCs.

  2. Aldo-keto reductase 1B10 and its role in proliferation capacity of drug-resistant cancers

    Directory of Open Access Journals (Sweden)

    Toshiyuki eMatsunaga

    2012-01-01

    Full Text Available The human aldo-keto reductase AKR1B10, originally identified as an aldose reductase-like protein and human small intestine aldose reductase, is a cytosolic NADPH-dependent reductase that metabolizes a variety of endogenous compounds, such as aromatic and aliphatic aldehydes and dicarbonyl compounds, and some drug ketones. The enzyme is highly expressed in solid tumors of several tissues including lung and liver, and as such has received considerable interest as a relevant biomarker for the development of those tumors. In addition, AKR1B10 has been recently reported to be significantly up-regulated in some cancer cell lines (medulloblastoma D341 and colon cancer HT29 acquiring resistance towards chemotherapeutic agents (cyclophosphamide and mitomycin c, suggesting the validity of the enzyme as a chemoresistance marker. Although the detailed information on the AKR1B10-mediated mechanisms leading to the drug resistance process is not well understood so far, the enzyme has been proposed to be involved in functional regulations of cell proliferation and metabolism of drugs and endogenous lipids during the development of chemoresistance. This article reviews the current literature focusing mainly on expression profile and roles of AKR1B10 in the drug resistance of cancer cells. Recent developments of AKR1B10 inhibitors and their usefulness in restoring sensitivity to anticancer drugs are also reviewed.

  3. Dysregulated connexin 43 in HER2-positive drug resistant breast cancer cells enhances proliferation and migration.

    Science.gov (United States)

    Yeh, Elizabeth S; Williams, Christina J; Williams, Carly Bess; Bonilla, Ingrid V; Klauber-DeMore, Nancy; Phillips, Stephanie L

    2017-12-12

    Connexin 43 (Cx43) is a gap junction protein whose function in the development of breast cancer and in breast cancer progression remains unclear. Evidence suggests that Cx43 ( GJA1 ) mRNA and protein expression is altered in breast tumors. However, reports indicate both increased and decreased Cx43 levels in human breast cancer samples. Studies also suggest that loss of Cx43 regulated gap junction intercellular communication is a common feature of breast malignancies that potentially correlates with histological stage. Further evidence suggests that Cx43 ( GJA1 ) mRNA expression is negatively correlated with HER2 positivity but a relationship between Cx43 and HER2 in breast cancer is not well defined. Therefore, in this study, we sought to evaluate the relationship between Cx43 activity, HER2, and drug resistance. Using HER2+ breast cancer cell lines that are sensitive or resistant to HER2 inhibitor, we evaluated Cx43 gap junction function. We found that Cx43 gap junction activity is completely lost in drug resistant HER2-positive (HER2+) breast cancer cells, whereas Cx43 gap junction activity can be restored by Cx43 overexpression in drug sensitive HER2+ cells. Moreover, the dysregulation of Cx43 resulted in increased tumorigenic and migratory capacity of the HER2+ drug resistant breast cancer cells.

  4. Pathway Aggregation in the Risk Assessment of Proliferation Resistance and Physical Protection (PR&PP) of Nuclear Energy Systems

    Energy Technology Data Exchange (ETDEWEB)

    Aldemir, Tunc [Ohio State Univ., Columbus, OH (United States); Denning, Richard [Ohio State Univ., Columbus, OH (United States); Catalyurek, Umit [Ohio State Univ., Columbus, OH (United States); Yilmaz, Alper [Ohio State Univ., Columbus, OH (United States); Yue, Meng [Brookhaven National Lab. (BNL), Upton, NY (United States); Cheng, Lap-Yan [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2015-01-23

    The framework for Proliferation Resistance and Physical Protection (PR & PP) evaluation is to define a set of challenges, to obtain the system responses, and to assess the outcomes. The assessment of outcomes heavily relies on pathways, defined as sequences of events or actions that could potentially be followed by a State or a group of individuals in order to achieve a proliferation objective, with the defined threats as initiating events. There may be large number of segments connecting pathway stages (e.g. acquisition, processing, and fabrication for PR) which can lead to even larger number of pathways or scenarios through possible different combinations of segment connections, each with associated probabilities contributing to the overall risk. Clustering of these scenarios in specified stage attribute intervals is important for their tractable analysis and outcome assessment. A software tool for scenario generation and clustering (OSUPR) is developed that utilizes the PRCALC code developed at the Brookhaven National Laboratory for scenario generation and the K- means, mean shift and adaptive mean shift algorithms as possible clustering schemes. The results of the study using the Example Sodium Fast Breeder as an example system show that clustering facilitates the probabilistic or deterministic analysis of scenarios to identify system vulnerabilities and communication of the major risk contributors to stakeholders. The results of the study also show that the mean shift algorithm has the most potential for assisting the analysis of the scenarios generated by PRCALC.

  5. Enhanced satellite cell proliferation with resistance training in elderly men and women

    DEFF Research Database (Denmark)

    Mackey, Abigail; Esmarck, B; Kadi, F

    2007-01-01

    In addition to the well-documented loss of muscle mass and strength associated with aging, there is evidence for the attenuating effects of aging on the number of satellite cells in human skeletal muscle. The aim of this study was to investigate the response of satellite cells in elderly men...... and women to 12 weeks of resistance training. Biopsies were collected from the m. vastus lateralis of 13 healthy elderly men and 16 healthy elderly women (mean age 76+/-SD 3 years) before and after the training period. Satellite cells were visualized by immunohistochemical staining of muscle cross.......15+/-0.06; mean+/-SD) and females (from 0.11+/-0.04 to 0.13+/-0.05). These results suggest that 12 weeks of resistance training is effective in enhancing the satellite cell pool in skeletal muscle in the elderly....

  6. Effect of copper nanoparticles administered in ovo on the activity of proliferating cells and on the resistance of femoral bones in broiler chickens

    DEFF Research Database (Denmark)

    Mroczek-Sosnowska, Natalia; Lukasiewicz, Monika; Adamek, Dobrochna

    2017-01-01

    The objective of this study was to evaluate bone resistance after in ovo administration of copper nanoparticles (NanoCu) and to determine the number of cells positive for proliferating cell nuclear antigen (PCNA) in the femoral bones of broiler chickens (n = 12 per group). The study demonstrated...... that femoral bones from the NanoCu group were characterised by a higher weight and volume and by significantly greater resistance to fractures compared to the Control group. NanoCu promoted the proliferation of PCNA-positive cells in the long bones of chickens. A significantly higher number of PCNA...

  7. BWR stability using a reducing dynamical model; Estabilidad de un BWR con un modelo dinamico reducido

    Energy Technology Data Exchange (ETDEWEB)

    Ballestrin Bolea, J. M.; Blazquez Martinez, J. B.

    1990-07-01

    BWR stability can be treated with reduced order dynamical models. When the parameters of the model came from dynamical models. When the parameters of the model came from experimental data, the predictions are accurate. In this work an alternative derivation for the void fraction equation is made, but remarking the physical structure of the parameters. As the poles of power/reactivity transfer function are related with the parameters, the measurement of the poles by other techniques such as noise analysis will lead to the parameters, but the system of equations is non-linear. Simple parametric calculation of decay ratio are performed, showing why BWRs become unstable when they are operated at low flow and high power. (Author)

  8. Advanced methods for BWR transient and stability analysis

    Energy Technology Data Exchange (ETDEWEB)

    Schmidt, A.; Wehle, F.; Opel, S.; Velten, R. [AREVA, AREVA NP, Erlangen (Germany)

    2008-07-01

    The design of advanced Boiling Water Reactor (BWR) fuel assemblies and cores is governed by the basic requirement of safe, reliable and flexible reactor operation with optimal fuel utilization. AREVA NP's comprehensive steady state and transient BWR methodology allows the designer to respond quickly and effectively to customer needs. AREVA NP uses S-RELAP5/RAMONA as the appropriate methodology for the representation of the entire plant. The 3D neutron kinetics and thermal-hydraulics code has been developed for the prediction of system, fuel and core behavior and provides additional margins for normal operation and transients. Of major importance is the extensive validation of the methodology. The validation is based on measurements at AREVA NP's test facilities, and comparison of the predictions with a great wealth of measured data gathered from BWR plants during many years of operation. Three of the main fields of interest are stability analysis, operational transients and reactivity initiated accidents (RIAs). The introduced 3D methodology for operational transients shows significant margin regarding the operational limit of critical power ratio, which has been approved by the German licensing authority. Regarding BWR stability a large number of measurements at different plants under various conditions have been performed and successfully post-calculated with RAMONA. This is the basis of reliable pre-calculations of the locations of regional and core-wide stability boundaries. (authors)

  9. RAMONA analysis of BWR stability at nuclear power plant Brunsbuettel

    Energy Technology Data Exchange (ETDEWEB)

    Kappes, C.; Velten, R.; Wehle, F. [AREVA NP GmbH, Erlangen (Germany); Huettmann, A.; Schuster, R. [Vattenfall Europe GmbH, Hamburg (Germany)

    2010-05-15

    For high power/low flow operating conditions associated with unfavorable core power distributions, BWR operation requires attention with respect to potential power and flow oscillations. Beside stability analyses based on highly validated methodology as RAMONA, also stability measurements are performed in BWR plants. Such measurements usually cover the evaluation of Average Power Range Monitor (APRM) and Local Power Range Monitor (LPRM) signals of the BWR core at several operating conditions. This paper presents the numerical simulation of stability phenomena which were recorded in the frame of a stability measurement at the nuclear power plant Brunsbuettel (KKB) on December 12{sup th} 2004 (Cycle 18). The measurement showed a local instability at most investigated operating points and a temporal global instability when the reactor was operated at conditions where four of the eight recirculation pumps were running. The numerical investigation with RAMONA-3 focuses on the operating point with four recirculation pumps when a temporal global instability has been measured. It will be shown that a local destabilization of a single Fuel Assembly (FA) can yield a global instability mode when the reactor is operating under high power and low flow conditions. Such phenomena have already been observed and analyzed for other BWR plants as e.g. Forsmark-1. (orig.)

  10. Fate and proliferation of typical antibiotic resistance genes in five full-scale pharmaceutical wastewater treatment plants

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jilu [College of Environmental Science and Engineering, Ministry of Education Key Laboratory of Pollution Processes and Environmental Criteria, Nankai University, Tianjin 300071 (China); Mao, Daqing, E-mail: mao@tju.edu.cn [School of Environmental Science and Engineering, Tianjin University, Tianjin 300072 (China); Mu, Quanhua [College of Environmental Science and Engineering, Ministry of Education Key Laboratory of Pollution Processes and Environmental Criteria, Nankai University, Tianjin 300071 (China); Luo, Yi, E-mail: luoy@nankai.edu.cn [College of Environmental Science and Engineering, Ministry of Education Key Laboratory of Pollution Processes and Environmental Criteria, Nankai University, Tianjin 300071 (China)

    2015-09-01

    This study investigated the characteristics of 10 subtypes of antibiotic resistance genes (ARGs) for sulfonamide, tetracycline, β-lactam and macrolide resistance and the class 1 integrase gene (intI1). In total, these genes were monitored in 24 samples across each stage of five full-scale pharmaceutical wastewater treatment plants (PWWTPs) using qualitative and real-time quantitative polymerase chain reactions (PCRs). The levels of typical ARG subtypes in the final effluents ranged from (2.08 ± 0.16) × 10{sup 3} to (3.68 ± 0.27) × 10{sup 6} copies/mL. The absolute abundance of ARGs in effluents accounted for only 0.6%–59.8% of influents of the five PWWTPs, while the majority of the ARGs were transported to the dewatered sludge with concentrations from (9.38 ± 0.73) × 10{sup 7} to (4.30 ± 0.81) × 10{sup 10} copies/g dry weight (dw). The total loads of ARGs discharged through dewatered sludge was 7–308 folds higher than that in the raw influents and 16–638 folds higher than that in the final effluents. The proliferation of ARGs mainly occurs in the biological treatment processes, such as conventional activated sludge, cyclic activated sludge system (CASS) and membrane bio-reactor (MBR), implying that significant replication of certain subtypes of ARGs may be attributable to microbial growth. High concentrations of antibiotic residues (ranging from 0.14 to 92.2 mg/L) were detected in the influents of selected wastewater treatment systems and they still remain high residues in the effluents. Partial correlation analysis showed significant correlations between the antibiotic concentrations and the associated relative abundance of ARG subtypes in the effluent. Although correlation does not prove causation, this study demonstrates that in addition to bacterial growth, the high antibiotic residues within the pharmaceutical WWTPs may influence the proliferation and fate of the associated ARG subtypes. - Highlights: • The ARGs in final discharges were 7

  11. Construction of an external electrode for determination of electrochemical corrosion potential in normal operational conditions of an BWR type reactor for hot cells; Construccion de un electrodo externo para determinacion del potencial de corrosion electroquimico en condiciones normales de operacion de un reactor tipo BWR para celdas calientes

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar T, J.A.; Rivera M, H.; Hernandez C, R. [Departamento de Sintesis y Caracterizacion de Materiales, Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    2001-07-01

    The behavior of the corrosion processes at high temperature requires of external devices that being capable to resist a temperature of 288 Centigrade and a pressure of 80 Kg/cm{sup 2}, to give stable and reproducible results of some variable and resisting physically and chemically the radiation. The external electrode of Ag/AgCl fulfils all the requirements in the determination of the electrochemical corrosion potential under normal operational conditions of a BWR type reactor in hot cells. (Author)

  12. Tesaglitazar, a dual peroxisome proliferator-activated receptor alpha/gamma agonist, improves apolipoprotein levels in non-diabetic subjects with insulin resistance

    DEFF Research Database (Denmark)

    Schuster, H.; Fagerberg, B.; Edwards, S.

    2008-01-01

    Aim: To determine the effects of the peroxisome proliferator-activated receptor (PPAR) alpha/gamma agonist tesaglitazar on serum levels of apolipoprotein (apo) A-I, apoB, and apoCIII in non-diabetic insulin-resistant subjects. Methods: This randomized, double-blind, multicentre, placebo...... associated with insulin resistance. (C) 2007 Elsevier Ireland Ltd. All rights reserved Udgivelsesdato: 2008/3...

  13. Peroxisome proliferator-activated receptor alpha improves pancreatic adaptation to insulin resistance in obese mice and reduces lipotoxicity in human islets

    NARCIS (Netherlands)

    Lalloyer, Fanny; Vandewalle, Brigitte; Percevault, Frederic; Torpier, Gerard; Kerr-Conte, Julie; Oosterveer, Maaike; Paumelle, Rejane; Fruchart, Jean-Charles; Kuipers, Folkert; Pattou, Francois; Fievet, Catherine; Staels, Bart

    Peroxisome proliferator-activated receptor (PPAR) alpha is a transcription factor controlling lipid and glucose homeostasis. PPAR alpha-deficient (-/-) mice are protected from high-fat diet-induced insulin resistance. However, the impact of PPAR alpha in the pathophysiological setting of

  14. Changes in expression of oestrogen regulated and proliferation genes with neoadjuvant treatment highlight heterogeneity of clinical resistance to the aromatase inhibitor, letrozole.

    Science.gov (United States)

    Miller, William R; Larionov, Alexey

    2010-01-01

    Clinical resistance is a major factor limiting benefits to endocrine therapy. Causes of resistance may be diverse and the mechanism of resistance in individual breast cancers is usually unknown. The present study illustrates how changes in the expression of proliferation and oestrogen-regulated genes occurring during neoadjuvant treatment with the aromatase inhibitor, letrozole, may define distinctive tumour subgroups and suggest different mechanisms of resistance in clinically endocrine resistant breast cancers. Postmenopausal women with large primary oestrogen-receptor (ER)-rich breast cancers were treated neoadjuvantly with letrozole (2.5 mg daily) for three months. Clinical response was determined by ultrasound changes in tumour volume. Tumour ribonucleic acid (RNA) from biopsies taken before, after 14 days and after three months of treatment was hybridized on Affymetrix U133A chips. Changes in expression of KIAA0101, TFF3, SERPINA3, IRS-1 and TFF1 were taken as markers of oestrogen regulation and those in CDC2, CKS-2, Cyclin B1, Thymidine Synthetase and PCNA as markers of proliferation. Fifteen tumours with < 50% volume reduction over three months of treatment were classified as being clinically non-responsive. Gene expression changes after 14 days of treatment with letrozole revealed different patterns of change in oestrogen regulated and proliferation genes in individual resistant tumours. Tumours could be separated into three different subgroups as follows: i) nine cases in which both proliferation and oestrogen signalling signatures were generally reduced on treatment (ii) four cases in which both signatures were generally unaffected or increased with treatment and (iii) two cases in which expression of the majority of oestrogen-regulated genes decreased whereas proliferation genes remained unchanged or increased. In 14 out of 15 tumours, RNA profiles were also available after three months of treatment. Patterns of change observed after 14 days were

  15. CNT1 expression influences proliferation and chemosensitivity in drug-resistant pancreatic cancer cells

    Science.gov (United States)

    Bhutia, Yangzom D.; Hung, Sau Wai; Patel, Bhavi; Lovin, Dylan; Govindarajan, Rajgopal

    2011-01-01

    Overcoming the inherent chemoresistance of pancreatic cancers remains a major goal of therapeutic investigations in this disease. In this study, we discovered a role for the human concentrative nucleoside transporter-1 (hCNT1; SLC28A1), a high-affinity pyrimidine nucleoside transporter, in determining the chemosensitivity of human pancreatic cancer cells to gemcitabine, the drug used presently as a standard of care. Compared with normal pancreas and pancreatic ductal epithelial cells, hCNT1 expression was frequently reduced in pancreatic tumors and tumor cell lines. In addition, hCNT1-mediated 3H-gemcitabine transport was lower in pancreatic cancer cell lines and correlated with cytotoxic IC50 estimations of gemcitabine. In contrast to gemcitabine-sensitive pancreatic cancer cell lines, MIA PaCa-2, a gemcitabine-resistant pancreatic cancer cell line exhibited relatively restrictive, cell cycle-dependent hCNT1 expression and transport. hCNT1 translation was suppressed in the late G1-enriched MIA PaCa-2 cell population possibly in an miRNA-dependent manner, which corresponded with the lowest hCNT1-mediated gemcitabine transport during this phase. While hCNT1 protein was induced during G1/S transition, increased hCNT1 trafficking resulted in maximal cell surface recruitment and transport-overshoot in the G2/M phase-enriched cell population. hCNT1 protein was directed predominantly to proteasomal or lysosomal degradation in S or G2/M phase MIA PaCa-2 cells, respectively. Pharmacological inhibition of hCNT1 degradation moderately increased cell surface hCNT1 expression and cellular gemcitabine transport in MIA PaCa-2 cells. Constitutive hCNT1 expression reduced clonogenic survival of MIA PaCa-2 cells and steeply augmented gemcitabine transport and chemosensitization. In addition to supporting a putative tumor suppressor role for hCNT1, our findings identify hCNT1 as a potential candidate to render drug-resistant pancreatic cancer cells amenable to chemotherapy. PMID

  16. Ell3 stimulates proliferation, drug resistance, and cancer stem cell properties of breast cancer cells via a MEK/ERK-dependent signaling pathway

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Hee-Jin [Department of Biomedical Science, College of Life Science, CHA University, Seoul (Korea, Republic of); Kim, Gwangil [Department of Pathology, CHA Bundang Medical Center, CHA University, Seoul (Korea, Republic of); Park, Kyung-Soon, E-mail: kspark@cha.ac.kr [Department of Biomedical Science, College of Life Science, CHA University, Seoul (Korea, Republic of)

    2013-08-09

    Highlights: •Ell3 enhances proliferation and drug resistance of breast cancer cell lines. •Ell3 is related to the cancer stem cell characteristics of breast cancer cell lines. •Ell3 enhances oncogenicity of breast cancer through the ERK1/2 signaling pathway. -- Abstract: Ell3 is a RNA polymerase II transcription elongation factor that is enriched in testis. The C-terminal domain of Ell3 shows strong similarities to that of Ell (eleven−nineteen lysine-rich leukemia gene), which acts as a negative regulator of p53 and regulates cell proliferation and survival. Recent studies in our laboratory showed that Ell3 induces the differentiation of mouse embryonic stem cells by protecting differentiating cells from apoptosis via the promotion of p53 degradation. In this study, we evaluated the function of Ell3 in breast cancer cell lines. MCF-7 cell lines overexpressing Ell3 were used to examine cell proliferation and cancer stem cell properties. Ectopic expression of Ell3 in breast cancer cell lines induces proliferation and 5-FU resistance. In addition, Ell3 expression increases the cancer stem cell population, which is characterized by CD44 (+) or ALDH1 (+) cells. Mammosphere-forming potential and migration ability were also increased upon Ell3 expression in breast cancer cell lines. Through biochemical and molecular biological analyses, we showed that Ell3 regulates proliferation, cancer stem cell properties and drug resistance in breast cancer cell lines partly through the MEK−extracellular signal-regulated kinase signaling pathway. Murine xenograft experiments showed that Ell3 expression promotes tumorigenesis in vivo. These results suggest that Ell3 may play a critical role in promoting oncogenesis in breast cancer by regulating cell proliferation and cancer stem cell properties via the ERK1/2 signaling pathway.

  17. Increasing alpha 7 beta 1-integrin promotes muscle cell proliferation, adhesion, and resistance to apoptosis without changing gene expression.

    Science.gov (United States)

    Liu, Jianming; Burkin, Dean J; Kaufman, Stephen J

    2008-02-01

    The dystrophin-glycoprotein complex maintains the integrity of skeletal muscle by associating laminin in the extracellular matrix with the actin cytoskeleton. Several human muscular dystrophies arise from defects in the components of this complex. The alpha(7)beta(1)-integrin also binds laminin and links the extracellular matrix with the cytoskeleton. Enhancement of alpha(7)-integrin levels alleviates pathology in mdx/utrn(-/-) mice, a model of Duchenne muscular dystrophy, and thus the integrin may functionally compensate for the absence of dystrophin. To test whether increasing alpha(7)-integrin levels affects transcription and cellular functions, we generated alpha(7)-integrin-inducible C2C12 cells and transgenic mice that overexpress the integrin in skeletal muscle. C2C12 myoblasts with elevated levels of integrin exhibited increased adhesion to laminin, faster proliferation when serum was limited, resistance to staurosporine-induced apoptosis, and normal differentiation. Transgenic expression of eightfold more integrin in skeletal muscle did not result in notable toxic effects in vivo. Moreover, high levels of alpha(7)-integrin in both myoblasts and in skeletal muscle did not disrupt global gene expression profiles. Thus increasing integrin levels can compensate for defects in the extracellular matrix and cytoskeleton linkage caused by compromises in the dystrophin-glycoprotein complex without triggering apparent overt negative side effects. These results support the use of integrin enhancement as a therapy for muscular dystrophy.

  18. Time-domain analysis of BWR core stability

    Energy Technology Data Exchange (ETDEWEB)

    Yokomizo, Osamu

    1983-01-01

    A time-domain stability analysis program for boiling water nuclear reactors (BWRs) has been developed and applied to analysis of a commercial size BWR. The program takes into account parallel channel effects. The model incorporates (a) one point neutron kinetics with weighted average reactivity feedback, (b) radial heat conduction and transfer in fuel rods, (c) fuel channel thermal hydraulics with quasi-equilibrium subcooled boiling approximation, and (d) recirculation hydrodynamics. Nonlinearity and parallel channel effects are examined through analyses of a commercial size BWR. Core behavior has been found virtually linear for small but finite amplitude oscillations, which proved the validity of frequency-domain stability analyses for finite disturbances. It has also been found that single channel analyses with core averaged thermal hydraulic properties give more stable results than parallel channel analyses.

  19. Condensate polishing guidelines for PWR and BWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Robbins, P.; Crinigan, P.; Graham, B.; Kohlmann, R.; Crosby, C.; Seager, J.; Bosold, R.; Gillen, J.; Kristensen, J.; McKeen, A.; Jones, V.; Sawochka, S.; Siegwarth, D.; Keeling, D.; Polidoroff, T.; Morgan, D.; Rickertsen, D.; Dyson, A.; Mills, W.; Coleman, L.

    1993-03-01

    Under EPRI sponsorship, an industry committee, similar in form and operation to other guideline committees, was created to develop Condensate Polishing Guidelines for both PWR and BWR systems. The committee reviewed the available utility and water treatment industry experience on system design and performance and incorporated operational and state-of-the-art information into document. These guidelines help utilities to optimize present condensate polisher designs as well as be a resource for retrofits or new construction. These guidelines present information that has not previously been presented in any consensus industry document. The committee generated guidelines that cover both deep bed and powdered resin systems as an integral part of the chemistry of PWR and BWR plants. The guidelines are separated into sections that deal with the basis for condensate polishing, system design, resin design and application, data management and performance and management responsibilities.

  20. Undermoderated spectrum MOX core study. Advanced fuel recycle by BWR

    Energy Technology Data Exchange (ETDEWEB)

    Matuyama, Shinichiro; Sakashita, Yoshiaki [Toshiba Corp., Kawasaki, Kanagawa (Japan)

    1998-09-01

    The concept of BARS (BWR with Advanced Recycle System) is described. This system is to recycle fuel for breeding type BWR using spent fuel by the dry reprocessing. At present, it is studying about the high spectrum core cooling with light water, the dry reprocessing, Vibration Compaction fuel and so on. In the dry reprocessing method used oxide, RE and DF are one of the technical issues. In the case that DF is about 10, RE doesn`t influence a core behavior. According to improve the present process, the possibility lies in making DF from 5 to equal or more than 10 sufficiently. Here, the outline, the development situation of these studies and the prospect of BARS from ability are explained. (author)

  1. Reduction of radiation exposure in Japanese BWR Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Morikawa, Yoshitake [ISOGO Nuclear Engineering Center, Yokohama (Japan)

    1995-03-01

    The reduction of occupational exposure to radiation during the annual inspection and maintenance outages of Japanese boiling water reactors (BWR) is one of the most important objectives for stable and reliable operation. It was shown that this radiation exposure is caused by radionuclides, such as Co-60, Co-58 and Mn-54 which are produced from the metal elements Co, Ni, and Fe present in the corrosion products of structural materials that had been irradiated by neutrons. Therefore, to reduce radiation sources and exposures in Japanese BWRs, attempts have been reinforced to remove corrosion products and activated corrosion products from the primary coolant system. This paper describes the progress of the application of these measures to Japanese BWRs. Most Japanese BWR-4 and BWR-5 type nuclear power plants started their commercial operations during the 1970s. With the elapse of time during operations, a problem came to the forefront, namely that occupational radiation exposure during plant outages gradually increased, which obstructed the smooth running of inspections and maintenance work. To overcome this problem, extensive studies to derive effective countermeasures for radiation exposure reduction were undertaken, based on the evaluation of the plants operation data.

  2. Determination of BWR Spent Nuclear Fuel Assembly Effective Thermal Conductivity

    Energy Technology Data Exchange (ETDEWEB)

    Matthew D. Hinds

    2001-10-17

    The purpose of this calculation is to provide an effective thermal conductivity for use in predicting peak cladding temperatures in boiling water reactor (BWR) fuel assemblies with 7x7,8x8, and 9x9 rod arrays. The first objective of this calculation is to describe the development and application of a finite element representation that predicts peak spent nuclear fuel temperatures for BWR assemblies. The second objective is to use the discrete representation to develop a basis for determining an effective thermal conductivity (described later) for a BWR assembly with srneared/homogeneous properties and to investigate the thermal behavior of a spent fuel assembly. The scope of this calculation is limited to a steady-state two-dimensional representation of the waste package interior region. This calculation is subject to procedure AP-3.124, Calculations (Ref. 27) and guided by the applicable technical work plan (Ref. 14). While these evaluations were originally developed for the thermal analysis of conceptual waste package designs emplaced in the potential repository at Yucca Mountain, the methodology applies to storage and transportation thermal analyses as well. Note that the waste package sketch in Attachment V depicts a preliminary design, and should not be interpreted otherwise.

  3. [Effects of triterpenoid from Psidium guajava leaves ursolic acid on proliferation, differentiation of 3T3-L1 preadipocyte and insulin resistance].

    Science.gov (United States)

    Lin, Juan-Na; Kuang, Qiao-Ting; Ye, Kai-He; Ye, Chun-Ling; Huang, Yi; Zhang, Xiao-Qi; Ye, Wen-Cai

    2013-08-01

    To investigate the influences of triterpenoid from Psidium guajava Leaves (ursolic acid) on the proliferation, differentiation of 3T3-L1 preadipocyte, and its possible mechanism treat for insulin resistance. 3T3-L1 preadipocyte was cultured in vitro. After adding ursolic acid to the culture medium for 48h, the cell viability was tested by MTT assay. Induced for 6 days, the lipid accumulation of adipocyte was measured by Oil Red O staining. The insulin resistant cell model was established with Dexamethasone. Cellular glucose uptake was determined with GOD-POD assays and FFA concentration was determined at the time of 48h. Secreted adiponectin were measured by ELISA. The protein levels of PPARgamma and PTP1B in insulin resistant adipocyte were measured by Western Blotting. Compared with medium control group, 30, 100 micromol/L ursolic acid could increase its proliferation and differentiation significantly (P 0.05). Ursolic acid can improve the proliferation and differentiation of 3T3-L1 preadipocyte, enhance cellular glucose uptake, inhibit the production of FFA, promote the secretion of adiponectin insulin resistant adipocyte, its mechanism may be related to upregulating the expression of PPARgamma protein.

  4. Condenser retubing-criteria manual. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Impagliazzo, A.M.; Bell, R.J.; Curlett, P.L.; Gordon, H.L.

    1982-05-01

    The objective of this document is to provide engineering assistance to utilities involved in retubing steam surface condensers with corrosion-resistant materials, such as titanium, and the recently developed high alloy pit-resistant steels. Field tests and recent operating experience have shown titanium and at least one of the high alloy pit-resistant steels to be virtually immune to the usual forms of corrosion occurring in steam surface condensers. This, together with the trend toward elimination of copper alloys in the circulating water system, has caused many utilities to retube their condensers with these materials.

  5. Thermohydraulic stability coupled to the neutronic in a BWR; Estabilidad termohidraulica acoplada a la neutronica en un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Calleros M, G.; Zapata Y, M.; Gomez H, R.A.; Mendez M, A. [Comision Federal de Electricidad, Central Nucleoelectrica de Laguna Verde, Carretera Cardel-Nautla Km. 42.5, Mpio. Alto Lucero, Veracruz (Mexico); Castlllo D, R. [ININ, Carretera Mexico-Toluca Km 36.5, La Marquesa, Estado de Mexico (Mexico)]. e-mail: gcm9acpp@cfe.gob.mx

    2006-07-01

    In a BWR type reactor the phenomenon of the nuclear fission is presented, in which are liberated in stochastic form neutrons, originating that the population of the same ones varies in statistic form around a mean value. This variation will cause that when the neutron flow impacts on the neutron detectors, its are had as a result neutron flow signals with fluctuations around an average value. In this article it is shown that it conforms it lapses the time, this variations in the neutron flow (and therefore, in the flow signal due only to the fission), they presented oscillations inside a stable range, which won't be divergent. Considering that the BWR is characterized because boiling phenomena are presented, which affect the moderation of the neutrons, additional variations will be had in the signal coming from the neutron detectors, with relationship to the fission itself, which will be influenced by the feedback of the moderator's reactivity and of the temperature of the fuel pellet. Also, as the BWR it has coupled control systems to maintain the coolant level one and of the thermal power of the reactor, for each control action it was affected the neutron population. This means that the reactor could end up straying of a stable state condition. By it previously described, the study of the thermohydraulic stability coupled to the neutronic is complex. In this work it is shown the phenomenology, the mathematical models and the theoretical behavior associated to the stability of the BWR type reactor; the variables that affect it are identified, the models that reproduce the behavior of the thermohydraulic stability coupled to the neutronic, the way to maintain stable the reactor and the instrumentation that can settle to detect and to suppress uncertainties is described. In particular, is make reference to the evolution of the methods to maintain the stability of the reactor and the detection system and suppression of uncertainties implemented in the

  6. IGF-1R and ErbB3/HER3 contribute to enhanced proliferation and carcinogenesis in trastuzumab-resistant ovarian cancer model

    Energy Technology Data Exchange (ETDEWEB)

    Jia, Yanhan [Department of Immunology, School of Basic Medical Sciences, Wuhan University, Wuhan, Hubei 430071 (China); Department of Immunology, Institute of Basic Medical Sciences, Beijing 100850 (China); Zhang, Yan [Department of Gynaecology and Obstetrics, PLA General Hospital, Beijing 100853 (China); Qiao, Chunxia; Liu, Guijun [Department of Immunology, Institute of Basic Medical Sciences, Beijing 100850 (China); Zhao, Qing [Department of Immunology, Institute of Basic Medical Sciences, Beijing 100850 (China); Department of Gynaecology and Obstetrics, PLA General Hospital, Beijing 100853 (China); Zhou, Tingting; Chen, Guojiang [Department of Immunology, Institute of Basic Medical Sciences, Beijing 100850 (China); Li, Yali [Department of Gynaecology and Obstetrics, PLA General Hospital, Beijing 100853 (China); Feng, Jiannan; Li, Yan [Department of Immunology, Institute of Basic Medical Sciences, Beijing 100850 (China); Zhang, Qiuping, E-mail: qpzhang@whu.edu.cn [Department of Immunology, School of Basic Medical Sciences, Wuhan University, Wuhan, Hubei 430071 (China); Peng, Hui, E-mail: p_h2002@hotmail.com [Department of Immunology, Institute of Basic Medical Sciences, Beijing 100850 (China); Cardiovascular Drug Research Center, Institute of Health and Environmental Medicine, Beijing 100850 (China)

    2013-07-12

    Highlights: •We established trastuzumab-resistant cell line SKOV3/T. •SKOV3/T enhances proliferation and in vivo carcinogenesis. •IGF-1R and HER3 genes were up-regulated in SKOV3/T based on microarray analysis. •Targeting IGF-1R and/or HER3 inhibited the proliferation of SKOV3/T. •Therapies targeting IGF-1R and HER3 might be effective in ovarian cancer. -- Abstract: Trastuzumab (Herceptin®) has demonstrated clinical potential in several types of HER2-overexpressing human cancers. However, primary and acquired resistance occurs in many HER2-positive patients with regimens. To investigate the possible mechanism of acquired therapeutic resistance to trastuzumab, we have developed a preclinical model of human ovarian cancer cells, SKOV3/T, with the distinctive feature of stronger carcinogenesis. The differences in gene expression between parental and the resistant cells were explored by microarray analysis, of which IGF-1R and HER3 were detected to be key molecules in action. Their correctness was validated by follow-up experiments of RT-PCR, shRNA-mediated knockdown, downstream signal activation, cell cycle distribution and survival. These results suggest that IGF-1R and HER3 differentially regulate trastuzumab resistance and could be promising targets for trastuzumab therapy in ovarian cancer.

  7. Cross-section adjustment techniques for BWR adaptive simulation

    Science.gov (United States)

    Jessee, Matthew Anderson

    Computational capability has been developed to adjust multi-group neutron cross-sections to improve the fidelity of boiling water reactor (BWR) modeling and simulation. The method involves propagating multi-group neutron cross-section uncertainties through BWR computational models to evaluate uncertainties in key core attributes such as core k-effective, nodal power distributions, thermal margins, and in-core detector readings. Uncertainty-based inverse theory methods are then employed to adjust multi-group cross-sections to minimize the disagreement between BWR modeling predictions and measured plant data. For this work, measured plant data were virtually simulated in the form of perturbed 3-D nodal power distributions with discrepancies with predictions of the same order of magnitude as expected from plant data. Using the simulated plant data, multi-group cross-section adjustment reduces the error in core k-effective to less than 0.2% and the RMS error in nodal power to 4% (i.e. the noise level of the in-core instrumentation). To ensure that the adapted BWR model predictions are robust, Tikhonov regularization is utilized to control the magnitude of the cross-section adjustment. In contrast to few-group cross-section adjustment, which was the focus of previous research on BWR adaptive simulation, multigroup cross-section adjustment allows for future fuel cycle design optimization to include the determination of optimal fresh fuel assembly designs using the adjusted multi-group cross-sections. The major focus of this work is to efficiently propagate multi-group neutron cross-section uncertainty through BWR lattice physics calculations. Basic neutron cross-section uncertainties are provided in the form of multi-group cross-section covariance matrices. For energy groups in the resolved resonance energy range, the cross-section uncertainties are computed using an infinitely-dilute approximation of the neutron flux. In order to accurately account for spatial and

  8. Y-box-binding protein-1 (YB-1) promotes cell proliferation, adhesion and drug resistance in diffuse large B-cell lymphoma

    Energy Technology Data Exchange (ETDEWEB)

    Miao, Xiaobing; Wu, Yaxun [Department of Pathology, Affiliated Cancer Hospital of Nantong University, Nantong 226361, Jiangsu (China); Wang, Yuchan [Department of Pathogen, Medical College, Nantong University, Nantong 226001, Jiangsu (China); Jiangsu Province Key Laboratory for Inflammation and Molecular Drug Target, Nantong University, Nantong 226001, Jiangsu (China); Zhu, Xinghua; Yin, Haibing [Department of Pathology, Affiliated Cancer Hospital of Nantong University, Nantong 226361, Jiangsu (China); He, Yunhua [Jiangsu Province Key Laboratory for Inflammation and Molecular Drug Target, Nantong University, Nantong 226001, Jiangsu (China); Li, Chunsun; Liu, Yushan; Lu, Xiaoyun; Chen, Yali; Shen, Rong [Department of Pathology, Affiliated Cancer Hospital of Nantong University, Nantong 226361, Jiangsu (China); Xu, Xiaohong, E-mail: xuxiaohongnantong@126.com [Department of Oncology, Affiliated Cancer Hospital of Nantong University, Nantong 226361, Jiangsu (China); He, Song, E-mail: hesongnt@126.com [Department of Pathology, Affiliated Cancer Hospital of Nantong University, Nantong 226361, Jiangsu (China)

    2016-08-15

    YB-1 is a multifunctional protein, which has been shown to correlate with resistance to treatment of various tumor types. This study investigated the expression and biologic function of YB-1 in diffuse large B-cell lymphoma (DLBCL). Immunohistochemical analysis showed that the expression statuses of YB-1 and pYB-1{sup S102} were reversely correlated with the clinical outcomes of DLBCL patients. In addition, we found that YB-1 could promote the proliferation of DLBCL cells by accelerating the G1/S transition. Ectopic expression of YB-1 could markedly increase the expression of cell cycle regulators cyclin D1 and cyclin E. Furthermore, we found that adhesion of DLBCL cells to fibronectin (FN) could increase YB-1 phosphorylation at Ser102 and pYB-1{sup S102} nuclear translocation. In addition, overexpression of YB-1 could increase the adhesion of DLBCL cells to FN. Intriguingly, we found that YB-1 overexpression could confer drug resistance through cell-adhesion dependent and independent mechanisms in DLBCL. Silencing of YB-1 could sensitize DLBCL cells to mitoxantrone and overcome cell adhesion-mediated drug resistance (CAM-DR) phenotype in an AKT-dependent manner. - Highlights: • The expression statuses of YB-1 and pYB-1{sup S102} are reversely correlated with outcomes of DLBCL patients. • YB-1 promotes cell proliferation by accelerating G1/S transition in DLBCL. • YB-1 confers drug resistance to mitoxantrone in DLBCL.

  9. Prony's method application for BWR instabilities characterization

    Energy Technology Data Exchange (ETDEWEB)

    Castillo, Rogelio, E-mail: rogelio.castillo@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carretera México-Toluca s/n, La Marquesa, Ocoyoacac, Estado de México 52750 (Mexico); Ramírez, J. Ramón, E-mail: ramon.ramirez@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carretera México-Toluca s/n, La Marquesa, Ocoyoacac, Estado de México 52750 (Mexico); Alonso, Gustavo, E-mail: gustavo.alonso@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carretera México-Toluca s/n, La Marquesa, Ocoyoacac, Estado de México 52750 (Mexico); Instituto Politecnico Nacional, Unidad Profesional Adolfo Lopez Mateos, Ed. 9, Lindavista, D.F. 07300 (Mexico); Ortiz-Villafuerte, Javier, E-mail: javier.ortiz@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carretera México-Toluca s/n, La Marquesa, Ocoyoacac, Estado de México 52750 (Mexico)

    2015-04-01

    Highlights: • Prony's method application for BWR instability events. • Several BWR instability benchmark are assessed using this method. • DR and frequency are obtained and a new parameter is proposed to eliminate false signals. • Adequate characterization of in-phase and out-of-phase events is obtained. • The Prony's method application is validated. - Abstract: Several methods have been developed for the analysis of reactor power signals during BWR power oscillations. Among them is the Prony's method, its application provides the DR and the frequency of oscillations. In this paper another characteristic of the method is proposed to determine the type of oscillations that can occur, in-phase or out-of-phase. Prony's method decomposes a given signal in all the frequencies that it contains, therefore the DR of the fundamental mode and the first harmonic are obtained. To determine the more dominant pole of the system a normalized amplitude W of the system is calculated, which depends on the amplitude and the damping coefficient. With this term, it can be analyzed which type of oscillations is present, if W of the fundamental mode frequency is the greater, the type of oscillations is in-phase, if W of the first harmonic frequency is the greater, the type of oscillations is out-of-phase. The method is applied to several stability benchmarks to assess its validity. Results show the applicability of the method as an alternative analysis method to determine the type of oscillations occurred.

  10. Prevention of organic iodide formation in BWR`s

    Energy Technology Data Exchange (ETDEWEB)

    Karjunen, T. [Finnish Centre for Radiation and Nuclear Safety, Helsinki (Finland); Laitinen, T.; Piippo, J.; Sirkiae, P. [VTT Manufacturing Technology (Finland)

    1996-12-01

    During an accident, many different forms of iodine may emerge. Organic iodides, such as methyl iodide and ethyl iodide, are relatively volatile, and thus their appearance leads to increased concentration of gaseous iodine. Since organic iodides are also relatively immune to most accident mitigation measures, such as sprays and filters, they can affect the accident source term significantly even when only a small portion of iodine is in organic form. Formation of organic iodides may not be limited by the amount of organic substances available. Excessive amounts of methane can be produced, for example, during oxidation of boron carbide, which is used in BWR`s as a neutron absorber material. Another important source is cable insulation. In a BWR, a large quantity of cables is placed below the pressure vessel. Thus a large quantity of pyrolyse gases will be produced, should the vessel fail. Organic iodides can be formed as a result of many different reactions, but at least in certain conditions the main reaction takes place between an organic radical produced by radiolysis and elemental iodine. A necessary requirement for prevention of organic iodide production is therefore that the pH in the containment water pools is kept high enough to eliminate formation of elemental iodine. In a typical BWR the suppression pool water is usually unbuffered. As a result, the pH may be dominated by chemicals introduced during an accident. If no system for adding basic chemicals is operable, the main factor affecting pool water pH may be hydrochloric acid released during cable degradation. Should this occur, the conditions could be very favorable for production of elemental iodine and, consequently, formation of organic iodides. Although high pH is necessary for iodine retention, it could have also adverse effects. High pH may, for example, accelerate corrosion of containment materials and alter the characteristics of the solid corrosion products. (author) 6 figs., 1 tab., 13 refs.

  11. Safety analysis of thorium-based fuels in the General Electric Standard BWR

    Energy Technology Data Exchange (ETDEWEB)

    Colby, M.J.; Townsend, D.B.; Kunz, C.L.

    1980-06-01

    A denatured (U-233/Th)O/sub 2/ fuel assembly has been designed which is energy equivalent to and hardware interchangeable with a modern boiling water reactor (BWR) reference reload assembly. Relative to the reference UO/sub 2/ fuel, the thorium fuel design shows better performance during normal and transient reactor operation for the BWR/6 product line and will meet or exceed current safety and licensing criteria. Power distributions are flattened and thermal operating margins are increased by reduced steam void reactivity coefficients caused by U-233. However, a (U-233/Th)O/sub 2/-fueled BWR will likely have reduced operating flexibility. A (U-235/Th)O/sub 2/-fueled BWR should perform similar to a UO/sub 2/-fueled BWR under all operating conditions. A (Pu/Th)O/sub 2/-fueled BWR may have reduced thermal margins and similar accident response and be less stable than a UO/sub 2/-fueled BWR. The assessment is based on comparisions of point model and infinite lattice predictions of various nuclear reactivity parameters, including void reactivity coefficients, Doppler reactivity coefficients, and control blade worths.

  12. Bone stroma-derived cells change coregulators recruitment to androgen receptor and decrease cell proliferation in androgen-sensitive and castration-resistant prostate cancer cells

    Energy Technology Data Exchange (ETDEWEB)

    Villagran, Marcelo A.; Gutierrez-Castro, Francisco A.; Pantoja, Diego F.; Alarcon, Jose C.; Fariña, Macarena A.; Amigo, Romina F.; Muñoz-Godoy, Natalia A. [Molecular Endocrinology and Oncology Laboratory, University of Concepcion, Concepcion (Chile); Pinilla, Mabel G. [Department of Medical Specialties, School of Medicine, University of Concepcion, Concepcion (Chile); Peña, Eduardo A.; Gonzalez-Chavarria, Ivan; Toledo, Jorge R.; Rivas, Coralia I.; Vera, Juan C. [Department of Physiopathology, School of Biological Sciences, University of Concepcion, Concepcion (Chile); McNerney, Eileen M. [Molecular Endocrinology and Oncology Laboratory, University of Concepcion, Concepcion (Chile); Onate, Sergio A., E-mail: sergio.onate@udec.cl [Molecular Endocrinology and Oncology Laboratory, University of Concepcion, Concepcion (Chile); Department of Medical Specialties, School of Medicine, University of Concepcion, Concepcion (Chile); Department of Urology, State University of New York at Buffalo, NY (United States)

    2015-11-27

    Prostate cancer (CaP) bone metastasis is an early event that remains inactive until later-stage progression. Reduced levels of circulating androgens, due to andropause or androgen deprivation therapies, alter androgen receptor (AR) coactivator expression. Coactivators shift the balance towards enhanced AR-mediated gene transcription that promotes progression to androgen-resistance. Disruptions in coregulators may represent a molecular switch that reactivates latent bone metastasis. Changes in AR-mediated transcription in androgen-sensitive LNCaP and androgen-resistant C4-2 cells were analyzed for AR coregulator recruitment in co-culture with Saos-2 and THP-1. The Saos-2 cell line derived from human osteosarcoma and THP-1 cell line representing human monocytes were used to display osteoblast and osteoclast activity. Increased AR activity in androgen-resistant C4-2 was due to increased AR expression and SRC1/TIF2 recruitment and decreased SMRT/NCoR expression. AR activity in both cell types was decreased over 90% when co-cultured with Saos-2 or THP-1 due to dissociation of AR from the SRC1/TIF2 and SMRT/NCoR coregulators complex, in a ligand-dependent and cell-type specific manner. In the absence of androgens, Saos-2 decreased while THP-1 increased proliferation of LNCaP cells. In contrast, both Saos-2 and THP-1 decreased proliferation of C4-2 in absence and presence of androgens. Global changes in gene expression from both CaP cell lines identified potential cell cycle and androgen regulated genes as mechanisms for changes in cell proliferation and AR-mediated transactivation in the context of bone marrow stroma cells. - Highlights: • Decreased corepressor expression change AR in androgen-resistance prostate cancer. • Bone stroma-derived cells change AR coregulator recruitment in prostate cancer. • Bone stroma cells change cell proliferation in androgen-resistant cancer cells. • Global gene expression in CaP cells is modified by bone stroma cells in co

  13. Neutron dosimetry. Environmental monitoring in a BWR type reactor; Dosimetria de neutrones. Monitoreo ambiental en un reactor del tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Tavera D, L.; Camacho L, M.E

    1991-01-15

    The measurements carried out on reactor dosimetry are applied mainly to the study on the effects of the radiation in 108 materials of the reactor; little is on the environmental dosimetry outside of the primary container of BWR reactors. In this work the application of a neutron spectrometer formed by plastic detectors of nuclear traces manufactured in the ININ, for the environmental monitoring in penetrations around the primary container of the unit I of the Laguna Verde central is presented. The neutron monitoring carries out with purposes of radiological protection, during the operational tests of the reactor. (Author)

  14. BWR Refill-Reflood Program, Task 4. 7 - model development: basic models for the BWR version of TRAC

    Energy Technology Data Exchange (ETDEWEB)

    Andersen, J G.M.; Chu, K H; Shaug, J C

    1983-09-01

    TRAC (Transient Reactor Analysis Code) is a computer code for best estimate analysis of the thermal hydraulic conditions in a reactor system. The constitutive correlations for shear and heat transfer developed for the Boiling Water Reactor (BWR) version of TRAC are described. A universal flow regime map has been developed to tie the regimes for shear and heat transfer into a consistent package. New models in the areas of interfacial shear, interfacial heat transfer and thermal radiation have been introduced. Improvements have also been made to the constitutive correlations and the numerical methods. All the models have been implemented into the GE version TRACB02 and extensively tested against data.

  15. Malignant T cells exhibit CD45 resistant Stat 3 activation and proliferation in cutaneous T cell lymphoma

    DEFF Research Database (Denmark)

    Krejsgaard, T; Helvad, Rikke; Ralfkiær, Elisabeth

    2010-01-01

    CD45 is a protein tyrosine phosphatase, which is well-known for regulating antigen receptor signalling in T and B cells via its effect on Src kinases. It has recently been shown that CD45 can also dephosphorylate Janus kinases (Jaks) and thereby regulate Signal transducer and activator...... of transcription (Stat) activation and cytokine-induced proliferation in lymphocytes. Consequently, CD45 dysregulation could be implicated in aberrant Jak/Stat activation and proliferation in lymphoproliferative diseases. Despite high expression of the CD45 ligand, Galectin-1, in skin lesions from cutaneous T......-cell lymphoma (CTCL), the malignant T cells exhibit constitutive activation of the Jak3/Stat3 signalling pathway and uncontrolled proliferation. We show that CD45 expression is down-regulated on malignant T cells when compared to non-malignant T cells established from CTCL skin lesions. Moreover, CD45 cross...

  16. Identification of Important Compounds Isolated from Natural Sources that Have Activity Against Multidrug-resistant Cancer Cell Lines: Effects on Proliferation, Apoptotic Mechanism and the Efflux Pump Responsible for Multi-resistance Phenotype.

    Science.gov (United States)

    Amaral, Leonard; Spengler, Gabriella; Molnar, Joseph

    2016-11-01

    The focus of this mini-review is to identify non-toxic compounds isolated from natural sources (plants) that exhibit specific activity against efflux pumps of specific multidrug-resistant (MDR) cancer cell lines, inhibit proliferation of the MDR cancer cell lines and inhibit the activity of overexpressed efflux pumps of the MDR cancer cell line. Copyright© 2016 International Institute of Anticancer Research (Dr. John G. Delinassios), All rights reserved.

  17. Effect of anisotropic scattering in neutronics analysis of BWR assembly

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Toshikazu [Division of Sustainable Energy and Environmental Engineering, Graduate School of Engineering, Osaka University, Yamadaoka 2-1, Suita, Osaka 565-0571 (Japan)]. E-mail: takeda@nucl.eng.osaka-u.ac.jp; Okamoto, Toshiki [Division of Sustainable Energy and Environmental Engineering, Graduate School of Engineering, Osaka University, Yamadaoka 2-1, Suita, Osaka 565-0571 (Japan); Inoue, Akira [Division of Sustainable Energy and Environmental Engineering, Graduate School of Engineering, Osaka University, Yamadaoka 2-1, Suita, Osaka 565-0571 (Japan); Kosaka, Shinya [TEPCO Systems Corporation, 2-37-28 Eitai, Koutou-ku, Tokyo 135-0034 (Japan); Ikeda, Hideaki [TEPCO Systems Corporation, 2-37-28 Eitai, Koutou-ku, Tokyo 135-0034 (Japan)

    2006-11-15

    The anisotropic scattering effect to keff is studied for UO{sub 2} and MOX fueled BWR assemblies. The anisotropic scattering effect increases the assembly k {sub {infinity}} by 0.44% {delta}k for the UO{sub 2} assembly with 0% void fraction, and by 0.21% {delta}k for the MOX assembly with 0% void fraction. This is because the anisotropic scattering effect flattens the intra-assembly thermal flux, and the absorption rate in the surrounding water gap is decreased, but the absorption rates in the MOX fuel rods are increased compared to the UO{sub 2} rods. Therefore, the total decrease in absorption rates in the UO{sub 2} assembly is relatively large, and the k {sub {infinity}} is increased in the UO{sub 2} assembly. The dependence of the anisotropic scattering effect on the void fraction is investigated, and the significant difference of 0.62% {delta}k/k is found for the 0% and the 80% void fractions. The BWR assemblies with Gd rods are also considered. Furthermore, the usefulness of the transport cross section is investigated, and it is found that the transport cross section gives reasonable anisotropic scattering effect, though not satisfactory.

  18. Hydrogen injection in BWR and related radiation chemistry

    Science.gov (United States)

    Ishigure, Kenkichi; Takagi, Junichi; Shiraishi, Hirotsugu

    Hydrogen injection to feed water systems in boiling water reactors (BWR) has drawn wide attention as one of the possible countermeasures to the stress corrosion cracking (SCC) of 304 type stainless steel piping. To confirm the effectiveness of the hydrogen injection, a computer simulation of the complicated radiolysis reactions was carried out. The result of the simulation showed that the reactor water data monitored at the usual sampling points in actual plants reflect mainly the reactions in the downcomer portion but not in the reactor core in BWR. The calculation claimed approximately 300 ppb hydrogen in feed water to reduce the oxygen concentration in the recirculation lines to a negligible level, while one order of magnitude higher level of hydrogen is necessary to suppress oxygen in the reactor core. The computer simulation requires many radiation chemical data as in-put, among which are G values of initial products for water radiolysis at high temperature. An experimental approach was made to confirm the G values for high temperature radiolysis of water. The result does not seem to be consistent with the high temperature G values reported by Burns.

  19. BWR Servicing and Refueling Improvement Program: Phase I summary report

    Energy Technology Data Exchange (ETDEWEB)

    Perry, D.R.

    1978-09-01

    Under the U.S. Department of Energy sponsorship, General Electric Co. (GE) undertook a study of boiling water reactor (BWR) refueling outages for the purpose of recommending the development and demonstration of critical path time savings improvements. The Tennessee Valley Authority (TVA) joined the study as a subcontractor, providing monitoring assistance and making the Browns Ferry Site available for improvement demonstrations. Agreement was also reached with Georgia Power Co., Power Authority of the State of New York, and Commonwealth Edison Co. for monitoring and data collection at Hatch 1, FitzPatrick, and Quad Cities 1 nuclear plants, respectively. The objective was to identify, develop, and demonstrate improved refueling, maintenance, and inspection procedures and equipment. The improvements recommended in this study are applicable to BWR nuclear plants currently in operation as well as those in the design and construction phases. The recommendations and outage information can be used as a basis to plan and conduct the first outages of new plants and to improve the planning and facilities of currently operating plants. Many of the recommendations can readily be incorporated in plants currently in the design and construction phases as well as in the design of future plants. Many of these recommended improvements can be implemented immediately by utilities without further technical development.

  20. Krüppel-like Factor 5 contributes to pulmonary artery smooth muscle proliferation and resistance to apoptosis in human pulmonary arterial hypertension

    Directory of Open Access Journals (Sweden)

    Paulin Roxane

    2011-09-01

    Full Text Available Background Pulmonary arterial hypertension (PAH is a vascular remodeling disease characterized by enhanced proliferation of pulmonary artery smooth muscle cell (PASMC and suppressed apoptosis. This phenotype has been associated with the upregulation of the oncoprotein survivin promoting mitochondrial membrane potential hyperpolarization (decreasing apoptosis and the upregulation of growth factor and cytokines like PDGF, IL-6 and vasoactive agent like endothelin-1 (ET-1 promoting PASMC proliferation. Krüppel-like factor 5 (KLF5, is a zinc-finger-type transcription factor implicated in the regulation of cell differentiation, proliferation, migration and apoptosis. Recent studies have demonstrated the implication of KLF5 in tissue remodeling in cardiovascular diseases, such as atherosclerosis, restenosis, and cardiac hypertrophy. Nonetheless, the implication of KLF5 in pulmonary arterial hypertension (PAH remains unknown. We hypothesized that KLF5 up-regulation in PAH triggers PASMC proliferation and resistance to apoptosis. Methods and results We showed that KFL5 is upregulated in both human lung biopsies and cultured human PASMC isolated from distal pulmonary arteries from PAH patients compared to controls. Using stimulation experiments, we demonstrated that PDGF, ET-1 and IL-6 trigger KLF-5 activation in control PASMC to a level similar to the one seen in PAH-PASMC. Inhibition of the STAT3 pathway abrogates KLF5 activation in PAH-PASMC. Once activated, KLF5 promotes cyclin B1 upregulation and promotes PASMC proliferation and triggers survivin expression hyperpolarizing mitochondria membrane potential decreasing PASMC ability to undergo apoptosis. Conclusion We demonstrated for the first time that KLF5 is activated in human PAH and implicated in the pro-proliferative and anti-apoptotic phenotype that characterize PAH-PASMC. We believe that our findings will open new avenues of investigation on the role of KLF5 in PAH and might lead to the

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-07-01

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

  2. LncRNA UCA1 promotes proliferation and cisplatin resistance of oral squamous cell carcinoma by sunppressing miR-184 expression.

    Science.gov (United States)

    Fang, Zheng; Zhao, Junfang; Xie, Weihong; Sun, Qiang; Wang, Haibin; Qiao, Bin

    2017-12-01

    Chemotherapy resistance has become the main obstacle for the effective treatment of human cancers. Long non-coding RNA urothelial cancer associated 1 (UCA1) is generally regarded as an oncogene in some cancers. However, the function and molecular mechanism of UCA1 implicated in cisplatin (CDDP) chemoresistance of oral squamous cell carcinoma (OSCC) is still not fully established. UCA1 expression in tumor tissues and cells was tested by qRT-PCR. MTT, flow cytometry and caspase-3 activity analysis were explored to evaluate the CDDP sensitivity in OSCC cells. Western blot analysis was used to measure BCL2, Bax and SF1 protein expression. Luciferase reporter assay was conducted to investigate the molecular relationship between UCA1, miR-184, and SF1. Nude mice model was used to confirm the functional role of UCA1 in CDDP resistance in vivo. UCA1 expression was upregulated in OSCC tissues, cell lines, and CDDP resistant OSCC cells. Function analysis revealed that UCA1 facilitated proliferation, enhanced CDDP chemoresistance, and suppressed apoptosis in OSCC cells. Mechanisms investigation indicated that UCA1 could interact with miR-184 to repress its expression. Rescue experiments suggested that downregulation of miR-184 partly reversed the tumor suppression effect and CDDP chemosensitivity of UCA1 knockdown in CDDP-resistant OSCC cells. Moreover, UCA1 could perform as a miR-184 sponge to modulate SF1 expression. The OSCC nude mice model experiments demonstrated that depletion of UCA1 further boosted CDDP-mediated repression effect on tumor growth. UCA1 accelerated proliferation, increased CDDP chemoresistance and restrained apoptosis partly through modulating SF1 via sponging miR-184 in OSCC cells, suggesting that targeting UCA1 may be a potential therapeutic strategy for OSCC patients. © 2017 The Authors. Cancer Medicine published by John Wiley & Sons Ltd.

  3. Malignant T cells exhibit CD45 resistant Stat 3 activation and proliferation in cutaneous T cell lymphoma

    DEFF Research Database (Denmark)

    Krejsgaard, T; Helvad, Rikke; Ralfkiær, Elisabeth

    2010-01-01

    CD45 is a protein tyrosine phosphatase, which is well-known for regulating antigen receptor signalling in T and B cells via its effect on Src kinases. It has recently been shown that CD45 can also dephosphorylate Janus kinases (Jaks) and thereby regulate Signal transducer and activator of transcr......CD45 is a protein tyrosine phosphatase, which is well-known for regulating antigen receptor signalling in T and B cells via its effect on Src kinases. It has recently been shown that CD45 can also dephosphorylate Janus kinases (Jaks) and thereby regulate Signal transducer and activator...... of transcription (Stat) activation and cytokine-induced proliferation in lymphocytes. Consequently, CD45 dysregulation could be implicated in aberrant Jak/Stat activation and proliferation in lymphoproliferative diseases. Despite high expression of the CD45 ligand, Galectin-1, in skin lesions from cutaneous T......-cell lymphoma (CTCL), the malignant T cells exhibit constitutive activation of the Jak3/Stat3 signalling pathway and uncontrolled proliferation. We show that CD45 expression is down-regulated on malignant T cells when compared to non-malignant T cells established from CTCL skin lesions. Moreover, CD45 cross...

  4. Mouse C2 myoblast cells resist HVJ (Sendai virus)-mediated cell fusion in the proliferating stage but become capable of fusion after differentiation.

    Science.gov (United States)

    Hirayama, E; Nakanishi, M; Honda, N; Kim, J

    1999-05-01

    To investigate the mechanism of myoblast fusion, we attempted to prepare artificial myotubes of mouse C2 myoblast cells using the hemagglutinating virus of Japan (HVJ, Sendai virus). Proliferating C2 cells showed strong resistance to HVJ-mediated cell fusion and remained morphologically unchanged even though massive numbers of virions adsorbed onto their surface. They showed no membrane disruption, which occurs in the early stage of cell fusion induced by HVJ. These observations suggest that proliferating C2 cells are resistant to HVJ-mediated cell fusion. However, upon induction of differentiation, C2 cells gradually became capable of fusion induced by HVJ and then even generated heterokaryons with Ehrlich ascites tumor cells. When differentiated C2 cells that had become fusion-sensitive were treated with HVJ in the presence of EDTA, they did not fuse but degenerated, suggesting that their cell membranes were transiently disrupted by interaction with HVJ. These results suggest that the cell membranes of myoblasts change to a fusion-capable state during the process of differentiation.

  5. BWR plant analyzer development at BNL (Brookhaven National Laboratory)

    Energy Technology Data Exchange (ETDEWEB)

    Wulff, W.; Cheng, H.S.; Mallen, A.N.

    1986-01-01

    An engineering plant analyzer has been developed at BNL for realistically and accurately simulating transients and severe abnormal events in BWR power plants. Simulations are being carried out routinely with high fidelity, high simulation speed, at low cost and with unsurpassed user convenience. The BNL Plant Analyzer is the only operating facility which (a) simulates more than two orders-of-magnitude faster than the CDC-7600 mainframe computer, (b) is accessible and fully operational in on-line interactive mode, remotely from anywhere in the US, from Europe or the Far East (Korea), via widely available IBM-PC compatible personal computers, standard modems and telephone lines, (c) simulates both slow and rapid transients seven times faster than real-time in direct access, and four times faster in remote access modes, (d) achieves high simulation speed without compromising fidelity, and (e) is available to remote access users at the low cost of $160 per hour.

  6. Oxide evolution on Alloy X-750 in simulated BWR environment

    Science.gov (United States)

    Tuzi, Silvia; Göransson, Kenneth; Rahman, Seikh M. H.; Eriksson, Sten G.; Liu, Fang; Thuvander, Mattias; Stiller, Krystyna

    2016-12-01

    In order to simulate the environment experienced by spacer grids in a boiling water reactor (BWR), specimens of the Ni-based Alloy X-750 were exposed to a water jet in an autoclave at a temperature of 286 °C and a pressure of 80 bar. The oxide microstructure of specimens exposed for 2 h, 24 h, 168 h and 840 h has been investigated mainly using electron microscopy. The specimens suffer mass loss due to dissolution during exposure. At the same time a complex layered oxide develops. After the longest exposure the oxide consists of two outer spinel layers consisting of blocky crystals, one intermediate layer of nickel oxide interspersed with Ti-rich oxide needles, and an inner layer of oxidized base metal. The evolution of the oxide leading up to this structure is discussed and a model is presented.

  7. Peroxisome Proliferator-Activated Receptors in Regulation of Cytochromes P450: New Way to Overcome Multidrug Resistance?

    OpenAIRE

    Katerina Cizkova; Anna Konieczna; Bela Erdosova; Radka Lichnovska; Jiri Ehrmann

    2012-01-01

    Embryonic and tumour cells are able to protect themselves against various harmful compounds. In human pathology, this phenomenon exists in the form of multidrug resistance (MDR) that significantly deteriorates success of anticancer treatment. Cytochromes P450 (CYPs) play one of the key roles in the xenobiotic metabolism. CYP expression could contribute to resistance of cancer cells to chemotherapy. CYP epoxygenases (CYP2C and CYP2J) metabolize about 20% of clinically important drugs. Besides ...

  8. PWR and BWR spent fuel assembly gamma spectra measurements

    Science.gov (United States)

    Vaccaro, S.; Tobin, S. J.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Hu, J.; Schwalbach, P.; Sjöland, A.; Trellue, H.; Vo, D.

    2016-10-01

    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative-Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

  9. PWR and BWR spent fuel assembly gamma spectra measurements

    Energy Technology Data Exchange (ETDEWEB)

    Vaccaro, S. [European Commission, DG Energy, Directorate EURATOM Safeguards Luxembourg (Luxembourg); Tobin, S.J.; Favalli, A. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Hu, J. [Oak Ridge National Laboratory, Oak Ridge (United States); Schwalbach, P. [European Commission, DG Energy, Directorate EURATOM Safeguards Luxembourg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company (SKB) (Sweden); Trellue, H.; Vo, D. [Los Alamos National Laboratory, Los Alamos, NM (United States)

    2016-10-11

    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of {sup 137}Cs, {sup 154}Eu, and {sup 134}Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

  10. Over-expression of 60s ribosomal L23a is associated with cellular proliferation in SAG resistant clinical isolates of Leishmania donovani.

    Directory of Open Access Journals (Sweden)

    Sanchita Das

    Full Text Available Sodium antimony gluconate (SAG unresponsiveness of Leishmania donovani (Ld had effectively compromised the chemotherapeutic potential of SAG. 60s ribosomal L23a (60sRL23a, identified as one of the over-expressed protein in different resistant strains of L.donovani as observed with differential proteomics studies indicates towards its possible involvement in SAG resistance in L.donovani. In the present study 60sRL23a has been characterized for its probable association with SAG resistance mechanism.The expression profile of 60s ribosomal L23a (60sRL23a was checked in different SAG resistant as well as sensitive strains of L.donovani clinical isolates by real-time PCR and western blotting and was found to be up-regulated in resistant strains. Ld60sRL23a was cloned, expressed in E.coli system and purified for raising antibody in swiss mice and was observed to have cytosolic localization in L.donovani. 60sRL23a was further over-expressed in sensitive strain of L.donovani to check its sensitivity profile against SAG (Sb V and III and was found to be altered towards the resistant mode.This study reports for the first time that the over expression of 60sRL23a in SAG sensitive parasite decreases the sensitivity of the parasite towards SAG, miltefosine and paramomycin. Growth curve of the tranfectants further indicated the proliferative potential of 60sRL23a assisting the parasite survival and reaffirming the extra ribosomal role of 60sRL23a. The study thus indicates towards the role of the protein in lowering and redistributing the drug pressure by increased proliferation of parasites and warrants further longitudinal study to understand the underlying mechanism.

  11. PPARγ activation alters fatty acid composition in adipose triglyceride, in addition to proliferation of small adipocytes, in insulin resistant high-fat fed rats.

    Science.gov (United States)

    Sato, Daisuke; Oda, Kanako; Kusunoki, Masataka; Nishina, Atsuyoshi; Takahashi, Kazuaki; Feng, Zhonggang; Tsutsumi, Kazuhiko; Nakamura, Takao

    2016-02-15

    It was reported that adipocyte size is potentially correlated in part to amount of long chain polyunsaturated fatty acids (PUFAs) and insulin resistance because several long chain PUFAs can be ligands of peroxisome proliferator-activated receptors (PPARs). In our previous study, marked reduction of PUFAs was observed in insulin-resistant high-fat fed rats, which may indicate that PUFAs are consumed to improve insulin resistance. Although PPARγ agonist, well known as an insulin sensitizer, proliferates small adipocytes, the effects of PPARγ agonist on FA composition in adipose tissue have not been clarified yet. In the present study, we administered pioglitazone, a PPARγ agonist, to high-fat fed rats, and measured their FA composition of triglyceride fraction in adipose tissue and adipocyte diameters in pioglitazone-treated (PIO) and non-treated (control) rats. Insulin sensitivity was obtained with hyperinsulinemic euglycemic clamp. Average adipocyte diameter in the PIO group were smaller than that in the control one without change in tissue weight. In monounsaturated FAs (MUFAs), 14:1n-5, 16:1n-7, and 18:1n-9 contents in the PIO group were lower than those, respectively, in the control group. In contrast, 22:6n-3, 20:3n-6, 20:4n-6, and 22:4n-6 contents in the PIO group were higher than those, respectively, in the control group. Insulin sensitivity was higher in the PIO group than in the control one. These findings suggest that PPARγ activation lowered MUFAs whereas suppressed most of C20 or C22 PUFAs reduction, and that the change of fatty acid composition may be relevant with increase in small adipocytes. Copyright © 2016 Elsevier B.V. All rights reserved.

  12. ENO1 promotes tumor proliferation and cell adhesion mediated drug resistance (CAM-DR) in Non-Hodgkin's Lymphomas

    Energy Technology Data Exchange (ETDEWEB)

    Zhu, Xinghua; Miao, Xiaobing; Wu, Yaxun; Li, Chunsun; Guo, Yan; Liu, Yushan; Chen, Yali; Lu, Xiaoyun [Department of Pathology, Affiliated Cancer Hospital of Nantong University, 30 North Tongyang Road, Pingchao, Nantong 226361, Jiangsu (China); Wang, Yuchan, E-mail: wangyuchannt@126.com [Department of Pathogen and Immunology, Medical College, Nantong University, 19 Qixiu Road, Nantong 226001, Jiangsu (China); He, Song, E-mail: hesongnt@126.com [Department of Pathology, Affiliated Cancer Hospital of Nantong University, 30 North Tongyang Road, Pingchao, Nantong 226361, Jiangsu (China)

    2015-07-15

    Enolases are glycolytic enzymes responsible for the ATP-generated conversion of 2-phosphoglycerate to phosphoenolpyruvate. In addition to the glycolytic function, Enolase 1 (ENO1) has been reported up-regulation in several tumor tissues. In this study, we investigated the expression and biologic function of ENO1 in Non-Hodgkin's Lymphomas (NHLs). Clinically, by western blot analysis we observed that ENO1 expression was apparently higher in diffuse large B-cell lymphoma than in the reactive lymphoid tissues. Subsequently, immunohistochemical staining of 144 NHLs suggested that the expression of ENO1 was significantly lower in the indolent lymphomas compared with the progressive lymphomas. Further, we identified ENO1 as an independent prognostic factor, and it was significantly correlated with overall survival of NHL patients. In addition, we found that ENO1 could promote cell proliferation, regulate cell cycle associated gene and PI3K/AKT signaling pathway in NHLs. Finally, we verified that ENO1 participated in the process of lymphoma cell adhesion mediated drug resistance (CAM-DR). Adhesion to FN or HS5 cells significantly protected OCI-Ly8 and Daudi cells from cytotoxicity compared with those cultured in suspension, and these effects were attenuated when transfected with ENO1-siRNA. Based on the study, we propose that inhibition of ENO1 expression may be a novel strategy for therapy for NHLs patients, and it may be a target for drug resistance. - Highlights: • ENO1 expression is reversely correlated with clinical outcomes of patients with NHLs. • ENO1 promotes the proliferation of NHL cells. • ENO1 regulates cell adhesion mediated drug resistance.

  13. The possibility and the effects of a steam explosion in the BWR lower head on recriticality of a BWR core

    Energy Technology Data Exchange (ETDEWEB)

    Sehgal, B.R.; Dinh, T.N. [Sehgal Consult (Sweden)

    2002-12-01

    The report describes an analysis considering a BWR postulated severe accident scenario during which the late vessel automatic depressurization brings the water below the level of the bottom core plate. The subsequent lack of ECCS leads to core heat up during which the control rods melt and the melt deposits on the core plate. At that point of time in the scenario, the core fuel bundles are still intact and the Zircaloy clad oxidation is about to start. The objective of the study is to provide the conditions of reflood into the hot core due to the level swell or a slug delivered from the lower head as the control rod melt drops into the water. These conditions are employed in the neutronic analysis with the RECRIT code to determine if the core recriticality may be achieved. (au)

  14. Hyperactivity, startle reactivity and cell-proliferation deficits are resistant to chronic lithium treatment in adult Nr2e1(frc/frc) mice.

    Science.gov (United States)

    Wong, B K Y; Hossain, S M; Trinh, E; Ottmann, G A; Budaghzadeh, S; Zheng, Q Y; Simpson, E M

    2010-10-01

    The NR2E1 region on Chromosome 6q21-22 has been repeatedly linked to bipolar disorder (BP) and NR2E1 has been associated with BP, and more specifically bipolar I disorder (BPI). In addition, patient sequencing has shown an enrichment of rare candidate-regulatory variants. Interestingly, mice carrying either spontaneous (Nr2e1(frc) ) or targeted (Tlx(-) ) deletions of Nr2e1 (here collectively known as Nr2e1-null) show similar neurological and behavioral anomalies, including hypoplasia of the cerebrum, reduced neural stem cell proliferation, extreme aggression and deficits in fear conditioning; these are the traits that have been observed in some patients with BP. Thus, NR2E1 is a positional and functional candidate for a role in BP. However, no Nr2e1-null mice have been fully evaluated for behaviors used to model BP in rodents or pharmacological responses to drugs effective in treating BP symptoms. In this study we examine Nr2e1(frc/frc) mice, homozygous for the spontaneous deletion, for abnormalities in activity, learning and information processing, and cell proliferation; these are the phenotypes that are either affected in patients with BP or commonly assessed in rodent models of BP. The effect of lithium, a drug used to treat BP, was also evaluated for its ability to attenuate Nr2e1(frc/frc) behavioral and neural stem cell-proliferation phenotypes. We show for the first time that Nr2e1-null mice exhibit extreme hyperactivity in the open field as early as postnatal day 18 and in the home cage, deficits in open-field habituation and passive avoidance, and surprisingly, an absence of acoustic startle. We observed a reduction in neural stem/progenitor cell proliferation in Nr2e1(frc/frc) mice, similar to that seen in other Nr2e1-null strains. These behavioral and cell-proliferation phenotypes were resistant to chronic-adult-lithium treatment. Thus, Nr2e1(frc/frc) mice exhibit behavioral traits used to model BP in rodents, but our results do not support Nr2e1(frc

  15. Hyperactivity, startle reactivity and cell-proliferation deficits are resistant to chronic lithium treatment in adult Nr2e1frc/frc mice

    Science.gov (United States)

    Wong, Bibiana K.Y.; Hossain, Sazzad M.; Trinh, Eric; Ottmann, Glen A.; Budaghzadeh, Saeed; Zheng, Qing Y.; Simpson, Elizabeth M.

    2012-01-01

    The NR2E1 region on Chromosome 6q21-22 has been repeatedly linked to bipolar disorder (BP) and NR2E1 has been associated with BP, and more specifically bipolar I disorder (BPI). In addition, patient sequencing has revealed an enrichment of rare candidate-regulatory variants. Interestingly, mice carrying either spontaneous (Nr2e1frc) or targeted (Tlx−) deletions of Nr2e1 (here collectively known as Nr2e1-null) show similar neurological and behavioral anomalies, including: hypoplasia of the cerebrum, reduced neural stem cell proliferation, extreme aggression, and deficits in fear conditioning; traits that have been observed in some patients with BP. Thus, NR2E1 is a positional and functional candidate for a role in BP. However, no Nr2e1-null mice have been fully evaluated for behaviors used to model BP in rodents or pharmacological responses to drugs effective in treating BP symptoms. In this study we examine Nr2e1frc/frc mice, homozygous for the spontaneous deletion, for abnormalities in activity, learning and information processing, and cell proliferation; phenotypes that are either affected in patients with BP or commonly assessed in rodent models of BP. The effect of lithium, a drug used to treat BP, was also evaluated for its ability to attenuate Nr2e1frc/frc behavioral and neural stem cell proliferation phenotypes. We show for the first time that Nr2e1-null mice exhibit extreme hyperactivity in the open field as early as postnatal day 18 and in the home cage, deficits in open-field habituation and passive avoidance, and, surprisingly, an absence of acoustic startle. We observed a reduction in neural stem/progenitor cell proliferation in Nr2e1frc/frc mice, similar to that seen in other Nr2e1-null strains. These behavioral and cell-proliferation phenotypes were resistant to chronic-adult-lithium treatment. Thus, Nr2e1frc/frc mice exhibit behavioral traits used to model BP in rodents, but our results do not support Nr2e1frc/frc mice as pharmacological models for

  16. The retardation of vasculopathy induced by attenuation of insulin resistance in the corpulent JCR:LA-cp rat is reflected by decreased vascular smooth muscle cell proliferation in vivo.

    Science.gov (United States)

    Absher, P M; Schneider, D J; Baldor, L C; Russell, J C; Sobel, B E

    1999-04-01

    Proliferation in vivo of vascular smooth muscle cells occurs early in the course of atherosclerosis. Cultured smooth muscle cells (SMCs) explanted from aortas of JCR:LA-cp corpulent rats known to exhibit metabolic derangements and insulin resistance typical of type II diabetes early in life and to develop atherosclerosis later in life exhibit increased proliferation compared with SMCs from lean, normal rats. Vascular smooth muscle proliferation in vitro was found to be positively and significantly correlated with plasma insulin levels in vivo. Proliferation of aortic SMCs from JCR:LA-cp cp/cp corpulent rats cultured in vitro exhibited increased proliferation in the presence of exogenous insulin. Exercise and diet, selected as interventions designed to ameliorate the insulin resistance and hyperinsulinemia in the JCR:LA-cp cp/cp rat, effectively lowered blood insulin levels and decreased subsequent proliferation in vitro of aortic SMCs explanted from these animals. The results indicate that assessment of proliferation of vascular smooth muscle cells ex vivo may provide insight into the presence and severity of atherogenicity in association with insulin resistance in diverse species under diverse circumstances. Accordingly, with appropriate controls, it may be possible to use SMC proliferation ex vivo as a marker of the extent to which an intervention such as administration of insulin sensitizers to experimental animals and human subjects results in a change in behavior of vessel wall elements potentially indicative of amelioration of atherogenicity and detectable as judged from reduced proliferative rates of the cells ex vivo when they have been harvested from vessels exposed to a milieu in which insulin resistance has been attenuated.

  17. Y-box-binding protein-1 (YB-1) promotes cell proliferation, adhesion and drug resistance in diffuse large B-cell lymphoma.

    Science.gov (United States)

    Miao, Xiaobing; Wu, Yaxun; Wang, Yuchan; Zhu, Xinghua; Yin, Haibing; He, Yunhua; Li, Chunsun; Liu, Yushan; Lu, Xiaoyun; Chen, Yali; Shen, Rong; Xu, Xiaohong; He, Song

    2016-08-15

    YB-1 is a multifunctional protein, which has been shown to correlate with resistance to treatment of various tumor types. This study investigated the expression and biologic function of YB-1 in diffuse large B-cell lymphoma (DLBCL). Immunohistochemical analysis showed that the expression statuses of YB-1 and pYB-1(S102) were reversely correlated with the clinical outcomes of DLBCL patients. In addition, we found that YB-1 could promote the proliferation of DLBCL cells by accelerating the G1/S transition. Ectopic expression of YB-1 could markedly increase the expression of cell cycle regulators cyclin D1 and cyclin E. Furthermore, we found that adhesion of DLBCL cells to fibronectin (FN) could increase YB-1 phosphorylation at Ser102 and pYB-1(S102) nuclear translocation. In addition, overexpression of YB-1 could increase the adhesion of DLBCL cells to FN. Intriguingly, we found that YB-1 overexpression could confer drug resistance through cell-adhesion dependent and independent mechanisms in DLBCL. Silencing of YB-1 could sensitize DLBCL cells to mitoxantrone and overcome cell adhesion-mediated drug resistance (CAM-DR) phenotype in an AKT-dependent manner. Copyright © 2016 Elsevier Inc. All rights reserved.

  18. Analysis of the behavior of irradiated BWR fuel rod in storage dry conditions; Analisis del comportamiento de una barra combustible irradiada BWR en condiciones de almacenamiento en seco

    Energy Technology Data Exchange (ETDEWEB)

    Munoz, A.; Montes, D.; Ruiz-Hervias, J.; Munoz-Reja, C.

    2014-07-01

    In order to complete previous studies of creep on PWR sheath material, developed a joint experimental program by CSN, ENRESA and ENUSA about BWR (Zircaloy-2) sheath material. This program consisted in creep tests and then on the material under creep, compression testing diametral obtaining the permissible displacement of the sheath to break. (Author)

  19. Silencing CAPN2 Expression Inhibited Castration-Resistant Prostate Cancer Cells Proliferation and Invasion via AKT/mTOR Signal Pathway

    Directory of Open Access Journals (Sweden)

    Pu Li

    2017-01-01

    Full Text Available The mRNA expression of CAPN2 was upregulated in CRPC cells (DU145 and PC3 than that in non-CRPC cells. Silencing CAPN2 expression could inhibit DU145 and PC3 cells proliferation by cell cycle arrest at G1 phase. Knockdown of CPAN2 level suppressed the migration and invasion capacity of CRPC cells by reducing matrix metalloproteinase-2 (MMP-2 and MMP-9 activation, as well as repressing the phosphorylation protein expression of AKT and mTOR. In addition, we found that the expression of CAPN2 was elevated in Pca tissues than that in normal control tissues. Therefore, we showed the important roles of CAPN2 in the development and progression in CRPC cells, suggesting a new therapeutic intervention for treating castration-resistant prostate cancer patients.

  20. MECHANISMS OF CELL RESISTANCE TO CYTOMEGALOVIRUS ARE CONNECTED WITH CELL PROLIFERATION STATE AND TRANSCRIPTION ACTIVITY OF LEUKOCYTE AND IMMUNE INTERFERON GENES

    Directory of Open Access Journals (Sweden)

    T. M. Sokolova

    2007-01-01

    Full Text Available Abstract. Cytomegalovirus (CMV infection in diploid human fibroblasts (HF and levels of cell resistance to this virus were shown to be in direct correlation with high α-interferon (IFNα gene activity and induction of IFNγ gene transcription. Regulation of IFNα mRNA transcription was revealed to be positively associated with cellular DNA synthesis. At the same time, activities of IFNβ and IFNγ genes were at the constantly low level and were not induced in DNA-synthetic phase (S-phase of the cells. Levels of IFNα mRNA synthesis are quite different for G0- vs S-phase-synchronized HF110044 cell cultures: appropriate values for dividing cells (S-phase proved to be 100-fold higher than in resting state (G0. The mode of CMV infection in resting HF-cell could be considered either as acute, or a productive one. On the contrary, proliferating cells exhibited lagging viral syntheses and delayed cell death. Arrest of CMV replication may be, to some extent, comparable with latent infectious state, being associated with high production of IFNα. Both basal and induced levels of IFNα mRNA in CMV-resistant adult human skin fibroblast cells (HSF-1608 were 10-fold higher than in human embryo lung cell line (HELF-977, which is highly sensitive to CMV. Moreover, a short-time induction of IFNγ genes was observed in resistant cells, whereas no such effect was noticed in highly sensitive cells. CMV reproduction in sensitive cell lines (HELF-977 and HELF-110044 partially inhibits IFNα mRNA transcription at the later stages of infection (24 to 48 hours. Thus, cellular resistance and control of CMV infection in diploid fibroblasts are associated predominantly with high transcription of IFNα gene, and with temporal induction of IFNγ gene. We did not reveal any participation of IFNβ genes in protection of human diploid fibroblasts from CMV.

  1. An A BWR demonstration simulator for training and developing technical staff

    Energy Technology Data Exchange (ETDEWEB)

    Powers, J. [Toshiba America Nuclear Energy, Charlotte, North Carolina (United States); Yonezawa, H.; Aoyagi, Y.; Kataoka, K., E-mail: jim.powers@toshiba.com [Toshiba Corporation, Kawasaki, Kanagawa (Japan)

    2015-09-15

    The US-Advanced Boiling Water Reactor (A BWR), certified by the US NRC, is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (STP3-4) Combined License Application (Cola). Nuclear Innovation North America (Nina) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. Toshiba has developed a Demonstration Simulator of the A BWR control room that provides a realistic experience for training and education on BWR principles and operations fundamentals. The Demonstration Simulator is located in the Toshiba America Nuclear Energy (Tane) office in Charlotte, North Carolina and is composed of standard office computer equipment set up in a specific arrangement that is representative of the layout of an A BWR control room. The Demonstration Simulator is not intended for licensed operator training, but can provide a framework for encouraging entry level technically oriented nuclear workers to enter the operations field; strengthening the linkage between university energy field curricula and real-life application of theory; and, improving understanding of integrated plant operations for developing station technical staff. This paper describes the A BWR Demonstration Simulator and its applications for training and educating future nuclear workers. (Author)

  2. RNAi-mediated knockdown of FANCF suppresses cell proliferation, migration, invasion, and drug resistance potential of breast cancer cells

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, L.; Li, N.; Yu, J.K.; Tang, H.T.; Li, Y.L.; He, M.; Yu, Z.J.; Bai, X.F. [Department of Pharmacology, School of Pharmacy, China Medical University, Heping Ward, Shenyang City, Liaoning (China); Zheng, Z.H.; Wang, E.H. [Institute of Pathology and Pathophysiology, China Medical University, Heping Ward, Shenyang City, Liaoning (China); Wei, M.J. [Department of Pharmacology, School of Pharmacy, China Medical University, Heping Ward, Shenyang City, Liaoning (China)

    2013-12-12

    Fanconi anemia complementation group F protein (FANCF) is a key factor, which maintains the function of FA/BRCA, a DNA damage response pathway. However, the functional role of FANCF in breast cancer has not been elucidated. We performed a specific FANCF-shRNA knockdown of endogenous FANCF in vitro. Cell viability was measured with a CCK-8 assay. DNA damage was assessed with an alkaline comet assay. Apoptosis, cell cycle, and drug accumulation were measured by flow cytometry. The expression levels of protein were determined by Western blot using specific antibodies. Based on these results, we used cell migration and invasion assays to demonstrate a crucial role for FANCF in those processes. FANCF shRNA effectively inhibited expression of FANCF. We found that proliferation of FANCF knockdown breast cancer cells (MCF-7 and MDA-MB-435S) was significantly inhibited, with cell cycle arrest in the S phase, induction of apoptosis, and DNA fragmentation. Inhibition of FANCF also resulted in decreased cell migration and invasion. In addition, FANCF knockdown enhanced sensitivity to doxorubicin in breast cancer cells. These results suggest that FANCF may be a potential target for molecular, therapeutic intervention in breast cancer.

  3. RNAi-mediated knockdown of FANCF suppresses cell proliferation, migration, invasion, and drug resistance potential of breast cancer cells

    Directory of Open Access Journals (Sweden)

    L. Zhao

    2014-01-01

    Full Text Available Fanconi anemia complementation group F protein (FANCF is a key factor, which maintains the function of FA/BRCA, a DNA damage response pathway. However, the functional role of FANCF in breast cancer has not been elucidated. We performed a specific FANCF-shRNA knockdown of endogenous FANCF in vitro. Cell viability was measured with a CCK-8 assay. DNA damage was assessed with an alkaline comet assay. Apoptosis, cell cycle, and drug accumulation were measured by flow cytometry. The expression levels of protein were determined by Western blot using specific antibodies. Based on these results, we used cell migration and invasion assays to demonstrate a crucial role for FANCF in those processes. FANCF shRNA effectively inhibited expression of FANCF. We found that proliferation of FANCF knockdown breast cancer cells (MCF-7 and MDA-MB-435S was significantly inhibited, with cell cycle arrest in the S phase, induction of apoptosis, and DNA fragmentation. Inhibition of FANCF also resulted in decreased cell migration and invasion. In addition, FANCF knockdown enhanced sensitivity to doxorubicin in breast cancer cells. These results suggest that FANCF may be a potential target for molecular, therapeutic intervention in breast cancer.

  4. Mistletoe (Viscum album) extract targets Axl to suppress cell proliferation and overcome cisplatin- and erlotinib-resistance in non-small cell lung cancer cells.

    Science.gov (United States)

    Kim, Soyoung; Kim, Kyung-Chan; Lee, ChuHee

    2017-12-01

    found to be induced and reduced by VAE treatment, respectively. Taken together, our data provide that VAE targets Axl to suppress cell proliferation and to circumvent cisplatin- and erlotinib-resistance in NSCLC cells. Copyright © 2017 Elsevier GmbH. All rights reserved.

  5. Environmental enrichment induces behavioral recovery and enhanced hippocampal cell proliferation in an antidepressant-resistant animal model for PTSD.

    Directory of Open Access Journals (Sweden)

    Hendrikus Hendriksen

    Full Text Available BACKGROUND: Post traumatic stress disorder (PTSD can be considered the result of a failure to recover after a traumatic experience. Here we studied possible protective and therapeutic aspects of environmental enrichment (with and without a running wheel in Sprague Dawley rats exposed to an inescapable foot shock procedure (IFS. METHODOLOGY/PRINCIPAL FINDINGS: IFS induced long-lasting contextual and non-contextual anxiety, modeling some aspects of PTSD. Even 10 weeks after IFS the rats showed reduced locomotion in an open field. The antidepressants imipramine and escitalopram did not improve anxiogenic behavior following IFS. Also the histone deacetylase (HDAC inhibitor sodium butyrate did not alleviate the IFS induced immobility. While environmental enrichment (EE starting two weeks before IFS did not protect the animals from the behavioral effects of the shocks, exposure to EE either immediately after the shock or one week later induced complete recovery three weeks after IFS. In the next set of experiments a running wheel was added to the EE to enable voluntary exercise (EE/VE. This also led to reduced anxiety. Importantly, this behavioral recovery was not due to a loss of memory for the traumatic experience. The behavioral recovery correlated with an increase in cell proliferation in hippocampus, a decrease in the tissue levels of noradrenalin and increased turnover of 5-HT in prefrontal cortex and hippocampus. CONCLUSIONS/SIGNIFICANCE: This animal study shows the importance of (physical exercise in the treatment of psychiatric diseases, including post-traumatic stress disorder and points out the possible role of EE in studying the mechanism of recovery from anxiety disorders.

  6. Development of controllers with high reliability for BWR plant

    Energy Technology Data Exchange (ETDEWEB)

    Asami, K.; Iida, H.; Eki, Y.; Hirose, M.; Ito, T. (Hitachi Ltd., Tokyo (Japan))

    1980-09-01

    Owing to the problems in system operation due to recent energy situation and the increase of nuclear power generation, high reliability and high rate of operation are required for nuclear power stations. In Hitachi Ltd., in order to meet these needs, efforts were exerted positively to make the controllers for BWR plants reliable. In this paper, three representative examples among them are described. The digital type flow rate control system for feed water recirculation ''D-FRC'' and the digital, electronic hydraulic type turbine control system ''D-EHC'' were developed by applying digital control technology and multiplication techniques, aiming at the high reliability of the most important control systems for nuclear power stations. The analog trip module enabled to attain high reliability and to reduce radiation exposure by improving preventive maintainability. Owing to the recent energy situation, the needs of safety and the high rate of operation in nuclear power stations have increased, but the attainment of high reliability in hardwares only has its limit, and the high reliability of systems is required. The measures for improving the reliability, the constitution of the D-FRC and its watching and diagnosing functions, the achievement of high quality control and the outline of the D-EHC and the analog trip module are described.

  7. BWR Anticipated Transients Without Scram Leading to Instability

    Energy Technology Data Exchange (ETDEWEB)

    Cheng L. Y.; Baek J.; Cuadra, A.; Aronson, A.; Diamond, D.; Yarsky, P.

    2013-11-10

    Anticipated transients without scram (ATWS) in aboiling water reactor (BWR) were simulated in order to understand reactor response and determine the effectiveness of automatic and operator actions to mitigate this beyond-design-basis accident. The events of interest herein are initiated by a turbine trip when the reactor is operating in the expanded operating domainMELLLA+ [maximum extended load line limit plus]. In these events the reactor may initially be at up to 120% of the original licensed thermal power (OLTP) and at flow rates as low as 80% of rated.For these (and similar) ATWS events the concern isthat when the reactor power decreases in response to a dual recirculation pump trip, the core will become unstable and large amplitude oscillations will begin. The occurrence of these power oscillations, if left unmitigated, may result in fuel damage, and the amplitude of the poweroscillations may hamper the effectiveness of the injection of dissolved neutron absorber through the standby liquid control system (SLCS).

  8. A New Methodology for Early Anomaly Detection of BWR Instabilities

    Energy Technology Data Exchange (ETDEWEB)

    Ivanov, K. N.

    2005-11-27

    The objective of the performed research is to develop an early anomaly detection methodology so as to enhance safety, availability, and operational flexibility of Boiling Water Reactor (BWR) nuclear power plants. The technical approach relies on suppression of potential power oscillations in BWRs by detecting small anomalies at an early stage and taking appropriate prognostic actions based on an anticipated operation schedule. The research utilizes a model of coupled (two-phase) thermal-hydraulic and neutron flux dynamics, which is used as a generator of time series data for anomaly detection at an early stage. The model captures critical nonlinear features of coupled thermal-hydraulic and nuclear reactor dynamics and (slow time-scale) evolution of the anomalies as non-stationary parameters. The time series data derived from this nonlinear non-stationary model serves as the source of information for generating the symbolic dynamics for characterization of model parameter changes that quantitatively represent small anomalies. The major focus of the presented research activity was on developing and qualifying algorithms of pattern recognition for power instability based on anomaly detection from time series data, which later can be used to formulate real-time decision and control algorithms for suppression of power oscillations for a variety of anticipated operating conditions. The research being performed in the framework of this project is essential to make significant improvement in the capability of thermal instability analyses for enhancing safety, availability, and operational flexibility of currently operating and next generation BWRs.

  9. Reactor core stability monitoring method for BWR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kanemoto, Shigeru; Ebata, Shigeo.

    1992-09-01

    In an operation for a BWR type reactor, reactor power is usually increased or decreased by controlling both of control rods and reactor core flow rate. Under a certain condition, the reactor core is made unstable by the coupling of nuclear and thermohydrodynamic characteristics in the reactor. Therefore, the reactor power and the reactor core flow rate are changed within a range predetermined by a design calculation. However, if reactor core stability can be always measured and monitored, it is useful for safe operation, as well as an existent operation range can be extended to enable more effective operation. That is, autoregressive a coefficient is determined successively on real time based on fluctuation components of neutron flux signals. Based on the result, an amplification ratio, as a typical measure of the reactor core stability, is determined on a real time. A time constant of the successive calculation for the autoregressive coefficient can be made variable by the amplification ratio. Then, the amplification ratio is estimated at a constant accuracy. With such procedures, the reactor core stability can be monitored successively in an ON-line manner at a high accuracy, thereby enabling to improve the operation performance. (I.S.).

  10. Beta and gamma dose calculations for PWR and BWR containments

    Energy Technology Data Exchange (ETDEWEB)

    King, D.B.

    1989-07-01

    Analyses of gamma and beta dose in selected regions in PWR and BWR containment buildings have been performed for a range of fission product releases from selected severe accidents. The objective of this study was to determine the radiation dose that safety-related equipment could experience during the selected severe accident sequences. The resulting dose calculations demonstrate the extent to which design basis accident qualified equipment could also be qualified for the severe accident environments. Surry was chosen as the representative PWR plant while Peach Bottom was selected to represent BWRs. Battelle Columbus Laboratory performed the source term release analyses. The AB epsilon scenario (an intermediate to large LOCA with failure to recover onsite or offsite electrical power) was selected as the base case Surry accident, and the AE scenario (a large break LOCA with one initiating event and a combination of failures in two emergency cooling systems) was selected as the base case Peach Bottom accident. Radionuclide release was bounded for both scenarios by including spray operation and arrested sequences as variations of the base scenarios. Sandia National Laboratories used the source terms to calculate dose to selected containment regions. Scenarios with sprays operational resulted in a total dose comparable to that (2.20 /times/ 10/sup 8/ rads) used in current equipment qualification testing. The base case scenarios resulted in some calculated doses roughly an order of magnitude above the current 2.20 /times/ 10/sup 8/ rad equipment qualification test region. 8 refs., 23 figs., 12 tabs.

  11. GOTHIC MODEL OF BWR SECONDARY CONTAINMENT DRAWDOWN ANALYSES

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, P.N.

    2004-10-06

    This article introduces a GOTHIC version 7.1 model of the Secondary Containment Reactor Building Post LOCA drawdown analysis for a BWR. GOTHIC is an EPRI sponsored thermal hydraulic code. This analysis is required by the Utility to demonstrate an ability to restore and maintain the Secondary Containment Reactor Building negative pressure condition. The technical and regulatory issues associated with this modeling are presented. The analysis includes the affect of wind, elevation and thermal impacts on pressure conditions. The model includes a multiple volume representation which includes the spent fuel pool. In addition, heat sources and sinks are modeled as one dimensional heat conductors. The leakage into the building is modeled to include both laminar as well as turbulent behavior as established by actual plant test data. The GOTHIC code provides components to model heat exchangers used to provide fuel pool cooling as well as area cooling via air coolers. The results of the evaluation are used to demonstrate the time that the Reactor Building is at a pressure that exceeds external conditions. This time period is established with the GOTHIC model based on the worst case pressure conditions on the building. For this time period the Utility must assume the primary containment leakage goes directly to the environment. Once the building pressure is restored below outside conditions the release to the environment can be credited as a filtered release.

  12. BWR/5 Pressure-Suppression Pool Response during an SBO

    Directory of Open Access Journals (Sweden)

    Javier Ortiz-Villafuerte

    2013-01-01

    Full Text Available RELAP/SCDAPSIM Mod 3.4 has been used to simulate a station blackout occurring at a BWR/5 power station. Further, a simplified model of a wet well and dry well has been added to the NSSS model to study the response of the primary containment during the evolution of this accident. The initial event leading to severe accident was considered to be a LOOP with simultaneous scram. The results show that RCIC alone can keep the core fully covered, but even in this case about 30% of the original liquid water inventory in the PSP is vaporized. During the SBO, without RCIC, this inventory is reduced about 5% more within six hours. Further, a significant pressure rise occurs in containment at about the time when a sharp increase of heat generation occurs in RPV due to cladding oxidation. Failure temperature of fuel clad is also reached at this point. As the accident progresses, conditions for containment venting can be reached in about nine hours, although there still exists considerable margin before reaching containment design pressure. Detailed information of accident progress in reactor vessel and containment is presented and discussed.

  13. Effect of Control Blade History, and Axial Coolant Density and Burnup Profiles on BWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, William BJ J [ORNL

    2016-01-01

    A technical basis for peak reactivity boiling water reactor (BWR) burnup credit (BUC) methods was recently generated, and the technical basis for extended BWR BUC is now being developed. In this paper, a number of effects related to extended BWR BUC are analyzed, including three major operational effects in BWRs: the coolant density axial distribution, the use of control blades during operation, and the axial burnup profile. Specifically, uniform axial moderator density profiles are analyzed and compared to previous results and an additional temporal fidelity study combing moderator density profiles for three different fuel assemblies is presented. Realistic control blade histories and cask criticality results are compared to previously generated constructed control blade histories. Finally, a preliminary study of the axial burnup profile is provided.

  14. Key Parameters for Operator Diagnosis of BWR Plant Condition during a Severe Accident

    Energy Technology Data Exchange (ETDEWEB)

    Clayton, Dwight A [ORNL; Poore III, Willis P [ORNL

    2015-01-01

    The objective of this research is to examine the key information needed from nuclear power plant instrumentation to guide severe accident management and mitigation for boiling water reactor (BWR) designs (specifically, a BWR/4-Mark I), estimate environmental conditions that the instrumentation will experience during a severe accident, and identify potential gaps in existing instrumentation that may require further research and development. This report notes the key parameters that instrumentation needs to measure to help operators respond to severe accidents. A follow-up report will assess severe accident environmental conditions as estimated by severe accident simulation model analysis for a specific US BWR/4-Mark I plant for those instrumentation systems considered most important for accident management purposes.

  15. Report on the BWR owners group radiation protection/ALARA Committee

    Energy Technology Data Exchange (ETDEWEB)

    Aldrich, L.R. [Commonwealth Edison Co., Downers Grove, IL (United States)

    1995-03-01

    Radiation protection programs at U.S. boiling water reactor (BWR) stations have evolved during the 1980s and early 1990s from a regulatory adherence-based endeavor to a proactive, risk-based radiation protection and prevention mission. The objectives are no longer to merely monitor and document exposure to radiation and radioactive materials. The focus of the current programs is the optimization of radiation protection of occupational workers consistent with the purpose of producing cost-effective electric power. The newly revised 10 CFR 20 defines the term ALARA (as low as reasonably achievable) to take into account the state of technology, the economics of improvements in relation to the state of the technology, and the benefits to the public health and safety. The BWR Owners Group (BWROG) initially formed the Radiation Protection/ALARA Committee in January 1990 to evaluate methods of reducing occupational radiation exposure during refueling outages. Currently, twenty U.S. BWR owner/operators (representing 36 of the operational 37 domestic BWR units), as well as three foreign BWR operators (associate members), have broadened the scope to promote information exchange between BWR radiation protection professionals and develop good practices which will affect optimization of their radiation protection programs. In search of excellence and the challenge of becoming {open_quotes}World Class{close_quotes} performers in radiation protection, the BWROG Radiation Protection/ALARA Committee has recently accepted a role in assisting the member utilities in improving radiation protection performance in a cost-effective manner. This paper will summarize the recent activities of this Committee undertaken to execute their role of exchanging information in pursuit of optimizing the improvement of their collective radiation protection performance.

  16. Coretran/Vipre assembly critical power assessment against Nupec BWR experiments

    Energy Technology Data Exchange (ETDEWEB)

    Aounallah, Y. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    2001-07-01

    This study has been performed, in the framework of the STARS project, to assess CORETRAN-01/VIPRE-02 code capability to predict critical heat flux conditions for BWR fuel assemblies. The assessment is based on comparisons of the code results with the NUPEC steady-state critical power measurements on full-scale assemblies tested under a range of flow conditions. Two assembly types were considered, the standard BWR 8 x 8 and the so-called ''high-burnup'' assembly, similar to GE-10. Code modelling options that have a significant impact on the results have been identified, along with code limitations. (author)

  17. Design of a redundant meteorological station for a BWR reactor; Diseno de una estacion meteorologica redundante para un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, R.; Celis del Angel, L.; Bucio, F.; Rivero, T.; Palacios, J. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: ramses@nuclear.inin.mx

    2008-07-01

    In this work the design of a meteorological station for a reactor type BWR is proposed. Two independent channels of data acquisition that allow him to have a bigger readiness is exposed. It is incorporate sensors without mobile parts to measure speed, wind direction and pluvial precipitation. It also counts, with sensors of global solar radiation, net radiation, barometric pressure, relative humidity and ambient temperature; with them they are possible to be calculated, moreover, other variables as temperature differential, dew point and atmospheric stability. The sensors are placed on a tower to different heights and send their information (each second) to a local registration system, the one which in turn, it remits the data to the monitoring office so that a computer is linked with the system, display and management the information in real time and automatic way. The redundant structure allows that in the event of maintenance the data acquisition is not interrupted, even if the information is transferred to another place. In all the station sections it is used protocols of standard communication to allow that a great quantity of devices can be connected without major problem. The above-mentioned would allow to the operators in the control room to have reliable information during the whole time of the reactor operation. (Author)

  18. Identification of the reduced order models of a BWR reactor; Identificacion de modelos de orden reducido de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez S, A. [UNAM, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: augusto@correo.unam.mx

    2004-07-01

    The present work has as objective to analyze the relative stability of a BWR type reactor. It is analyzed that so adaptive it turns out to identify the parameters of a model of reduced order so that this it reproduces a condition of given uncertainty. This will take of a real fact happened in the La Salle plant under certain operation conditions of power and flow of coolant. The parametric identification is carried out by means of an algorithm of recursive least square and an Output Error model (Output Error), measuring the output power of the reactor when the instability is present, and considering that it is produced by a change in the reactivity of the system in the same way that a sign of type step. Also it is carried out an analytic comparison of the relative stability, analyzing two types of answers: the original answer of the uncertainty of the reactor vs. the obtained response identifying the parameters of the model of reduced order, reaching the conclusion that it is very viable to adapt a model of reduced order to study the stability of a reactor, under the only condition to consider that the dynamics of the reactivity is of step type. (Author)

  19. pH-dependent effects of 2,4-dinitrophenol (DNP) on proliferation, endocytosis, fine structure and DNP resistance in Tetrahymena

    DEFF Research Database (Denmark)

    Nilsson, Jytte R.

    1995-01-01

    Biologi, 2,4-dinitrophenol, pH-dependence, fine structure, cell proliferation, phagocytosis, Tetrahymena pyriformis......Biologi, 2,4-dinitrophenol, pH-dependence, fine structure, cell proliferation, phagocytosis, Tetrahymena pyriformis...

  20. Resistant carbohydrates stimulate cell proliferation and crypt fission in wild-type mice and in the Apc(Min/+) mouse model of intestinal cancer, association with enhanced polyp development.

    Science.gov (United States)

    Mandir, Nikki; Englyst, Hans; Goodlad, Robert A

    2008-10-01

    Fermentation of carbohydrates in the colon can stimulate cell proliferation and could thus be a cancer risk. The effects of resistant carbohydrates, i.e. those not digested and absorbed in the small intestine, on cell proliferation, crypt fission and polyp development were investigated in wild-type and adenomatous polyposis coli multiple intestinal neoplasia (Apc(Min/+)) mice. Fifteen 4-week-old female wild-type and fifteen Apc(Min/+) mice were used for each group and fed a chow diet, a semi-synthetic diet or the semi-synthetic supplemented with wheat bran or an apple pomace preparation, both high in resistant carbohydrates, for 8 weeks. Tissue from all mice was used to measure cell proliferation and crypt fission and tissue from the Apc(Min/+) mice was scored for polyp number and tumour burden. There were slight reductions in intestinal mass in the mice fed the semi-synthetic diets and this was increased by the inclusion of resistant carbohydrates. The Apc(Min/+) mice had elevated cell proliferation and crypt fission in the distal small intestine and colon and these were increased by the resistant carbohydrates. Bran or apple pomace significantly increased polyp number in the proximal third of the small intestine. Apple pulp more than doubled polyp number throughout the small bowel (99.2 (SEM 11.1) v. 40.0 (SEM 8.2), Ptypes of resistant carbohydrates increased polyp number and tumour burden and this was associated with elevated epithelial cell proliferation and crypt fission.

  1. Applied methods for mitigation of damage by stress corrosion in BWR type reactors; Metodos aplicados para la mitigacion del dano por corrosion bajo esfuerzo en reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez C, R.; Diaz S, A.; Gachuz M, M.; Arganis J, C. [Instituto Nacional de Investigaciones Nucleares, Gerencia de Ciencia de Materiales, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    1998-07-01

    The Boiling Water nuclear Reactors (BWR) have presented stress corrosion problems, mainly in components and pipes of the primary system, provoking negative impacts in the performance of energy generator plants, as well as the increasing in the radiation exposure to personnel involucred. This problem has caused development of research programs, which are guided to find solution alternatives for the phenomena control. Among results of greater relevance the control for the reactor water chemistry stands out particularly in the impurities concentration and oxidation of radiolysis products; as well as the supervision in the materials selection and the stresses levels reduction. The present work presents the methods which can be applied to diminish the problems of stress corrosion in BWR reactors. (Author)

  2. THERMAL EVALUATION OF THE USE OF BWR MOX SNF IN THE WASTE PACKAGE DESIGN (SCPB: N/A)

    Energy Technology Data Exchange (ETDEWEB)

    H. Wang

    1997-01-23

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) as specified in the Waste Package Implementation Plan (pp. 4-8,4-11,4-24, 5-1, and 5-13; Ref. 5.10) and Waste Package Plan (pp. 3-15,3-17, and 3-24; Ref. 5.9). The design data request addressed herein is: (1) Characterize the conceptual 40 BWR and 24 BWR Multi-Purpose Canister (MPC) Waste Package (WP) design to show that the design is feasible for use in the MGDS environment when loaded with BWR MOX SNF. (2) Characterize the conceptual 44 BWR and 24 BWR Uncanistered Fuel (UCF) Waste Package (WP) design to show that the design is feasible for use in the MGDS environment when loaded with BWR MOX SNF. The purpose of this analysis is to respond to a concern that the long-term disposal thermal issues for the WP Design, if used with SNF designed for a MOX fuel cycle, do not preclude WP compatibility with the MGDS. The objective of this analysis is to provide thermal parameter information for the conceptual WP design with disposal container which is loaded with BWR MOX SNF under nominal MGDS repository conditions. The results are intended to show that the design has a reasonable chance to meet the MGDS design requirements for normal MGDS operation, and to provide the required guidance to determining the major design issues for future design efforts, and to show that the BWR MOX SNF loaded WP performance is similar to an WP loaded with commercial BWR SNF.

  3. Propagation of cracks by stress corrosion in conditions of BWR type reactor; Propagacion de grietas por corrosion bajo esfuerzo en condiciones de reactor de agua en ebullicion (BWR)

    Energy Technology Data Exchange (ETDEWEB)

    Merino C, F.J. [ININ, 52045 Estado de Mexico (Mexico); Fuentes C, P. [ITT, Metepec, Estado de Mexico (Mexico)]. E-mail: fjmc@nuclear.inin.mx

    2004-07-01

    In this work, the obtained results when applying the Hydrogen Chemistry to a test tube type Compact Tension (CT), built in austenitic stainless steel 304l, simulating the conditions to those that it operates a Boiling Water Reactor (BWR), temperature 288 C and pressure of 8 MPa are presented. With the application of this water chemistry, seeks to be proven the diminution of the crack propagation speed. (Author)

  4. Bisphenol A activates EGFR and ERK promoting proliferation, tumor spheroid formation and resistance to EGFR pathway inhibition in estrogen receptor-negative inflammatory breast cancer cells.

    Science.gov (United States)

    Sauer, Scott J; Tarpley, Michael; Shah, Imran; Save, Akshay V; Lyerly, H Kim; Patierno, Steven R; Williams, Kevin P; Devi, Gayathri R

    2017-03-01

    Emerging evidence from epidemiological studies suggests a link between environmental chemical exposure and progression of aggressive breast cancer subtypes. Of all clinically distinct types of breast cancers, the most lethal phenotypic variant is inflammatory breast cancer (IBC). Overexpression of epidermal growth factor receptors (EGFR/HER2) along with estrogen receptor (ER) negativity is common in IBC tumor cells, which instead of a solid mass present as rapidly proliferating diffuse tumor cell clusters. Our previous studies have demonstrated a role of an adaptive response of increased antioxidants in acquired resistance to EGFR-targeting drugs in IBC. Environmental chemicals are known to induce oxidative stress resulting in perturbations in signal transduction pathways. It is therefore of interest to identify chemicals that can potentiate EGFR mitogenic effects in IBC. Herein, we assessed in ER-negative IBC cells a subset of chemicals from the EPA ToxCast set for their effect on EGFR activation and in multiple cancer phenotypic assays. We demonstrated that endocrine-disrupting chemicals such as bisphenol A (BPA) and 2,2-bis(p-hydroxyphenyl)-1,1,1-trichloroethane can increase EGFR/ERK signaling. BPA also caused a corresponding increase in expression of SOD1 and anti-apoptotic Bcl-2, key markers of antioxidant and anti-apoptotic processes. BPA potentiated clonogenic growth and tumor spheroid formation in vitro, reflecting IBC-specific pathological characteristics. Furthermore, we identified that BPA was able to attenuate the inhibitory effect of an EGFR targeted drug in a longer-term anchorage-independent growth assay. These findings provide a potential mechanistic basis for environmental chemicals such as BPA in potentiating a hyperproliferative and death-resistant phenotype in cancer cells by activating mitogenic pathways to which the tumor cells are addicted for survival. © The Author 2017. Published by Oxford University Press. All rights reserved. For

  5. Blueberry intake alters skeletal muscle and adipose tissue peroxisome proliferator-activated receptor activity and reduces insulin resistance in obese rats.

    Science.gov (United States)

    Seymour, E Mitchell; Tanone, Ignasia I; Urcuyo-Llanes, Daniel E; Lewis, Sarah K; Kirakosyan, Ara; Kondoleon, Michael G; Kaufman, Peter B; Bolling, Steven F

    2011-12-01

    Metabolic syndrome can precede the development of type 2 diabetes and cardiovascular disease and includes phenotypes such as obesity, systemic inflammation, insulin resistance, and hyperlipidemia. A recent epidemiological study indicated that blueberry intake reduced cardiovascular mortality in humans, but the possible genetic mechanisms of this effect are unknown. Blueberries are a rich source of anthocyanins, and anthocyanins can alter the activity of peroxisome proliferator-activated receptors (PPARs), which affect energy substrate metabolism. The effect of blueberry intake was assessed in obesity-prone rats. Zucker Fatty and Zucker Lean rats were fed a higher-fat diet (45% of kcal) or a lower-fat diet (10% of kcal) containing 2% (wt/wt) freeze-dried whole highbush blueberry powder or added sugars to match macronutrient and calorie content. In Zucker Fatty rats fed a high-fat diet, the addition of blueberry reduced triglycerides, fasting insulin, homeostasis model index of insulin resistance, and glucose area under the curve. Blueberry intake also reduced abdominal fat mass, increased adipose and skeletal muscle PPAR activity, and affected PPAR transcripts involved in fat oxidation and glucose uptake/oxidation. In Zucker Fatty rats fed a low-fat diet, the addition of blueberry also significantly reduced liver weight, body weight, and total fat mass. Finally, Zucker Lean rats fed blueberry had higher body weight and reduced triglycerides, but all other measures were unaffected. In conclusion, whole blueberry intake reduced phenotypes of metabolic syndrome in obesity-prone rats and affected PPAR gene transcripts in adipose and muscle tissue involved in fat and glucose metabolism.

  6. Analysis CFD for the hydrogen transport in the primary containment of a BWR; Analisis CFD para el transporte de hidrogeno en la contencion primaria de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez P, D. A.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Edificio 9, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Gomez T, A. M., E-mail: guerreroazteca_69@hotmail.com [ININ, Departamento de Sistemas Nucleares, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    This study presents a qualitative and quantitative comparison among the CFD GASFLOW and OpenFOAM codes which are related with the phenomenon of hydrogen transport and other gases in the primary containment of a Boiling Water Reactor (BWR). GASFLOW is a commercial license code that is well validated and that was developed in Germany for the analysis of the gases transport in containments of nuclear reactors. On the other hand, OpenFOAM is an open source code that offers several evaluation solvers for different types of phenomena; in this case, the solver reacting-Foam is used to analyze the hydrogen transport inside the primary containment of the BWR. The results that offer the solver reacting-Foam of OpenFOAM are evaluated in the hydrogen transport calculation and the results are compared with those of the program of commercial license GASFLOW to see if is viable the use of the open source code in the case of the hydrogen transport in the primary containment of a BWR. Of the obtained results so much quantitative as qualitative some differences were identified between both codes, the differences (with a percentage of maximum error of 4%) in the quantitative results are small and they are considered acceptable for this analysis type, also, these differences are attributed mainly to the used transport models, considering that OpenFOAM uses a homogeneous model and GASFLOW uses a heterogeneous model. (Author)

  7. Synthesis of Potent and Selective Inhibitors of Aldo-Keto Reductase 1B10 and Their Efficacy against Proliferation, Metastasis, and Cisplatin Resistance of Lung Cancer Cells.

    Science.gov (United States)

    Endo, Satoshi; Xia, Shuang; Suyama, Miho; Morikawa, Yoshifumi; Oguri, Hiroaki; Hu, Dawei; Ao, Yoshinori; Takahara, Satoyuki; Horino, Yoshikazu; Hayakawa, Yoshihiro; Watanabe, Yurie; Gouda, Hiroaki; Hara, Akira; Kuwata, Kazuo; Toyooka, Naoki; Matsunaga, Toshiyuki; Ikari, Akira

    2017-10-26

    Aldo-keto reductase 1B10 (AKR1B10) is overexpressed in several extraintestinal cancers, particularly in non-small-cell lung cancer, where AKR1B10 is a potential diagnostic marker and therapeutic target. Selective AKR1B10 inhibitors are required because compounds should not inhibit the highly related aldose reductase that is involved in monosaccharide and prostaglandin metabolism. Currently, 7-hydroxy-2-(4-methoxyphenylimino)-2H-chromene-3-carboxylic acid benzylamide (HMPC) is known to be the most potent competitive inhibitor of AKR1B10, but it is nonselective. In this study, derivatives of HMPC were synthesized by removing the 4-methoxyphenylimino moiety and replacing the benzylamide with phenylpropylamide. Among them, 4c and 4e showed higher AKR1B10 inhibitory potency (IC50 4.2 and 3.5 nM, respectively) and selectivity than HMPC. The treatments with the two compounds significantly suppressed not only migration, proliferation, and metastasis of lung cancer A549 cells but also metastatic and invasive potentials of cisplatin-resistant A549 cells.

  8. Candesartan cilexetil prevents diet-induced insulin resistance via peroxisome proliferator-activated receptor-γ activation in an obese rat model.

    Science.gov (United States)

    Yan, Wen-Hua; Pan, Chang-Yu; Dou, Jing-Tao; Meng, Jun-Hua; Wang, Bao-An; Mu, Yi-Ming

    2016-07-01

    Angiotensin II type 1 receptor (AT1R) blockers (ARBs) have been shown to reduce the incidence of type 2 diabetes mellitus; however, the underlying molecular mechanism is unknown. Peroxisome proliferator-activated receptor γ (PPARγ) is the central regulator of insulin and glucose metabolism, which improves insulin sensitivity. Whether candesartan cilexetil, as a prodrug of the AT1R blocker candesartan, has PPARγ-activating properties remains to be elucidated. The aim of the present study was to investigate the effects of oral administration of candesartan cilexetil on glucose tolerance and the actions of PPARγ on liver and adipose tissue in the insulin-resistant obese rat induced by high-fat diet. Animals treated with candesartan cilexetil showed an improved glucose tolerance after oral glucose challenge. Whole-body insulin sensitivity was evaluated using the hyperinsulinemic-euglycemic clamp technique. During high-fat feeding in high-fat diet (HF) rats, the glucose infusion rate (GIR) was 52.3% lower than that in normal chow (NC) rats. However, the GIR was significantly enhanced following candesartan cilexetil treatment. Angiotensin II receptor antagonism also resulted in significant increases in PPARγ protein expression in adipose and liver tissue. These results indicate that PPARγ activation by candesartan cilexetil may provide novel therapeutic options in the treatment of patients with metabolic syndrome.

  9. Peroxisome Proliferator-Activated Receptor α Activates Human Multidrug Resistance Transporter 3/ATP-Binding Cassette Protein Subfamily B4 Transcription and Increases Rat Biliary Phosphatidylcholine Secretion

    Science.gov (United States)

    Ghonem, Nisanne S.; Ananthanarayanan, Meenakshisundaram; Soroka, Carol J.; Boyer, James L.

    2014-01-01

    Multidrug resistance transporter 3/ATP-binding cassette protein subfamily B4 (MDR3/ABCB4) is a critical determinant of biliary phosphatidylcholine (PC) secretion. Clinically, mutations and partial deficiencies in MDR3 result in cholestatic liver injury. Thus, MDR3 is a potential therapeutic target for cholestatic liver disease. Fenofibrate is a peroxisome proliferator-activated receptor (PPAR) α ligand that has antiinflammatory actions and regulates bile acid detoxification. Here we examined the mechanism by which fenofibrate regulates MDR3 gene expression. Fenofibrate significantly up-regulated MDR3 messenger RNA (mRNA) and protein expression in primary cultured human hepatocytes, and stimulated MDR3 promoter activity in HepG2 cells. In silico analysis of 5′-upstream region of human MDR3 gene revealed a number of PPARα response elements (PPRE). Electrophoretic mobility shift (EMSA) and chromatin immunoprecipitation (ChIP) assays demonstrated specific binding of PPARα to the human MDR3 promoter. Targeted mutagenesis of three novel PPREs reduced inducibility of the MDR3 promoter by fenofibrate. In collagen sandwich cultured rat hepatocytes, treatment with fenofibrate increased secretion of fluorescent PC into bile canaliculi. Conclusion Fenofibrate transactivates MDR3 gene transcription by way of the binding of PPARα to three novel and functionally critical PPREs in the MDR3 promoter. Fenofibrate treatment further stimulates biliary phosphatidylcholine secretion in rat hepatocytes, thereby providing a functional correlate. We have established a molecular mechanism that may contribute to the beneficial use of fenofibrate therapy in human cholestatic liver disease. PMID:24122873

  10. ADP-ribosylation factor 1 (ARF1) takes part in cell proliferation and cell adhesion-mediated drug resistance (CAM-DR).

    Science.gov (United States)

    Xu, Xiaohong; Wang, Qiru; He, Yunhua; Ding, Linlin; Zhong, Fei; Ou, Yangyu; Shen, Yaodong; Liu, Hong; He, Song

    2017-05-01

    Cell adhesion-mediated drug resistance (CAM-DR) remains the primary obstacle in human multiple myeloma (MM) therapy. In this study, we aimed at investigating the expression and biologic function of ARF1 in MM. We determined that ARF1 expression was positively correlated with cell proliferation and knockdown of ARF1 contributed to CAM-DR. The enhancement in the adhesion of MM cells to fibronectin (FN) or the bone marrow stroma cell line HS-5 cells translated to an increased CAM-DR phenotype. Importantly, we showed that this CAM-DR phenotype was correlated with the phosphorylation of Akt and ERK in MM cells. Moreover, we sought to determine whether ARF1 could interact with p27 in RPMI8226 cells. Knockdown of ARF1 also significantly decreased pT157-p27 protein expression in RPMI8226 cells. Our research shows ARF1 may reverse CAM-DR by regulating phosphorylation of p27 at T157 in MM. Taken together, our data shed new light on the molecular mechanism of CAM-DR in MM, and targeting ARF1 may be a novel therapeutic approach for improving the effectiveness of chemotherapy in MM.

  11. Signal analysis of acoustic and flow-induced vibrations of BWR main steam line

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa-Paredes, G., E-mail: gepe@xanum.uam.mx [División de Ciencias Básicas e Ingeniería, Universidad Autónoma Metropolitana-Iztapalapa, México, D.F. 09340 (Mexico); Prieto-Guerrero, A. [División de Ciencias Básicas e Ingeniería, Universidad Autónoma Metropolitana-Iztapalapa, México, D.F. 09340 (Mexico); Núñez-Carrera, A. [Comisión Nacional de Seguridad Nuclear y Salvaguardias, Doctor Barragán 779, Col. Narvarte, México, D.F. 03020 (Mexico); Vázquez-Rodríguez, A. [División de Ciencias Básicas e Ingeniería, Universidad Autónoma Metropolitana-Iztapalapa, México, D.F. 09340 (Mexico); Centeno-Pérez, J. [Instituto Politécnico Nacional, Escuela Superior de Física y Matemáticas Unidad Profesional “Adolfo López Mateos”, Av. IPN, s/n, México, D.F. 07738 (Mexico); Espinosa-Martínez, E.-G. [Departamento de Sistemas Energéticos, Universidad Nacional Autónoma de México, México, D.F. 04510 (Mexico); and others

    2016-05-15

    Highlights: • Acoustic and flow-induced vibrations of BWR are analyzed. • BWR performance after extended power uprate is considered. • Effect of acoustic side branches (ASB) is analyzed. • The ASB represents a reduction in the acoustic loads to the steam dryer. • Methodology developed for simultaneous analyzing the signals in the MSL. - Abstract: The aim of this work is the signal analysis of acoustic waves due to phenomenon known as singing in Safety Relief Valves (SRV) of the main steam lines (MSL) in a typical BWR5. The acoustic resonance in SRV standpipes and fluctuating pressure is propagated from SRV to the dryer through the MSL. The signals are analyzed with a novel method based on the Multivariate Empirical Mode Decomposition (M-EMD). The M-EMD algorithm has the potential to find common oscillatory modes (IMF) within multivariate data. Based on this fact, we implement the M-EMD technique to find the oscillatory mode in BWR considering the measurements obtained collected by the strain gauges located around the MSL. These IMF, analyzed simultaneously in time, allow obtaining an estimation of the effects of the multiple-SRV in the MSL. Two scenarios are analyzed: the first is the signal obtained before the installation of the acoustic dampers (ASB), and the second, the signal obtained after installation. The results show the effectiveness of the ASB to damp the strong resonances when the steam flow increases, which represents an important reduction in the acoustic loads to the steam dryer.

  12. Development and Testing of CTF to Support Modeling of BWR Operating Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Salko, Robert K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wysocki, Aaron [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Collins, Benjamin S. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Godfrey, Andrew T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gosdin, Chris [Pennsylvania State Univ., University Park, PA (United States); Avramova, Maria [North Carolina State Univ., Raleigh, NC (United States)

    2016-01-29

    This milestone supports developing and assessing COBRA-TF (CTF) for the modeling of boiling water reactors (BWRs). This is achieved in three stages. First, a new preprocessor utility that is capable of handling BWR-specic design elements (e.g., channel boxes and large water rods) is developed. A previous milestone (L3:PHI.CTF.P12.01) led to the development of this preprocessor capability for single assembly models. This current milestone expands this utility so that it is applicable to multi-assembly BWR models that can be modeled in either serial or parallel. The second stage involves making necessary modications to CTF so that it can execute these new models. Specically, this means implementing an outer-iteration loop, specic to BWR models, that equalizes the pressure loss over all assemblies in the core (which are not connected due to the channel boxes) by adjusting inlet mass ow rate. A third stage involves assessing the standard convergence metrics that are used by CTF to determine when a simulation is steady-state. The nal stage has resulted in the implementation of new metrics in the code that give a better indication of how steady the solution is at convergence. This report summarizes these eorts and provides a demonstration of CTF's BWR-modeling capabilities. CASL-U-2016-1030-000

  13. Flex concept for US-A BWR extended loss of AC power events

    Energy Technology Data Exchange (ETDEWEB)

    Powers, J. [Toshiba America Nuclear Energy, Charlotte, North Carolina (United States); Aoyagi, Y.; Kataoka, K. [Toshiba Corporation, Kawasaki, Kanagawa (Japan); Thomas, S.; Mookhoek, B., E-mail: jim.powers@toshiba.com [Nuclear Innovation North America, Lake Jackson, Texas (United States)

    2015-09-15

    The US-Advanced Boiling Water Reactor (US-A BWR), certified by the US NRC, is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (Stp 3 and 4) Combined License Application (Cola) and incorporates numerous design and technology enhancements for improved safety performance. Nuclear Innovation North America (NINA) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. The Stp 3 and 4 project has finished the US NRC technical review of the Cola, and the final safety evaluation report (FSER) is scheduled to be issued by the US NRC in 2015. Following the accident at the Fukushima Dai-ichi plant, the US-A BWR was reviewed for Beyond Design Basis Event (BDBE) safety using industry and regulatory guidance for US NRC Order EA-12-049 Order Modifying Licenses with Regard to Requirements for Mitigation of Beyond Design Basis External Events (BDBEE). By virtue of the design approach, the US-A BWR is capable of providing an indefinite coping period for a station blackout. The use of installed systems with extended coping times is a significant advantage of the US-A BWR compared to most of the plants currently operating in the U.S. In addition, the Stp 3 and 4 design incorporates enhancements consistent with the current US industry Diverse and Flexible Coping Strategies (Flex) initiative. The final technical topic requiring review by the US NRC Advisory Committee on Reactor Safeguards was the Flex Integrated Plan submitted by NINA, and this review was successfully completed. This paper summarizes the progress of the US-A BWR in licensing the Flex Integrated Plan for the project, and describes the technology and features of the US-A BWR design that contribute to safety post-Fukushima. It also provides an informational comparison of the design capabilities of the US-A BWR for extreme external events, and relates these capabilities to re

  14. Assessment of the Prony's method for BWR stability analysis

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz-Villafuerte, Javier, E-mail: javier.ortiz@inin.gob.m [Gerencia de Ciencias Aplicadas, Instituto Nacional de Investigaciones Nucleares, Carr. Mexico-Toluca S/N, La Marquesa, Ocoyoacac, Edo. Mexico 52750 (Mexico); Castillo-Duran, Rogelio, E-mail: rogelio.castillo@inin.gob.m [Gerencia de Ciencias Aplicadas, Instituto Nacional de Investigaciones Nucleares, Carr. Mexico-Toluca S/N, La Marquesa, Ocoyoacac, Edo. Mexico 52750 (Mexico); Palacios-Hernandez, Javier C., E-mail: javier.palacios@inin.gob.m [Gerencia de Ciencias Aplicadas, Instituto Nacional de Investigaciones Nucleares, Carr. Mexico-Toluca S/N, La Marquesa, Ocoyoacac, Edo. Mexico 52750 (Mexico)

    2011-05-15

    Highlights: This paper describes a method to determine the degree of stability of a BWR. Performance comparison between Prony's and common AR techniques is presented. Benchmark data and actual BWR transient data are used for comparison. DR and f results are presented and discussed. The Prony's method is shown to be a robust technique for BWR stability. - Abstract: It is known that Boiling Water Reactors are susceptible to present power oscillations in regions of high power and low coolant flow, in the power-flow operational map. It is possible to fall in one of such instability regions during reactor startup, since both power and coolant flow are being increased but not proportionally. One other possibility for falling into those areas is the occurrence of a trip of recirculation pumps. Stability monitoring in such cases can be difficult, because the amount or quality of power signal data required for calculation of the stability key parameters may not be enough to provide reliable results in an adequate time range. In this work, the Prony's Method is presented as one complementary alternative to determine the degree of stability of a BWR, through time series data. This analysis method can provide information about decay ratio and oscillation frequency from power signals obtained during transient events. However, so far not many applications in Boiling Water Reactors operation have been reported and supported to establish the scope of using such analysis for actual transient events. This work presents first a comparison of decay ratio and frequency oscillation results obtained by Prony's method and those results obtained by the participants of the Forsmark 1 and 2 Boiling Water Reactor Stability Benchmark using diverse techniques. Then, a comparison of decay ratio and frequency oscillation results is performed for four real BWR transient event data, using Prony's method and two other techniques based on an autoregressive modeling. The four

  15. Axial profiles of burned and fraction of holes for calculations of criticality with credit for BWR fuel burning; Perfiles axiales de quemado y fraccion de huecos para calculos de criticidad con credito al quemado para combustible BWR

    Energy Technology Data Exchange (ETDEWEB)

    Casado Sanchez, C.; Rubio Oviedo, P.

    2014-07-01

    This paper presents a method to define surround profiles of burning and fraction of holes suited for use in applications of credit to burning of BWR fuel from results obtained with the module STARBUCS of SCALE. (Author)

  16. Mobile crud and transportation of radioactivity in BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hermansson, H-P. [Studsvik Nuclear AB, Nykoping (Sweden); LTU, Div. of Chemical Engineering, Lulea (Sweden); Hagg, J. [Ringhals AB, Varobacka (Sweden)

    2010-07-01

    Mobile crud is here referred to as a generic term for all types of particles that occur in the reactor water in BWRs and that are able to carry radioactivity. Previous results in this on-going series of studies in Swedish BWRs suggest that there are particles of different origins and function. A share may come from fuel crud and others may come from detachment, precipitation and dissolution processes in different parts of the BWR primary system, as well as from other system parts, such as the turbine/condenser. In addition, crud particles in this sense may come from purely mechanical processes such as degradation of graphite containing parts of the control rod drives. Therefore, the overall aim was to evaluate which particles are responsible for the transportation and distribution of radioactivity and also to clarify the chemical conditions under which they are formed. Furthermore the aim was to draw conclusions about how the chemistry would be like in order to avoid or at least minimize the formation of radioactivity distributing particles. A specific objective has also been to look into the importance of particle size for spreading of radioactivity in the primary system. Different types of crud particles are likely to have different characteristics in terms of function associated with transportation of radioactivity. The fuel crud is radioactive from the source and other types of crud can via surface processes, co-precipitation and other chemical and mechanical processes potentially affect the distribution of radioactivity in the primary system. In order to predict how operating parameters (e.g. stable, full power operation and scram) and chemical parameters (NWC/HWC/Zn, etc.) will affect the activity build-up on the system surfaces, it is important to know how the different types of crud are affected by these and related parameters. Fuel crud fixed on cladding ring samples, as well as mobile crud from the reactor water captured on filters, were examined by

  17. Proliferation Resistant Nuclear Reactor Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Gray, L W; Moody, K J; Bradley, K S; Lorenzana, H E

    2011-02-18

    Global appetite for fission power is projected to grow dramatically this century, and for good reason. Despite considerable research to identify new sources of energy, fission remains the most plentiful and practical alternative to fossil fuels. The environmental challenges of fossil fuel have made the fission power option increasingly attractive, particularly as we are forced to rely on reserves in ecologically fragile or politically unstable corners of the globe. Caught between a globally eroding fossil fuel reserve as well as the uncertainty and considerable costs in the development of fusion power, most of the world will most likely come to rely on fission power for at least the remainder of the 21st century. Despite inevitable growth, fission power faces enduring challenges in sustainability and security. One of fission power's greatest hurdles to universal acceptance is the risk of potential misuse for nefarious purposes of fissionable byproducts in spent fuel, such as plutonium. With this issue in mind, we have discussed intrinsic concepts in this report that are motivated by the premise that the utility, desirability, and applicability of nuclear materials can be reduced. In a general sense, the intrinsic solutions aim to reduce or eliminate the quantity of existing weapons usable material; avoid production of new weapons-usable material through enrichment, breeding, extraction; or employ engineering solutions to make the fuel cycle less useful or more difficult for producing weapons-usable material. By their nature, these schemes require modifications to existing fuel cycles. As such, the concomitants of these modifications require engagement from the nuclear reactor and fuel-design community to fully assess their effects. Unfortunately, active pursuit of any scheme that could further complicate the spread of domestic nuclear power will probably be understandably unpopular. Nevertheless, the nonproliferation and counterterrorism issues are paramount, and we posit that the exploration, development, and implementation of intrinsic mechanisms such as discussed here are part of a balanced approach aimed at preventing the misuse of nuclear material for nuclear-energy applications.

  18. Exosome-like vesicles released from lipid-induced insulin-resistant muscles modulate gene expression and proliferation of beta recipient cells in mice.

    Science.gov (United States)

    Jalabert, Audrey; Vial, Guillaume; Guay, Claudiane; Wiklander, Oscar P B; Nordin, Joel Z; Aswad, Hala; Forterre, Alexis; Meugnier, Emmanuelle; Pesenti, Sandra; Regazzi, Romano; Danty-Berger, Emmanuelle; Ducreux, Sylvie; Vidal, Hubert; El-Andaloussi, Samir; Rieusset, Jennifer; Rome, Sophie

    2016-05-01

    The crosstalk between skeletal muscle (SkM) and beta cells plays a role in diabetes aetiology. In this study, we have investigated whether SkM-released exosome-like vesicles (ELVs) can be taken up by pancreatic beta cells and can deliver functional cargoes. Mice were fed for 16 weeks with standard chow diet (SCD) or with standard diet enriched with 20% palmitate (HPD) and ELVs were purified from quadriceps muscle. Fluorescent ELVs from HPD or SCD quadriceps were injected i.v. or intramuscularly (i.m.) into mice to determine their biodistributions. Micro (mi)RNA quantification in ELVs was determined using quantitative real-time RT-PCR (qRT-PCR)-based TaqMan low-density arrays. Microarray analyses were performed to determine whether standard diet ELVs (SD-ELVs) and high palmitate diet ELVs (HPD-ELVs) induced specific transcriptional signatures in MIN6B1 cells. In vivo, muscle ELVs were taken up by pancreas, 24 h post-injection. In vitro, both SD-ELVs and HPD-ELVs transferred proteins and miRNAs to MIN6B1 cells and modulated gene expressions whereas only HPD-ELVs induced proliferation of MIN6B1 cells and isolated islets. Bioinformatic analyses suggested that transferred HPD-ELV miRNAs may participate in these effects. To validate this, we demonstrated that miR-16, which is overexpressed in HPD-ELVs, was transferred to MIN6B1 cells and regulated Ptch1, involved in pancreas development. In vivo, islets from HPD mice showed increased size and altered expression of genes involved in development, including Ptch1, suggesting that the effect of palm oil on islet size in vivo was reproduced in vitro by treating beta cells with HPD-ELVs. Our data suggest that muscle ELVs might have an endocrine effect and could participate in adaptations in beta cell mass during insulin resistance.

  19. Characterization of welding of AISI 304l stainless steel similar to the core encircling of a BWR reactor; Caracterizacion de soldaduras de acero inoxidable AISI 304L similares a las de la envolvente del nucleo de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Gachuz M, M.E.; Palacios P, F.; Robles P, E.F. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2003-07-01

    Plates of austenitic stainless steel AISI 304l of 0.0381 m thickness were welded by means of the SMAW process according to that recommended in the Section 9 of the ASME Code, so that it was reproduced the welding process used to assemble the encircling of the core of a BWR/5 reactor similar to that of the Laguna Verde Nucleo electric plant, there being generated the necessary documentation for the qualification of the one welding procedure and of the welder. They were characterized so much the one base metal, as the welding cord by means of metallographic techniques, scanning electron microscopy, X-ray diffraction, mechanical essays and fracture mechanics. From the obtained results it highlights the presence of an area affected by the heat of up to 1.5 mm of wide and a value of fracture tenacity (J{sub IC}) to ambient temperature for the base metal of 528 KJ/m{sup 2}, which is diminished by the presence of the welding and by the increment in the temperature of the one essay. Also it was carried out an fractographic analysis of the fracture zone generated by the tenacity essays, what evidence a ductile fracture. The experimental values of resistance and tenacity are important for the study of the structural integrity of the encircling one of the core. (Author)

  20. Current and anticipated use of thermal-hydraulic codes for BWR transient and accident analyses in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Arai, Kenji; Ebata, Shigeo [Toshiba Corp., Yokohama (Japan)

    1997-07-01

    This paper summarizes the current and anticipated use of the thermal-hydraulic and neutronic codes for the BWR transient and accident analyses in Japan. The codes may be categorized into the licensing codes and the best estimate codes for the BWR transient and accident analyses. Most of the licensing codes have been originally developed by General Electric. Some codes have been updated based on the technical knowledge obtained in the thermal hydraulic study in Japan, and according to the BWR design changes. The best estimates codes have been used to support the licensing calculations and to obtain the phenomenological understanding of the thermal hydraulic phenomena during a BWR transient or accident. The best estimate codes can be also applied to a design study for a next generation BWR to which the current licensing model may not be directly applied. In order to rationalize the margin included in the current BWR design and develop a next generation reactor with appropriate design margin, it will be required to improve the accuracy of the thermal-hydraulic and neutronic model. In addition, regarding the current best estimate codes, the improvement in the user interface and the numerics will be needed.

  1. Design and axial optimization of nuclear fuel for BWR reactors; Diseno y optimizacion axial de combustible nuclear para reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Garcia V, M.A

    2006-07-01

    In the present thesis, the modifications made to the axial optimization system based on Tabu Search (BT) for the axial design of BWR fuel type are presented, developed previously in the Nuclear Engineering Group of the UNAM Engineering Faculty. With the modifications what is mainly looked is to consider the particular characteristics of the mechanical design of the GE12 fuel type, used at the moment in the Laguna Verde Nucleo electric Central (CNLV) and that it considers the fuel bars of partial longitude. The information obtained in this thesis will allow to plan nuclear fuel reloads with the best conditions to operate in a certain cycle guaranteeing a better yield and use in the fuel burnt, additionally people in charge in the reload planning will be favored with the changes carried out to the system for the design and axial optimization of nuclear fuel, which facilitate their handling and it reduces their execution time. This thesis this developed in five chapters that are understood in the following way in general: Chapter 1: It approaches the basic concepts of the nuclear energy, it describes the physical and chemical composition of the atoms as well as that of the uranium isotopes, the handling of the uranium isotope by means of the nuclear fission until arriving to the operation of the nuclear reactors. Chapter 2: The nuclear fuel cycle is described, the methods for its extraction, its conversion and its enrichment to arrive to the stages of the nuclear fuel management used in the reactors are described. Beginning by the radial design, the axial design and the core design of the nuclear reactor related with the fuel assemblies design. Chapter 3: the optimization methods of nuclear fuel previously used are exposed among those that are: the genetic algorithms method, the search methods based on heuristic rules and the application of the tabu search method, which was used for the development of this thesis. Chapter 4: In this part the used methodology to the

  2. Coolant Density and Control Blade History Effects in Extended BWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Ade, Brian J [ORNL; Marshall, William BJ J [ORNL; Bowman, Stephen M [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL; Martinez-Gonzalez, Jesus S [ORNL

    2015-01-01

    Oak Ridge National Laboratory and the US Nuclear Regulatory Commission have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase investigates the applicability of peak reactivity methods currently used for spent fuel pools to spent fuel storage and transportation casks and the validation of reactivity (keff) calculations and predicted spent fuel compositions. The second phase focuses on extending BUC beyond peak reactivity. This paper documents work performed to date investigating some aspects of extended BUC. (The technical basis for application of peak reactivity methods to BWR fuel in storage and transportation systems is presented in a companion paper.) Two reactor operating parameters are being evaluated to establish an adequate basis for extended BWR BUC: (1) the effect of axial void profile and (2) the effect of control blade utilization during operation. A detailed analysis of core simulator data for one cycle of a modern operating BWR plant was performed to determine the range of void profiles and the variability of the profile experienced during irradiation. Although a single cycle does not provide complete data, the data obtained are sufficient to determine the primary effects and to identify conservative modeling approaches. These data were used in a study of the effect of axial void profile. The first stage of the study was determination of the necessary moderator density temporal fidelity in depletion modeling. After the required temporal fidelity was established, multiple void profiles were used to examine the effect on cask reactivity. The results of these studies are being used to develop recommendations for conservatively modeling the void profile effects for BWR depletion calculations. The second operational parameter studied was control blade history. Control blades are inserted in

  3. Extended Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

    Energy Technology Data Exchange (ETDEWEB)

    Ade, Brian J [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bowman, Stephen M [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauld, Ian C [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ilas, Germina [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Martinez, J. S. [Univ. Politecnica de Madrid (Spain). Dept. of Nuclear Engineering

    2015-01-01

    [Full Text] Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase investigates the applicability of peak reactivity methods currently used for spent fuel pools to spent fuel storage and transportation casks and the validation of reactivity (keff) calculations and depleted fuel compositions. The second phase focuses on extending BUC beyond peak reactivity. This paper documents work performed to date, investigating some aspects of extended BUC, and it also describes the plan to complete the evaluations. The technical basis for application of peak reactivity methods to BWR fuel in storage and transportation systems is presented in a companion paper. Two reactor operating parameters are being evaluated to establish an adequate basis for extended BWR BUC, including investigation of the axial void profile effect and the effect of control blade utilization during operation. A detailed analysis of core simulator data for one cycle of an operating BWR plant was performed to determine the range of void profiles and the variability of the profile experienced during irradiation. While a single cycle does not provide complete data, the data obtained are sufficient to use to determine the primary effects and identify conservative modeling approaches. Using data resulting from a single cycle, the axial void profile is studied by first determining the temporal fidelity necessary in depletion modeling, and then using multiple void profiles to examine the effect of the void profile on cask reactivity. The results of these studies are being used to develop recommendations for conservatively modeling the void profile effects for BWR depletion calculations. The second operational parameter studied is control blade exposure. Control blades

  4. BWR Spent Nuclear Fuel Interfacial Bonding Efficiency Study

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jiang, Hao [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-04-30

    The objective of this project is to perform a systematic study of spent nuclear fuel (SNF, also known as “used nuclear fuel” [UNF]) integrity under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL) in August 2013. Under Nuclear Regulatory Commission (NRC) sponsorship, ORNL completed four benchmark tests, four static tests, and twelve dynamic or cycle tests on H. B. Robinson (HBR) high burn-up (HBU) fuel. The clad of the HBR fuels was made of Zircaloy-4. Testing was continued in fiscal year (FY) 2014 using Department of Energy (DOE) funds. Additional CIRFT testing was conducted on three HBR rods; two specimens failed, and one specimen was tested to over 2.23 × 107 cycles without failing. The data analysis on all the HBR SNF rods demonstrated that it is necessary to characterize the fatigue life of the SNF rods in terms of (1) the curvature amplitude and (2) the maximum absolute of curvature extremes. The maximum extremes are significant because they signify the maximum tensile stress for the outer fiber of the bending rod. CIRFT testing has also addressed a large variation in hydrogen content on the HBR rods. While the load amplitude is the dominant factor that controls the fatigue life of bending rods, the hydrogen content also has an important effect on the lifetime attained at each load range tested. In FY 15, eleven SNF rod segments from the Limerick BWR were tested using the ORNL CIRFT equipment; one test under static conditions and ten tests under dynamic loading conditions. Under static unidirectional loading, a moment of 85 N·m was obtained at a maximum curvature of 4.0 m-1. The specimen did not show any sign of failure during three repeated loading cycles to a similar maximum curvature. Ten cyclic tests were conducted with amplitudes varying from 15.2 to 7.1 N·m. Failure was observed in nine of

  5. Analysis of results of AZTRAN and AZKIND codes for a BWR; Analisis de resultados de los codigos AZTRAN y AZKIND para un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Bastida O, G. E.; Vallejo Q, J. A.; Galicia A, J.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico); Xolocostli M, J. V.; Rodriguez H, A.; Gomez T, A. M., E-mail: gbo729@yahoo.com.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2016-09-15

    This paper presents an analysis of results obtained from simulations performed with the neutron transport code AZTRAN and the kinetic code of neutron diffusion AZKIND, based on comparisons with models corresponding to a typical BWR, in order to verify the behavior and reliability of the values obtained with said code for its current development. For this, simulations of different geometries were made using validated nuclear codes, such as CASMO, MCNP5 and Serpent. The results obtained are considered adequate since they are comparable with those obtained and reported with other codes, based mainly on the neutron multiplication factor and the power distribution of the same. (Author)

  6. Radiation field control at the latest BWR plants -- design principle, operational experience and future subjects

    Energy Technology Data Exchange (ETDEWEB)

    Uchida, Shunsuke [Energy Research Lab., Ibaraki (Japan); Ohsumi, Katsumi; Takashima, Yoshie [Hitachi Works, Ibaraki (Japan)

    1995-03-01

    Improvements of operational procedures to control water chemistry, e.g., nickel/iron control, as well as application of hardware improvements for reducing radioactive corrosion products resulted in an extremely low occupational exposure of less than 0.5 man.Sv/yr without any serious impact on the radwaste system, for BWR plants involved in the Japanese Improvement and Standardization Program. Recently, {sup 60}C radioactively in the reactor water has been increasing due to less crud fixation on the two smooth surfaces of new type high performance fuels and to the pH drop caused by chromium oxide anions released from stainless steel structures and pipings. This increase must be limited by changes in water chemistry, e.g., applications of modified nickel/iron ratio control and weak alkali control. Controlled water chemistry to optimize three points, the plant radiation level and integrities of fuel and structural materials, is the primary future subject for BWR water chemistry.

  7. Effect of Control Blade History, and Axial Coolant Density and Burnup Profiles on BWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Ade, Brian J [ORNL; Marshall, William BJ J [ORNL; Martinez-Gonzalez, Jesus S [ORNL

    2015-05-01

    Oak Ridge National Laboratory (ORNL) and the US Nuclear Regulatory Commission (NRC) have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling water reactor (BWR) fuel in storage and transportation systems (often referred to as casks) and spent fuel pools (SFPs). This work is divided into two main phases. The first phase investigated the applicability of peak reactivity methods currently used in SFPs to transportation and storage casks and the validation of reactivity calculations and spent fuel compositions within these methods. The second phase focuses on extending BUC beyond peak reactivity. This paper documents the analysis of the effects of control blade insertion history, and moderator density and burnup axial profiles for extended BWR BUC.

  8. Impact of Reactor Operating Parameters on Cask Reactivity in BWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Germina [ORNL; Betzler, Benjamin R [ORNL; Ade, Brian J [ORNL

    2017-01-01

    This paper discusses the effect of reactor operating parameters used in fuel depletion calculations on spent fuel cask reactivity, with relevance for boiling-water reactor (BWR) burnup credit (BUC) applications. Assessments that used generic BWR fuel assembly and spent fuel cask configurations are presented. The considered operating parameters, which were independently varied in the depletion simulations for the assembly, included fuel temperature, bypass water density, specific power, and operating history. Different operating history scenarios were considered for the assembly depletion to determine the effect of relative power distribution during the irradiation cycles, as well as the downtime between cycles. Depletion, decay, and criticality simulations were performed using computer codes and associated nuclear data within the SCALE code system. Results quantifying the dependence of cask reactivity on the assembly depletion parameters are presented herein.

  9. Development of a scatter search optimization algorithm for BWR fuel lattice design

    Energy Technology Data Exchange (ETDEWEB)

    Francois, J.L.; Martin-del-Campo, C. [Mexico Univ. Nacional Autonoma, Facultad de Ingenieria (Mexico); Morales, L.B.; Palomera, M.A. [Mexico Univ. Nacional Autonoma, Instituto de Investigaciones en Matematicas Aplicadas y Sistemas, D.F. (Mexico)

    2005-07-01

    A basic Scatter Search (SS) method, applied to the optimization of radial enrichment and gadolinia distributions for BWR fuel lattices, is presented in this paper. Scatter search is considered as an evolutionary algorithm that constructs solutions by combining others. The goal of this methodology is to enable the implementation of solution procedures that can derive new solutions from combined elements. The main mechanism for combining solutions is such that a new solution is created from the strategic combination of two other solutions to explore the solutions' space. Results show that the Scatter Search method is an efficient optimization algorithm applied to the BWR design and optimization problem. Its main features are based on the use of heuristic rules since the beginning of the process, which allows directing the optimization process to the solution, and to use the diversity mechanism in the combination operator, which allows covering the search space in an efficient way. (authors)

  10. Decontamination as a precursor to decommissioning. Status report Task 2: process evaluation. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Divine, J.R.; Woodruff, E.M.; McPartland, S.A.; Zima, G.E.

    1983-05-01

    As part of the US Nuclear Regulatory Commission's program to reduce occupational exposure and waste volumes, the Pacific Northwest Laboratory is studying decontamination as a precursor to decommissioning. Eleven processes or solvents were examined for their behavior in decontaminating BWR carbon steel samples. The solvents included NS-1, a proprietary solvent of Dow Chemical Corporation, designed for BWR use, and AP-Citrox, a well-known, two-step process designed for PWR stainless steel; it was used to provide a reference for later comparison to other systems and processes. The decontamination factors observed in the tests performed in a small laboratory scale recirculating loop ranged from about 1 (no effect) to 222 (about 99.6% of the initial activity removed. Coordinated corrosion measurements were made using twelve chemical solvents and eight metal alloys found in a range of reactor types.

  11. Passive BWR integral LOCA testing at the Karlstein test facility INKA

    Energy Technology Data Exchange (ETDEWEB)

    Drescher, Robert [AREVA GmbH, Erlangen (Germany); Wagner, Thomas [AREVA GmbH, Karlstein am Main (Germany); Leyer, Stephan [TH Univ. of Applied Sciences, Deggendorf (Germany)

    2014-07-01

    KERENA is an innovative AREVA GmbH boiling water reactor (BWR) with passive safety systems (Generation III+). In order to verify the functionality of the reactor design an experimental validation programme was executed. Therefore the INKA (Integral Teststand Karlstein) test facility was designed and erected. It is a mockup of the BWR containment with integrated pressure suppression system. In March 2013 the first integral test - Main Steam Line Break (MSLB) - was executed. The main target was to demonstrate the ability of the passive systems to ensure core coverage, decay heat removal and to maintain the containment within defined limits. The results of the test showed that the passive safety systems are capable to bring the plant to stable conditions meeting all required safety targets with sufficient margins. (orig.)

  12. Development of membrane moisture separator for BWR off-gas system

    Energy Technology Data Exchange (ETDEWEB)

    Ogata, H.; Kawamura, S. [Tokyo Electric Power Co., Inc. (Japan); Kumasaka, M. [Hitachi Ltd., Ibaraki (Japan); Nishikubo, M. [Toshiba Corp., Yokohama (Japan)

    2001-07-01

    In BWR plant off-gas treatment systems, dehumidifiers are used to maintain noble gas adsorption efficiency in the first half of the charcoal hold-up units. From the perspective of simplifying and reducing the cost of such a dehumidification system, Japanese BWR utilities and plant fabricators have been developing a dehumidification system employing moisture separation membrane of the type already proven in fields such as medical instrumentation and precision measuring apparatus. The first part of this development involved laboratory testing to simulate the conditions found in an actual off-gas system, the results of which demonstrated satisfactory results in terms of moisture separation capability and membrane durability, and suggested favorable prospects for application in actual off-gas systems. Further, in-plant testing to verify moisture separation capability and membrane durability in the presence of actual gases is currently underway, with results so far suggesting that the system is capable of obtaining good moisture separation capability. (author)

  13. Study on reactor vessel replacement (RVR) for 1100 MW class BWR plants in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Mizutani, J.; Kawamura, S. [Tokyo Electric Power Co., Inc. (Japan); Aoki, M. [Hitachi Ltd., Ibaraki (Japan); Mori, T. [Toshiba Corp., Yokihama (Japan)

    2001-07-01

    Plant Life Management (PLM) is being studied in Japan, and reactor vessel replacement (RVR) is being considered as one option. Since reactor internals, except for reusable parts, and the reactor pressure vessel (RPV) are replaced, the RVR provides an effective technology for extending the service life of nuclear power plants substantially. At ICONE 7, we reported on the technical viability of the RVR for BWR4-type 800 MWe class plants. This time, we rationalized the RVR method through a study for BWR5-type 1100 MWe class plants to reduce the RVR duration and evaluated the technical viability and the economic efficiency of the method. In addition, we discuss how to dispose of the RPV to complete a scenario of the process from the RVR to its final disposal. (author)

  14. Probabilistic analysis of PWR and BWR fuel rod performance using the code CASINO-SLEUTH

    Energy Technology Data Exchange (ETDEWEB)

    Bull, A.J.

    1987-05-01

    This paper presents a brief description of the Monte Carlo and response surface techniques used in the code, and a probabilistic analysis of fuel rod performance in PWR and BWR applications. The analysis shows that fission gas release predictions are very sensitive to changes in certain of the code's inputs, identifies the most dominant input parameters and compares their effects in the two cases.

  15. Current understanding on the neutron irradiation embrittlement of BWR reactor pressure vessel steels in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Asano, K.; Nishiyama, T. [TEPCO (Japan); Soneda, N.; Dohi, K.; Nishida, K.; Nomoto, A. [CRIEPI (Japan); Ohta, T. [Japan Atomic Power Co. (Japan); Ishimaru, Y. [Chugoku EPCO (Japan); Yoneda, H. [Hokuriku EPCO (Japan); Lida, J. [Tohoku EPCO (Japan); Yuya, H. [Chubu EPCO (Japan)

    2011-07-01

    Neutron irradiation embrittlement of reactor pressure vessel (RPV) steels has been of concern primarily for the pressurized water reactors (PWRs). After long operation experiences, we are now becoming aware of the situation that the neutron irradiation embrittlement is also of concern for some of the boiling water reactors (BWRs) particularly with Cu-containing RPV steels. The surveillance data of Cu-containing BWR RPV steels show relatively larger shift in ductile-to-brittle transition temperature of fracture toughness than predicted by the embrittlement correlation method developed in late eighties and early nineties. Accurate evaluation of the amount of embrittlement is now very important for long-term operation of BWRs. In this paper, we will describe the neutron irradiation embrittlement of BWR RPVs in Japan. Some of the materials that show relatively large transition temperature shifts are investigated to understand the causes of embrittlement using state-of-the-art microstructural characterization techniques. Furthermore, some archive materials of such RPVs are irradiated in a material testing reactor with high neutron flux to understand the effect of flux on transition temperature shifts and corresponding microstructural changes. Microstructural evolution under irradiation, solute clustering in particular could explain the differences in transition temperature shift of the analyzed specimens. Larger BWR RPVs, which have larger water gaps, receive less neutron irradiation and harmful impurities in steels such as copper are well controlled since 1980 so irradiation embrittlement in BWR vessels can now be considered a concern only in old and small plants. All the new information obtained through these activities was considered in the development of new embrittlement correlation that is now adopted in JEAC 4201- 2007 of Japan Electric Association

  16. Analysis of BWR/Mark III drywell failure during degraded core accidents

    Energy Technology Data Exchange (ETDEWEB)

    Yang, J.W.

    1983-01-01

    The potential for a hydrogen detonation due to the accumulation of a large amount of hydrogen in the drywell region of a BWR Mark III containment is analyzed. Loss of integrity of the drywell wall causes a complete bypass of the suppression pool and leads to pressurization of the containment building. However, the predicted peak containment pressure does not exceed the estimates of containment failure pressure.

  17. In vivo and ex vivo regulation of breast cancer resistant protein (Bcrp) by peroxisome proliferator-activated receptor alpha (Pparα) at the blood-brain barrier.

    Science.gov (United States)

    Hoque, Md Tozammel; Shah, Arpit; More, Vijay; Miller, David S; Bendayan, Reina

    2015-12-01

    Breast cancer resistance protein (Bcrp/Abcg2) localized at the blood-brain barrier (BBB) limits permeability into the brain of many xenobiotics, including pharmacological agents. Peroxisome proliferator-activated receptor α (Pparα), a ligand-activated transcription factor, primarily involved in lipid metabolism, has been shown to regulate the functional expression of Bcrp in human cerebral microvascular endothelial cells (hCMEC/D3). The aim of this study was to investigate ex vivo and in vivo, the regulation of Bcrp by Pparα in an intact BBB. Ex vivo quantitative real-time PCR and immunoblot analyses showed significant up-regulation of Abcg2/Bcrp mRNA and protein levels in CD-1 mouse brain capillaries incubated with clofibrate, a Pparα ligand. Fluorescence-based transport assays in CD-1 and C57BL/6 brain capillaries showed that exposure to clofibrate significantly increased Bcrp transport activity. This increase was not observed in capillaries isolated from Pparα knockout mice. In vivo, we found: i) significant Bcrp protein up-regulation in clofibrate-dosed CD-1 and C57BL/6 capillary lysates, but no effect in Pparα knockout capillary lysates, and ii) significantly increased Bcrp transport activity in capillaries isolated from clofibrate-treated mice. These results demonstrate an increase in Bcrp functional expression by Pparα in brain capillaries, and suggest that Pparα is another nuclear receptor that can contribute to the regulation of membrane efflux transporters and drug permeability at the BBB. We propose the involvement of the following pathways in clofibrate-mediated induction of the drug transporter Abcg2/Bcrp mRNA, protein expression and function by the nuclear receptor Pparα, in mouse brain capillary endothelial cells. Upon activation with clofibrate (Pparα, ligand), Pparα complex translocates from the cytoplasm into the nucleus and further recruits coactivators and transcription machinery which induce the transcription of Abcg2 gene and

  18. Identification and assessment of BWR in-vessel severe accident mitigation strategies

    Energy Technology Data Exchange (ETDEWEB)

    Hodge, S.A.; Cleveland, J.C.; Kress, T.S.; Petek, M. [Oak Ridge National Lab., TN (United States)

    1992-10-01

    This report provides the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to develop a technical basis for evaluating the effectiveness and feasibility of current and proposed strategies for boiling water reactor (BWR) severe accident management. First, the findings of an assessment of the current status of accident management strategies for the mitigation of in-vessel events for BWR severe accident sequences are described. This includes a review of the BWR Owners` Group Emergency Procedure Guidelines (EPGSs) to determine the extent to which they currently address the characteristic events of an unmitigated severe accident and to provide the basis for recommendations for enhancement of accident management procedures. Second, where considered necessary, new candidate accident management strategies are proposed for mitigation of the late-phase (after core damage has occurred) events. Finally, recommendations are made for consideration of additional strategies where warranted, and two of the four candidate strategies identified by this effort are assessed in detail: (1) preparation of a boron solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and (2) containment flooding to maintain the core debris within the reactor vessel if the injection systems cannot be restored.

  19. Analysis of Core Physics Experiments on Irradiated BWR MOX Fuel in REBUS Program

    Energy Technology Data Exchange (ETDEWEB)

    Yamamoto, Toru; Ando, Yoshihira [Japan Nuclear Energy Safety Organization, Safety Standard Division, Tokyo (Japan); Hayashi, Yamato [Toshiba Corporation, Power System Company, Yokohama (Japan)

    2008-07-01

    As part of analyses of experimental data of a critical core containing a irradiated BWR MOX test bundle in the REBUS program, depletion calculations was performed for the BWR MOX fuel assemblies from that the MOX test rods were selected by using a general purpose neutronics code system SRAC. The core analyses were carried out using SRAC and a continuous energy Monte Carlo code MVP. The calculated k{sub eff}s were compared with those of the core containing a fresh MOX fuel bundle in the program. The SRAC-diffusion calculation underestimates k{sub eff}s of the both cores by 1.0 to 1.3 %dk and the k{sub eff}s of MVP are 1.001. The difference in k{sub eff} between the irradiated BWR MOX test bundle core and the fresh MOX one is 0.4 %dk in the SRAC-diffusion calculation and 0.0 %dk in the MVP calculation. The calculated fission rate distributions are in good agreement with the measurement in the SRAC-diffusion and MVP calculations. The calculated neutron flux distributions are also in good agreement with the measurement. The calculated burnup reactivity in the both calculations well reproduce the measurements. (authors)

  20. Status report: Intergranular stress corrosion cracking of BWR core shrouds and other internal components

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-03-01

    On July 25, 1994, the US Nuclear Regulatory Commission (NRC) issued Generic Letter (GL) 94-03 to obtain information needed to assess compliance with regulatory requirements regarding the structural integrity of core shrouds in domestic boiling water reactors (BWRs). This report begins with a brief description of the safety significance of intergranular stress corrosion cracking (IGSCC) as it relates to the design and function of BWR core shrouds and other internal components. It then presents a brief history of shroud cracking events both in the US and abroad, followed by an indepth summary of the industry actions to address the issue of IGSCC in BWR core shrouds and other internal components. This report summarizes the staff`s basis for issuing GL 94-03, as well as the staff`s assessment of plant-specific responses to GL 94-03. The staff is continually evaluating the licensee inspection programs and the results from examinations of BWR core shrouds and other internal components. This report is representative of submittals to and evaluations by the staff as of September 30, 1995. An update of this report will be issued at a later date.

  1. High fidelity analysis of BWR fuel assembly with COBRA-TF/PARCS and trace codes

    Energy Technology Data Exchange (ETDEWEB)

    Abarca, A.; Miro, R.; Barrachina, T.; Verdu, G., E-mail: aabarca@isirym.upv.es, E-mail: rmiro@iqn.upv.es, E-mail: tbarrachina@isirym.upv.es, E-mail: gverdu@iqn.upv.es [Universitat Politecnica de Valencia, (ISIRYM/UPV), (Spain). Institute for Industrial, Radiophysical and Environmental Safety; Concejal, A.; Melara, J.; Albendea, M., E-mail: acbe@iberdrola.es, E-mail: jls@iberdrola.es, E-mail: manuel.albendea@iberdrola.es [Iberdrola, Madrid (Spain); Soler, A., E-mail: asoler@iberdrola.es [SEA Propulsion SL, Madrid (Spain)

    2013-07-01

    The growing importance of detailed reactor core and fuel assembly description for light water reactors (LWRs) as well as the sub-channel safety analysis requires high fidelity models and coupled neutronic/thermalhydraulic codes. Hand in hand with advances in the computer technology, the nuclear safety analysis is beginning to use a more detailed thermal hydraulics and neutronics. Previously, a PWR core and a 16 by 16 fuel assembly models were developed to test and validate our COBRA-TF/PARCS v2.7 (CTF/PARCS) coupled code. In this work, a comparison of the modeling and simulation advantages and disadvantages of modern 10 by 10 BWR fuel assembly with CTF/PARCS and TRACE codes has been done. The objective of the comparison is making known the main advantages of using the sub-channel codes to perform high resolution nuclear safety analysis. The sub-channel codes, like CTF, permits obtain accurate predictions, in two flow regime, of the thermalhydraulic parameters important to safety with high local resolution. The modeled BWR fuel assembly has 91 fuel rods (81 full length and 10 partial length fuel rods) and a big square central water rod. This assembly has been modeled with high level of detail with CTF code and using the BWR modeling parameters provided by TRACE. The same neutronic PARCS's model has been used for the simulation with both codes. To compare the codes a coupled steady state has be performed. (author)

  2. MELCOR 1.8.2 assessment: The DF-4 BWR Damaged Fuel experiment

    Energy Technology Data Exchange (ETDEWEB)

    Tautges, T.J.

    1993-10-01

    MELCOR is a fully integrated, engineering-level computer code being developed at Sandia National Laboratories for the USNRC, that models the entire spectrum of severe accident phenomena in a unified framework for both BWRs and PWRs. As a part of an ongoing assessment, program, MELCOR has been used to model the ACRR in-pile DF-4 Damaged Fuel experiment. DF-4 provided data for early phase melt progression in BWR fuel assemblies, particularly for phenomena associated with eutectic interactions in the BWR control blade and zircaloy oxidation in the canister and cladding. MELCOR provided good agreement with experimental data in the key areas of eutectic material behavior and canister and cladding oxidation. Several shortcomings associated with the MELCOR modeling of BWR geometries were found and corrected. Twenty-five sensitivity studies were performed on COR, HS and CVH parameters. These studies showed that the new MELCOR eutectics model played an important role in predicting control blade behavior. These studies revealed slight time step dependence and no machine dependencies. Comparisons made with the results from four best-estimate codes showed that MELCOR did as well as these codes in matching DF-4 experimental data.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-07-01

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

  4. Logical model for the control of a BWR turbine;Modelo logico para el control de una turbina de un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Vargas O, Y. [Universidad del Valle de Mexico, Campus Toluca, Av. Las Palmas No. 136, Col. San Jorge Pueblo Nuevo, 52140 Metepec, Estado de Mexico (Mexico); Amador G, R.; Ortiz V, J.; Castillo D, R., E-mail: yonaeton@hotmail.co [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2009-07-01

    In this work a design of a logical model is presented for the turbine control of a nuclear power plant with a BWR like energy source. The model is sought to implement later on inside the thermal hydraulics code of better estimate RELAP/SCDAPSIM. The logical model is developed for the control and protection of the turbine, and the consequent protection to the BWR, considering that the turbine control will be been able to use for one or several turbines in series. The quality of the present design of the logical model of the turbine control is that it considers the most important parameters in the operation of a turbine, besides that they have incorporated to the logical model the secondary parameters that will be activated originally as true when the turbine model is substituted by a detailed model. The development of the logical model of a turbine will be of utility in the short and medium term to carry out analysis on the turbine operation with different operation conditions, of vapor extraction, specific steps of the turbine to feed other equipment s, in addition to analyze the separate and the integrated effect. (Author)

  5. Spent fuel data base: commercial light water reactors. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hauf, M.J.; Kniazewycz, B.G.

    1979-12-01

    As a consequence of this country's non-proliferation policy, the reprocessing of spent nuclear fuel has been delayed indefinitely. This has resulted in spent light water reactor (LWR) fuel being considered as a potential waste form for disposal. Since the Nuclear Regulatory Commission (NRC) is currently developing methodologies for use in the regulation of the management and disposal of high-level and transuranic wastes, a comprehensive data base describing LWR fuel technology must be compiled. This document provides that technology baseline and, as such, will support the development of those evaluation standards and criteria applicable to spent nuclear fuel.

  6. MicroRNA-99a inhibits insulin-induced proliferation, migration, dedifferentiation, and rapamycin resistance of vascular smooth muscle cells by inhibiting insulin-like growth factor-1 receptor and mammalian target of rapamycin.

    Science.gov (United States)

    Zhang, Zi-Wei; Guo, Rui-Wei; Lv, Jin-Lin; Wang, Xian-Mei; Ye, Jin-Shan; Lu, Ni-Hong; Liang, Xing; Yang, Li-Xia

    2017-04-29

    Patients with type 2 diabetes mellitus (T2DM) are characterized by insulin resistance and are subsequently at high risk for atherosclerosis. Hyperinsulinemia has been associated with proliferation, migration, and dedifferentiation of vascular smooth muscle cells (VSMCs) during the pathogenesis of atherosclerosis. Moreover, insulin-like growth factor-1 receptor (IGF-1R) and mammalian target of rapamycin (mTOR) have been demonstrated to be the underlying signaling pathways. Recently, microRNA-99a (miR-99a) has been suggested to regulate the phenotypic changes of VSMCs in cancer cells. However, whether it is involved in insulin-induced changes of VSCMs has not been determined. In this study, we found that insulin induced proliferation, migration, and dedifferentiation of mouse VSMCs in a dose-dependent manner. Furthermore, the stimulating effects of high-dose insulin on proliferation, migration, and dedifferentiation of mouse VSMCs were found to be associated with the attenuation of the inhibitory effects of miR-99a on IGF-1R and mTOR signaling activities. Finally, we found that the inducing effect of high-dose insulin on proliferation, migration, and dedifferentiation of VSMCs was partially inhibited by an active mimic of miR-99a. Taken together, these results suggest that miR-99a plays a key regulatory role in the pathogenesis of insulin-induced proliferation, migration, and phenotype conversion of VSMCs at least partly via inhibition of IGF-1R and mTOR signaling. Our results provide evidence that miR-99a may be a novel target for the treatment of hyperinsulinemia-induced atherosclerosis. Copyright © 2017 Elsevier Inc. All rights reserved.

  7. SHP-1 activation inhibits vascular smooth muscle cell proliferation and intimal hyperplasia in a rodent model of insulin resistance and diabetes

    DEFF Research Database (Denmark)

    Qi, Weier; Li, Qian; Liew, Chong Wee

    2017-01-01

    as Map2k1) and increased DNA methylation of the Shp-1 promoter. VSMCs from Shp-1-Tg mice exhibited impaired platelet-derived growth factor (PDGF)-stimulated tyrosine phosphorylation with a concomitant decrease in PDGF-stimulated VSMC proliferation and migration. Similarly, HFD-fed Shp-1-Tg mice and mice...

  8. Determination of stresses caused by fluctuation of acoustic load in the steam dryers of a BWR; Determinacion de esfuerzos originados por fluctuacion de carga acustica en los secadores de vapor de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Centeno P, J.; Quezada G, S.; Prieto G, A.; Vazquez R, A.; Espinosa P, G. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Av. San Rafael Atlixco No. 186, Col. Vicentina, 09340 Mexico D. F. (Mexico); Nunez C, A., E-mail: javcuami26@hotmail.com [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

    2014-10-15

    The extended power up-rate (EPU) in a nuclear power plant cause various problems in BWR components also in the steam system. This due to increased steam flow generated in the reactor and is conveyed to the turbine by the four main steam lines (MSL). One of the most serious problems is the generation of acoustic pressure loads in the metal structure of the steam dryer which eventually leads to fatigue failure and even the appearance of cracks, and in turn it causes loose parts that are entrained by the steam and transported in the MSL. This problem is due to the fluctuation of load acoustics caused by the union of the safety or relief valves (SRV) with the MSL, spreading through these to reach the reactor pressure vessel (RPV) where the effect of resonance of the acoustic wave is amplified and impacts directly in the supporting structure of the steam dryer, skirt and the panels where the mixture liquid-steam is dried, by centrifugation effect and runoff of liquid water. Efforts in the steam dryer operating conditions of EPU for two cases will be analyzed in this work, the first is before the installation of Acoustic Side Branch (ASB), and in the second case we consider the installation of said ASB in the standpipes of SRV. The analysis was performed with numerical experiments on a platform for computational fluid dynamics with virtual geometries previously designed based on the actual components of the reactor and steam system. The model to study is delimited by the top of the RPV, the steam dryer and a section of each of the four MSL with ten standpipes of SRV. With the obtained data and considering the mechanical-structural properties of the steam dryer material, we can evaluate the mechanical resistance to impacts by acoustic pressure load and its possible deformation or cracking. (Author)

  9. CAF-derived HGF promotes cell proliferation and drug resistance by up-regulating the c-Met/PI3K/Akt and GRP78 signalling in ovarian cancer cells.

    Science.gov (United States)

    Deying, Wei; Feng, Geng; Shumei, Liang; Hui, Zhao; Ming, Liu; Hongqing, Wang

    2017-04-28

    The tumour microenvironment is a highly heterogeneous entity that plays crucial roles in cancer progression. As the most prominent stromal cell types, cancer-associated fibroblasts (CAFs) produce a variety of factors into the tumour microenvironment. In the present study, we firstly isolated CAFs from tumour tissues of the patients with ovarian cancer and demonstrated that the hepatocyte growth factor (HGF) was highly expressed in the supernatants of CAFs. CAF-derived HGF or human recombinant HGF promoted cell proliferation in human ovarian cell lines SKOV3 and HO-8910 cells. Western blotting analysis also showed that CAF-derived HGF or recombinant HGF activated c-Met/phosphoinositide 3-kinase (PI3K)/Akt and glucose-regulated protein 78 (GRP78) signalling pathways in ovarian cancer cells, and these effects could be abrogated by anti-HGF and c-Met inhibitor INCB28060. Moreover, HGF in CAF matrix attenuated paclitaxel (PAC)-caused inhibition of cell proliferation and increase in cell apoptosis through activating c-Met/PI3K/Akt and GRP78 pathways in SKOV3 and HO-8910 cells. The results in vitro were further validated in nude mice. These findings suggest that CAF-derived HGF plays crucial roles in cell proliferation and drug resistance in ovarian cancer cells. © 2017 The Author(s).

  10. Analysis of assemblies exchange in the core of a reactor BWR; Analisis del intercambio de ensambles en el nucleo de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Kauil U, J. S. [Universidad Autonoma de Yucatan, Facultad de Ingenieria, Av. Industrias no contaminantes por Anillo Periferico Norte s/n, Apdo. Postal 150 Cordemex, Merida, Yucatan (Mexico); Fuentes M, L.; Castillo M, J. A.; Ortiz S, J. J.; Perusquia del Cueto, R., E-mail: san_dino@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-10-15

    The performance of the core of a boiling water reactor (BWR) was evaluated when two assemblies are exchanged during the fuel reload in erroneous way. All with the purpose of analyzing the value of the neutrons effective multiplication factor and the thermal limits for an exchange of assemblies. In their realization the mentioned study was based in a transition cycle of the Unit 1 of the nuclear power plant of Laguna Verde. The obtained results demonstrate that when carrying out an exchange between two fuel assemblies in erroneous way, with regard to the original reload, the changes in the neutrons effective multiplication factor do not present a serious problem, unless the exchange has been carried out among a very burnt assembly with one fresh, where this last is taken to the periphery. (Author)

  11. Obtention control bars patterns for a BWR using Tabo search; Obtencion de patrones de barras de control para un BWR usando busqueda Tabu

    Energy Technology Data Exchange (ETDEWEB)

    Castillo, A.; Ortiz, J.J.; Alonso, G. [Instituto Nacional de Investigaciones Nucleares, Km. 36.5 Carretera Mexico-Toluca, Ocoyoacac, Estado de Mexico 52045 (Mexico); Morales, L.B. [UNAM, IIMAS, Ciudad Universitaria, D. F. 04510 (Mexico); Valle, E. del [IPN, ESFM, Unidad Profesional ' Adolfo Lopez Mateos' , Col. Lindavista 07738, D. F. (Mexico)]. e-mail: jacm@nuclear.inin.mx

    2004-07-01

    The obtained results when implementing the technique of tabu search, for to optimize patterns of control bars in a BWR type reactor, using the CM-PRESTO code are presented. The patterns of control bars were obtained for the designs of fuel reloads obtained in a previous work, using the same technique. The obtained results correspond to a cycle of 18 months using 112 fresh fuels enriched at the 3.53 of U-235. The used technique of tabu search, prohibits recently visited movements, in the position that correspond to the axial positions of the control bars, additionally the tiempo{sub t}abu matrix is used for to manage a size of variable tabu list and the objective function is punished with the frequency of the forbidden movements. The obtained patterns of control bars improve the longitude of the cycle with regard to the reference values and they complete the restrictions of safety. (Author)

  12. Thermal hydraulics characterization of the core and the reactor vessel type BWR; Caracterizacion termohidraulica del nucleo y de la vasija de un reactor tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Zapata Y, M.; Lopez H, L.E. [CFE, Carretera Cardel-Nautla Km. 42.5, Municipio Alto Lucero, Veracruz (Mexico)]. e-mail: marxlenin.zapata@cfe.gob.mx

    2008-07-01

    The thermal hydraulics design of a reactor type BWR 5 as the employees in the nuclear power plant of Laguna Verde involves the coupling of at least six control volumes: Pumps jet region, Stratification region, Core region, Vapor dryer region, Humidity separator region and Reactor region. Except by the regions of the core and reactor, these control volumes only are used for design considerations and their importance as operative data source is limited. It is for that is fundamental to complement the thermal hydraulics relations to obtain major data that allow to determine the efficiency of internal components, such as pumps jet, humidity separator and vapor dryer. Like example of the previous thing, calculations are realized on the humidity of the principal vapor during starting, comparing it with the values at the moment incorporated in the data banks of the computers of process of both units. (Author)

  13. Study of transient rod extraction failure without RBM in a BWR; Estudio del transitorio error de extraccion de barra sin RBM en un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Vallejo Q, J. A.; Martin del Campo M, C.; Fuentes M, L.; Francois L, J. L., E-mail: amhed_jvq@hotmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico)

    2015-09-15

    The study and analysis of the operational transients are important for predicting the behavior of a system to short-term events and the impact that would cause this transient. For the nuclear industry these studies are indispensable due to economic, environmental and social impacts that could cause an accident during the operation of a nuclear reactor. In this paper the preparation, simulation and analysis results of the transient rod extraction failure in which not taken into operation the RBM is presented. The study was conducted for a BWR of 2027 MWt, in an intermediate cycle of its useful life and using the computer code Simulate-3K a scenario of anomalies was created in the core reactivity which gave a coherent prediction to the type of presented event. (Author)

  14. Study of transient turbine shot without bypass in a BWR; Estudio del transitorio disparo de turbina sin bypass en un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Vallejo Q, J. A.; Martin del Campo M, C.; Fuentes M, L.; Francois L, J. L., E-mail: amhed_jvq@hotmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico)

    2015-09-15

    The study and analysis of operational transients are important for predicting the behavior of a system to short-terms events and the impact that would cause this transition. For the nuclear industry these studies are indispensable due to economic, environmental and social impacts that could result in an accident during the operation of a nuclear reactor. In this paper the preparation, simulation and analysis of results of a turbine shot transient, which is not taken into operation the bypass is presented. The study is realized for a BWR of 2027 MWt, to an intermediate cycle life and using the computer code Simulate-3K a depressurization stage of the vessel is created which shows the response of other security systems and gives a coherent prediction to the event presented type. (Author)

  15. Optimization of fuel reloads for a BWR using the ant colony system; Optimizacion de recargas de combustible para un BWR usando el sistema de colonia de hormigas

    Energy Technology Data Exchange (ETDEWEB)

    Esquivel E, J. [Universidad Autonoma del Estado de Mexico, Facultad de Ingenieria, Cerro de Coatepec s/n, Ciudad Universitaria, 50110 Toluca, Estado de Mexico (Mexico); Ortiz S, J. J. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)], e-mail: jaime.es.jaime@gmail.com

    2009-10-15

    In this work some results obtained during the development of optimization systems are presented, which are employees for the fuel reload design in a BWR. The systems use the ant colony optimization technique. As first instance, a system is developed that was adapted at travel salesman problem applied for the 32 state capitals of Mexican Republic. The purpose of this implementation is that a similarity exists with the design of fuel reload, since the two problems are of combinatorial optimization with decision variables that have similarity between both. The system was coupled to simulator SIMULATE-3, obtaining good results when being applied to an operation cycle in equilibrium for reactors of nuclear power plant of Laguna Verde. (Author)

  16. Over-expression of CHAF1A promotes cell proliferation and apoptosis resistance in glioblastoma cells via AKT/FOXO3a/Bim pathway

    Energy Technology Data Exchange (ETDEWEB)

    Peng, Honghai; Du, Bin [Department of Neurosurgery, Jinan Central Hospital Affiliated to Shandong University, Jinan, Shandong 250013 (China); Jiang, Huili [Friendship Nephrology and Blood Purification Center, Jinan Central Hospital Affiliated to Shandong University, Jinan, Shandong 250013 (China); Gao, Jun, E-mail: gaoj1666@126.com [Department of Neurosurgery, Jinan Central Hospital Affiliated to Shandong University, Jinan, Shandong 250013 (China)

    2016-01-22

    Chromatinassembly factor 1 subunit A (CHAF1A) has been reported to be involved in several human diseases including cancer. However, the biological and clinical significance of CHAF1A in glioblastoma progression remains largely unknown. In this study, we found that up-regulation of CHAF1A happens frequently in glioblastoma tissues and is associated with glioblastoma prognosis. Knockout of CHAF1A by CRISPR/CAS9 technology induce G1 phase arrest and apoptosis in glioblastoma cell U251 and U87. In addition, inhibition of CHAF1A influenced the signal transduction of the AKT/FOXO3a/Bim axis, which is required for glioblastoma cell proliferation. Taken together, these results show that CHAF1A contributes to the proliferation of glioblastoma cells and may be developed as a de novo drug target and prognosis biomarker of glioblastoma.

  17. Estimation of dose rate around the spent control rods of a BWR; Estimacion de la rapidez de dosis alrededor de las barras de control gastadas de un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Cancino P, G.

    2016-10-01

    The energy can come from fossil renewable sources (solar (natural gas, oil), wind, hydro, tidal, geothermal, biomass, bio energy and nuclear. Nuclear power can be obtained by fission reactions and fusion (still under investigation) atomic nuclei. Fission, is a partition of a very heavy nucleus (Uranium 235, for example) into two lighter nuclei. Much of the world's electric power is generated from the energy released by fission processes. In a nuclear power reactor, light water as the BWR, there are many important elements that allow safe driving operation, one of them are the elements or control systems, the burnable poison or neutron absorber inherently allow control power reactor. The control rods, which consist mostly of stainless steel and absorbing elements (such as boron carbide, hafnium, cadmium, among others) of thermal neutrons is able to initiate, regulate or stop the reactor power. These, due to the use of depleted burned or absorbing material and therefore reach their lifespan, which can be 15 years or have other values depending on the manufacturer. Control rods worn should be removed, stored or confined in expressly places. Precisely at this stage arises the importance of knowing their radiological condition to manipulate safely and without incident to the people health responsible for conducting these proceedings state arises. This thesis consists in the estimation of the dose rate in spent control rod made of boron carbide, from a typical BWR reactor. It will be estimated by direct radiation measurements with measurement equipment for radiotherapy ionization chamber, in six spent control rods, which were taken at different reactor operating cycles and are in a spent fuel pool. Using bracket electromechanical and electronic equipment for positioning and lifting equipment for radiation measurement around the control rod in the axial and radial arrangement for proper scanning. Finally will be presented a graphic corresponding to the dose

  18. Crack growth rate in core shroud horizontal welds using two models for a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Arganis Juárez, C.R., E-mail: carlos.arganis@inin.gob.mx; Hernández Callejas, R.; Medina Almazán, A.L.

    2015-05-15

    Highlights: • Two models were used to predict SCC growth rate in a core shroud of a BWR. • A weld residual stress distribution with 30% stress relaxation by neutron was used. • Agreement is shown between the measurements of SCC growth rate and the predictions. • Slip–oxidation model is better at low fluences and empirical model at high fluences. - Abstract: An empirical crack growth rate correlation model and a predictive model based on the slip–oxidation mechanism for Stress Corrosion Cracking (SCC) were used to calculate the crack growth rate in a BWR core shroud. In this study, the crack growth rate was calculated by accounting for the environmental factors related to aqueous environment, neutron irradiation to high fluence and the complex residual stress conditions resulting from welding. In estimating the SCC behavior the crack growth measurements data from a Boiling Water Reactor (BWR) plant are referred to, and the stress intensity factor vs crack depth throughout thickness is calculated using a generic weld residual stress distribution for a core shroud, with a 30% stress relaxation induced by neutron irradiation. Quantitative agreement is shown between the measurements of SCC growth rate and the predictions of the slip–oxidation mechanism model for relatively low fluences (5 × 10{sup 24} n/m{sup 2}), and the empirical model predicted better the SCC growth rate than the slip–oxidation model for high fluences (>1 × 10{sup 25} n/m{sup 2}). The relevance of the models predictions for SCC growth rate behavior depends on knowing the model parameters.

  19. Non-disturbance technique to estimate the core stability of BWR

    Energy Technology Data Exchange (ETDEWEB)

    Shieh, D.J.; Chang, S.I.

    1980-09-01

    The feasibility of estimating the core instability margin of BWR by analyzing the fluctuation of normal operation neutron flux signal is surveyed. It concludes that the instability margin estimated by this technique will be comparable with the small reactivity perturbation test result. The DDS method used to estimate the natural frequency and decay ratio is briefly described and verified by some simulation cases. Then the APRM signal taken from Chin-Shan Nuclear Power Station Unit 1 is analyzed. The result shows that the natural frequency is 0.777 +- 0.039 Hz, and the decay ratio is 0.277 +- 0.045.

  20. An on-line method to monitor BWR core stability based on an autocorrelation method

    Energy Technology Data Exchange (ETDEWEB)

    Yokomizo, Osamu; Masuhara, Yasuhiro (Hitachi Ltd., Ibaraki (Japan). Energy Research Lab.); Yoshimoto, Yuichiro (Hitachi Ltd., Ibaraki (Japan). Hitachi Works)

    1990-03-01

    A on-line monitoring method is introduced for BWR core stability. The method utilizes only autocorrelation values for two delay time intervals. Its simplicity makes it suitable for an on-line monitor. Accuracy of the core decay ratio calculated by the method improves as the core condition approaches instability. The error in the decay ratio for regional limit cycle oscillations is 0.2% when calculated from local signals in the most unstable region, and 4% when calculated from core averaged signals. (orig.).

  1. Effect of nonuniform 3-D void distribution within BWR fuel assemblies on neutronic characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Toshikazu; Hyoudou, Hideaki; Kitada, Takanori [Osaka Univ., Suita (Japan). Dept. of Nuclear Engineering]. E-mail: takeda@nucl.eng.osaka-u.ac.jp; Kosaka, Shinya; Ikeda, Hideaki [Tepco Systems Corp., Tokyo (Japan)]. E-mail: kisaka-shinya@tepsys.co.jp

    2003-07-01

    The effect of nonuniform distribution of void fractions within fuel assemblies of BWR cores on the neutronic characteristics has been evaluated by coupling the thermal hydraulic and neutronic calculations. It is found that the effect is large for assemblies with Gd rods because of the small void fraction for subchannels with Gd rods. The axially averaged K{sub {infinity}} is reduced by about 0.1 {approx} 0.3% {delta}k/k compared with the case with uniform void distribution. The power peaking is decreased by 0.6 {approx} 1.3%. The reason of these changes is discussed. (author)

  2. TRACE code validation for BWR spray cooling injection based on GOTA facility experiments

    Energy Technology Data Exchange (ETDEWEB)

    Racca, S. [San Piero a Grado Nuclear Research Group (GRNSPG), Pisa (Italy); Kozlowski, T. [Royal Inst. of Tech., Stockholm (Sweden)

    2011-07-01

    Best estimate codes have been used in the past thirty years for the design, licensing and safety of NPP. Nevertheless, large efforts are necessary for the qualification and the assessment of such codes. The aim of this work is to study the main phenomena involved in the emergency spray cooling injection in a Swedish designed BWR. For this purpose, data from the Swedish separate effect test facility GOTA have been simulated using TRACE version 5.0 Patch 2. Furthermore, uncertainty calculations have been performed with the propagation of input errors method and the identification of the input parameters that mostly influence the peak cladding temperature has been performed. (author)

  3. Animation of Antimicrobial Resistance

    Medline Plus

    Full Text Available ... Animal & Veterinary Safety & Health Antimicrobial Resistance Animation of Antimicrobial Resistance Share Tweet Linkedin Pin it More sharing ... CVM) produced a nine-minute animation explaining how antimicrobial resistance both emerges and proliferates among bacteria. Over ...

  4. Animation of Antimicrobial Resistance

    Medline Plus

    Full Text Available ... Veterinary Safety & Health Antimicrobial Resistance Animation of Antimicrobial Resistance Share Tweet Linkedin Pin it More sharing options ... produced a nine-minute animation explaining how antimicrobial resistance both emerges and proliferates among bacteria. Over time, ...

  5. Animation of Antimicrobial Resistance

    Science.gov (United States)

    ... Veterinary Safety & Health Antimicrobial Resistance Animation of Antimicrobial Resistance Share Tweet Linkedin Pin it More sharing options ... produced a nine-minute animation explaining how antimicrobial resistance both emerges and proliferates among bacteria. Over time, ...

  6. Bisphenol A activates EGFR and ERK promoting proliferation, tumor spheroid formation and resistance to EGFR pathway inhibition in estrogen receptornegative inflammatory breast cancer cells

    Science.gov (United States)

    Background: Inflammatory breast cancer (IBC) is a distinct and the deadliest breast cancer variant, which shows a rapid rate of progression and acquired therapeutic resistance. Epidemiological studies suggest that chemical exposure in the environment and consumer products can aff...

  7. Nicotine promotes cell proliferation and induces resistance to cisplatin by α7 nicotinic acetylcholine receptor‑mediated activation in Raw264.7 and El4 cells.

    Science.gov (United States)

    Wang, Yan Yan; Liu, Yao; Ni, Xiao Yan; Bai, Zhen Huan; Chen, Qiong Yun; Zhang, Ye; Gao, Feng Guang

    2014-03-01

    Although nicotine is a risk factor for carcinogenesis and atherosclerosis, epidemiological data indicate that nicotine has therapeutic benefits in treating Alzheimer's disease. Our previous studies also showed that nicotine-treated dendritic cells have potential antitumor effects. Hence, the precise effects of nicotine on the biological characterizations of cells are controversial. The aim of the present study was to assess the roles of α7 nicotinic acetylcholine receptors (nAChRs), Erk1/2-p38-JNK and PI3K-Akt pathway in nicotine-mediated proliferation and anti-apoptosis effects. The results firstly showed that nicotine treatment clearly augmented cell viability and upregulated PCNA expression in both Raw264.7 and El4 cells. Meanwhile, nicotine afforded protection against cisplatin-induced toxicity through inhibiting caspase-3 activation and upregulating anti-apoptotic protein expression. Further exploration demonstrated that nicotine efficiently abolished cisplatin-promoted mitochondria translocation of Bax and the release of cytochrome c. The pretreatment of α-bungarotoxin and tubocurarine chloride significantly attenuated nicotine-augmented cell viability, abolished caspase-3 activation and α7 nAChR upregulation. Both Erk-JNK-p38 and PI3K-Akt signaling pathways could be activated by nicotine treatment in Raw264.7 and El4 cells. Notably, when Erk-JNK and PI3K-Akt activities were inhibited, nicotine-augmented cell proliferation and anti-apoptotic effects were abolished accordingly. The results presented here indicate that nicotine could achieve α7 nAChR-mediated proliferation and anti-apoptotic effects by activating Erk-JNK and PI3K-Akt pathways respectively, providing potential therapeutic molecules to deal with smoking-associated human diseases.

  8. Interpretation of the VENUS VIP-BWR program with APOLLO2.8 reference and optimized MOC schemes for 8 x 8 UO{sub 2} and MOX BWR assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Blaise, P., E-mail: patrick.blaise@cea.f [CEA - Commissariat a l' Energie Atomique, Centre de CADARACHE - DEN/CAD/DER/SPRC, F-13108 Saint Paul-Lez-Durance (France); Vidal, J.-F.; Santamarina, A.; Bernard, D. [CEA - Commissariat a l' Energie Atomique, Centre de CADARACHE - DEN/CAD/DER/SPRC, F-13108 Saint Paul-Lez-Durance (France)

    2010-11-15

    The interpretation of the VIP-BWR program conducted in the CEN.SCK Mol VENUS critical facility (Belgium), has been performed with the new APOLLO2.8 product and its CEA2005V4.1 library based on the JEFF3.1.1 file. Both reference SHEM-MOC (281groups without equivalence) and Optimized BWR 26G (26 groups with equivalence) schemes are used for UO{sub 2} and MOX BWR assembly calculations. The VIP-BWR program was aimed to provide an experimental database for BWR neutronics tools in mixed Gd poisoned configurations with 8 x 8 UO{sub 2} and MOX assemblies. The experimental conditions are relatively representative of actual industrial BWR core characteristics, at least in terms of void fraction. Measured pin-by-pin power distributions enable to exact valuable information at various interfaces. For fresh (UO{sub 2}/UO{sub 2}-Gd) and recycled UO{sub 2} (UO{sub 2} only) cores loadings, the information is given through the 'UO{sub 2}' core. In the case of partial MOX loadings (UO{sub 2}/MOX interface), the power distributions are available through the 'T-MOX' core. All critical sizes are predicted within 1 with SHEM-MOC reference calculation scheme. For UO{sub 2} core, the (C-E) on k{sub eff} are (95 {+-} 266) pcm and (203 {+-} 266) pcm for SHEM-MOC and Optimized scheme respectively. For MOX core, the results are (87 {+-} 214) pcm and (283 {+-} 214) pcm. The uncertainties take into account both measurement uncertainties and technological uncertainties such as enrichment, clad thicknesses, grid pitch or fuel densities. Average (C-E)/E on radial fission rate distributions are in excellent agreement in the UO{sub 2} and UO{sub 2}-Gd assemblies. In mixed loading cores, the calculation shows an over prediction in the MOX fuel pin. This points out a systematic swing between UO{sub 2} and MOX. For average fission rates in UO{sub 2}, the C/E results are within {+-} 0.2% with SHEM-MOC and OptimizedBWR26G. For average power in MOX, the C/E results are within {+-}0

  9. Internal Tandem Duplication in FLT3 Attenuates Proliferation and Regulates Resistance to the FLT3 Inhibitor AC220 by Modulating p21Cdkn1a and Pbx1 in Hematopoietic Cells.

    Directory of Open Access Journals (Sweden)

    Mariko Abe

    Full Text Available Internal tandem duplication (ITD mutations in the Fms-related tyrosine kinase 3 (FLT3 gene (FLT3-ITD are associated with poor prognosis in patients with acute myeloid leukemia (AML. Due to the development of drug resistance, few FLT3-ITD inhibitors are effective against FLT3-ITD+ AML. In this study, we show that FLT3-ITD activates a novel pathway involving p21Cdkn1a (p21 and pre-B cell leukemia transcription factor 1 (Pbx1 that attenuates FLT3-ITD cell proliferation and is involved in the development of drug resistance. FLT3-ITD up-regulated p21 expression in both mouse bone marrow c-kit+-Sca-1+-Lin- (KSL cells and Ba/F3 cells. The loss of p21 expression enhanced growth factor-independent proliferation and sensitivity to cytarabine as a consequence of concomitantly enriching the S+G2/M phase population and significantly increasing the expression of Pbx1, but not Evi-1, in FLT3-ITD+ cells. This enhanced cell proliferation following the loss of p21 was partially abrogated when Pbx1 expression was silenced in FLT3-ITD+ primary bone marrow colony-forming cells and Ba/F3 cells. When FLT3-ITD was antagonized with AC220, a selective inhibitor of FLT3-ITD, p21 expression was decreased coincident with Pbx1 mRNA up-regulation and a rapid decline in the number of viable FLT3-ITD+ Ba/F3 cells; however, the cells eventually became refractory to AC220. Overexpressing p21 in FLT3-ITD+ Ba/F3 cells delayed the emergence of cells that were refractory to AC220, whereas p21 silencing accelerated their development. These data indicate that FLT3-ITD is capable of inhibiting FLT3-ITD+ cell proliferation through the p21/Pbx1 axis and that treatments that antagonize FLT3-ITD contribute to the subsequent development of cells that are refractory to a FLT3-ITD inhibitor by disrupting p21 expression.

  10. Development of a prototype pin-by-pin fine mesh calculation code for BWR core analysis

    Energy Technology Data Exchange (ETDEWEB)

    Tada, Kenichi; Yamamoto, Akio; Yamane, Yoshihiro [Nagoya University, Nagoya (Japan); Kosaka, Shinya; Hirano, Gou [TEPCO SYSTEMS CORPORATION, Tokyo (Japan)

    2008-07-01

    A prototype core analysis code for BWR, SUBARU, which is based on the three-dimensional pin-by-pin fine-mesh calculation, is being developed. The SUBARU code has several features, e.g., incorporation of the SP3 transport theory, capability to treat the staggered meshes, and so on. In this paper, to estimate the prediction accuracy of this core analysis code, a hypothetical 2D ABWR core which is consisted by 8x8 low-enrichment UO{sub 2} fuel assembly, 9x9 high-enrichment UO{sub 2} fuel assembly, and 10x10 MOX fuel assembly is analyzed. To investigate the prediction accuracy, we compared the pin-wise fission rate distribution which was obtained by the cell-heterogeneous transport calculation by MOC. To evaluate the computational costs, a hypothetical 3D ABWR core is also used. These results suggest that SUBARU would have enough accuracy and reasonable calculation costs for the reference BWR core analysis when further investigation is taken into account. (authors)

  11. Design study of Thorium-232 and Protactinium-231 based fuel for long life BWR

    Science.gov (United States)

    Trianti, N.; Su'ud, Z.; Riyana, E. S.

    2012-06-01

    A preliminary design study for the utilization of thorium added with 231Pa based fuel on BWR type reactor has been performed. In the previous research utilization of fuel based Thorium-232 and Uranium-233 show 10 years operation time with maximum excess-reactivity about 4.075% dk/k. To increase reactor operation time and reduce excess-reactivity below 1% dk/k, Protactinium (Pa-231) is used as Burnable Poison. Protactinium-231 has very interesting neutronic properties, which enable the core to reduce initial excess-reactivity and simultaneously increase production of 233U to 231Pa in burn-up process. Optimizations of the content of 231Pa in the core enables the BWR core to sustain long period of operation time with reasonable burn-up reactivity swing. Based on the optimization of fuel element composition (Th and Pa) in various moderation ratio we can get reactor core with longer operation time, 20 ˜ 30 years operation without fuel shuffling or refuelling, with average power densities maximum of about 35 watt/cc, and maximum excess-reactivity 0.56% dk/k.

  12. On the Decay Ratio Determination in BWR Stability Analysis by Auto-Correlation Function Techniques

    Energy Technology Data Exchange (ETDEWEB)

    Behringer, K.; Hennig, D

    2002-11-01

    A novel auto-correlation function (ACF) method has been investigated for determining the oscillation frequency and the decay ratio in BWR stability analyses. The neutron signals are band-pass filtered to separate the oscillation peak in the power spectral density (PSD) from background. Two linear second-order oscillation models are considered. These models, corrected for signal filtering and including a background term under the peak in the PSD, are then least-squares fitted to the ACF of the previously filtered neutron signal, in order to determine the oscillation frequency and the decay ratio. Our method uses fast Fourier transform techniques with signal segmentation for filtering and ACF estimation. Gliding 'short-term' ACF estimates on a record allow the evaluation of uncertainties. Numerical results are given which have been obtained from neutron data of the recent Forsmark I and Forsmark II NEA benchmark project. Our results are compared with those obtained by other participants in the benchmark project. The present PSI report is an extended version of the publication K. Behringer, D. Hennig 'A novel auto-correlation function method for the determination of the decay ratio in BWR stability studies' (Behringer, Hennig, 2002)

  13. Fracture toughness of irradiated wrought and cast austenitic stainless steels in BWR environment

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O.K.; Gruber, E.E.; Shack, W.J. [Argonne National Lab., Nuclear Engineering Div., Argonne, Illinois (United States)

    2007-07-01

    Experimental data are presented on the fracture toughness of wrought and cast austenitic stainless steels (SSs) that were irradiated to a fluence of {approx} 1.5 x 10{sup 21} n/cm{sup 2} (E > 1 MeV){sup *} ({approx} 2.3 dpa) at 296-305{sup o}C. To evaluate the possible effects of test environment and crack morphology on the fracture toughness of these steels, all tests were conducted in normal-water-chemistry boiling water reactor (BWR) environments at {approx} 289{sup o}C. Companion tests were also conducted in air on the same material for comparison. The fracture toughness J-R curves for SS weld heat-affected-zone materials in BWR water were found to be comparable to those in air. However, the results of tests on sensitized Type 304 SS and thermally aged cast CF-8M steel suggested a possible effect of water environment. The available fracture toughness data on irradiated austenitic SSs were reviewed to assess the potential for radiation embrittlement of reactor-core internal components. The synergistic effects of thermal and radiation embrittlement of cast austenitic SS internal components are also discussed. (author)

  14. Transient and stability analysis of a BWR core with thorium-uranium fuel

    Energy Technology Data Exchange (ETDEWEB)

    Nunez-Carrera, Alejandro [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779 Col. Narvarte, 03020 Mexico, DF (Mexico); Espinosa-Paredes, Gilberto [Division de Ciencias Basicas e Ingenieria, Universidad Autonoma Metropolitana, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico, DF (Mexico)], E-mail: gepe@xanum.uam.mx; Francois, Juan-Luis [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532, 62550 Jiutepec Mor. (Mexico)

    2008-08-15

    The kinetic response of a boiling water reactor (BWR) equilibrium core using thorium as a nuclear material, in an integrated blanket-seed assembly, is presented in this work. Additionally an in-house code was developed to evaluate this core under steady state and transient conditions including a stability analysis. The code has two modules: (a) the time domain module for transient analysis and (b) the frequency domain module for stability analysis. The thermal-hydraulic process is modeled by a set of five equations, considering no homogeneous flow with drift-flux approximation and non-equilibrium thermodynamic. The neutronic process is calculated with a point kinetics model. Typical BWR reactivity effects are considered: void fraction, fuel temperature, moderator temperature and control rod density. Collapsed parameters were included in the code to represent the core using an average fuel channel. For the stability analysis, in the frequency domain, the transfer function is determined by applying Laplace-transforming to the calculated pressure drop perturbations in each of the considered regions where a constant total pressure drop was considered. The transfer function was used to study the system response in the frequency domain when an inlet flow perturbation is applied. The results show that the neutronic behavior of the core with thorium uranium fuel is similar to a UO{sub 2} core, even during transient conditions. The stability and transient analysis show that the thorium-uranium fuel can be operated safely in current BWRs.

  15. Corrosion fatigue initiation behaviour of wrought austenitic stainless pipe steels under simulated BWR/HWC and PWR conditions

    Energy Technology Data Exchange (ETDEWEB)

    Leber, H.J.; Ritter, S.; Seifert, H.P [Paul Scherrer Institute, Nuclear Energy and Safety Research Department, Laboratory for Nuclear Materials, 5232 Villigen PSI (Switzerland)

    2011-07-01

    The corrosion fatigue (CF) initiation and short crack growth behavior of different low-carbon and stabilized austenitic stainless steels was characterized under simulated BWR and primary PWR conditions by cyclic fatigue tests with sharply notched fracture mechanics specimens in the temperature range from 70 to 320 C. Environmental reduction of fatigue initiation life was observed in all stainless steels at strain rates {<=} 0.1 %/s in BWR and PWR environment. The stationary short crack CF crack growth rates after crack advances of 50 to 300 {mu}m from the notch-root were in the typical range of corresponding results from tests with long cracks (pre-cracked specimens) and also showed the same system parameter response. The effect of environment on the initiation process ({Delta}a = 10 {mu}m) was relevantly stronger than on the subsequent stationary short crack growth. Both, under BWR/HWC and PWR conditions, a relevant environmental reduction of fatigue initiation life occurred for the combination of temperatures {>=} 100 C, notch strain rates {<=} 0.1 %/s and notch strain amplitudes {>=} 0.3 %. If these conjoint threshold conditions were simultaneously satisfied, the environmental enhancement increased with decreasing strain rate and increasing temperature. Material and water chemistry parameters usually only had a little effect. Sensitization affected the CF behavior under highly oxidizing BWR/NWC conditions only. Preliminary block loading experiments did not reveal significant static load hold period effects on the technical corrosion fatigue initiation life. If the critical requirements were satisfied, the BWR/HWC and PWR environments usually resulted in acceleration of short fatigue crack growth by a factor of 5 to 20 with respect to air. Solution annealed steels showed slightly shorter CF initiation lives, but also lower stationary short CF crack growth rates under BWR/HWC and PWR conditions with low ECPs than under highly oxidizing BWR/NWC conditions. A very

  16. THERMAL EVALUATION OF THE USE OF BWR MOX SNF IN THE MULTI-PURPOSE CANISTER (MPC) WITH ACD DISPOSAL CONTAINER (SCPB: N/A)

    Energy Technology Data Exchange (ETDEWEB)

    T.L. Lotz

    1995-11-13

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) as specified in the Waste Package Implementation Plan (pp. 4-8,4-11,4-24,5-1, and 5-13; Ref. 5.10) and Waste Package Plan (pp. 3-15,3-17, and 3-24; Ref. 5.9). The design data request addressed herein is: (1) Characterize the conceptual 40 BWR Multi-Purpose Canister (MPC) Waste Package (WP) design to show that the design is feasible for use in the MGDS environment when loaded with BWR MOX SNF. (2) Characterize the conceptual 24 BWR Multi-Purpose Canister (MPC) Waste Package (WP) design to show that the design is feasible for use in the MGDS environment when loaded with BWR MOX SNF. The purpose of this analysis is to respond a concern that the long-term disposal thermal issues for the Multi-Purpose Canister (MPC) Subsystem Design, if used with SNF designed for a MOX fuel cycle, do not preclude MPC compatibility with the MGDS. The objective of this analysis is to provide thermal parameter information for the conceptual MPC design with disposal container which is loaded with BWR MOX SNF under nominal MGDS repository conditions. The results are intended to show that the design has a reasonable chance to meet the MGDS design requirements for normal MGDS operation, to provide the required guidance to determining the major design issues for future design efforts, and to show that the BWR MOX SNF loaded MPC performance is similar to an MPC loaded with commercial BWR SNF. Future design efforts will focus on specific MPC vendor designs and BWR MOX SNF designs when they become available.

  17. Artificial intelligence applied to fuel management in BWR type reactors; Inteligencia artificial aplicada a la administracion de combustible en reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz S, J.J

    1998-10-01

    In this work two techniques of artificial intelligence, neural networks and genetic algorithms were applied to a practical problem of nuclear fuel management; the determination of the optimal fuel reload for a BWR type reactor. This is an important problem in the design of the operation cycle of the reactor. As a result of the application of these techniques, comparable or even better reloads proposals than those given by expert companies in the subject were obtained. Additionally, two other simpler problems in reactor physics were solved: the determination of the axial power profile and the prediction of the value of some variables of interest at the end of the operation cycle of the reactor. Neural networks and genetic algorithms have been applied to solve many problems of engineering because of their versatility but they have been rarely used in the area of fuel management. The results obtained in this thesis indicates the convenience of undertaking further work on this area and suggest the application of these techniques of artificial intelligence to the solution of other problems in nuclear reactor physics. (Author)

  18. Optimization of fuel cells for BWR based in Tabu modified search; Optimizacion de celdas de combustible para BWR basada en busqueda Tabu modificada

    Energy Technology Data Exchange (ETDEWEB)

    Martin del Campo M, C.; Francois L, J.L. [Facultad de Ingenieria, UNAM, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico); Palomera P, M.A. [Facultad de Ingenieria, UNAM, Posgrado en Ingenieria en Computacion, Circuito exterior s/n, Ciudad Universitaria, Mexico, D.F. (Mexico)]. e-mail: cmcm@fi-b.unam.mx

    2004-07-01

    The advances in the development of a computational system for the design and optimization of cells for assemble of fuel of Boiling Water Reactors (BWR) are presented. The method of optimization is based on the technique of Tabu Search (Tabu Search, TS) implemented in progressive stages designed to accelerate the search and to reduce the time used in the process of optimization. It was programed an algorithm to create the first solution. Also for to diversify the generation of random numbers, required by the technical TS, it was used the Makoto Matsumoto function obtaining excellent results. The objective function has been coded in such a way that can adapt to optimize different parameters like they can be the enrichment average or the peak factor of radial power. The neutronic evaluation of the cells is carried out in a fine way by means of the HELIOS simulator. In the work the main characteristics of the system are described and an application example is presented to the design of a cell of 10x10 bars of fuel with 10 different enrichment compositions and gadolinium content. (Author)

  19. Prediction of the stability of BWR reactors during the start-up process; Prediccion de la estabilidad de reactores BWR durante el proceso de arranque

    Energy Technology Data Exchange (ETDEWEB)

    Ruiz E, J.A.; Castillo D, R. [ININ, Km. 36.5 Carretera Mexico-Toluca, 52045 Salazar, Estado de Mexico (Mexico); Blazquez M, J.B. [Centro de Investigaciones Energetics, Medioambientales y Tecnologicas, Av Complutense 22, 28040 Madrid (Spain)

    2004-07-01

    The Boiling Water Reactors (BWR) are susceptible of uncertainties of power when they are operated to low flows of coolant (W) and high powers (P), being presented this situation mainly in the start-up process. The start-up process could be made but sure if the operator knew the value of the stability index Decay reason (Dr) before going up power and therefore to guarantee the stability. The power and the flow are constantly measures, the index Dr could also be considered its value in real time. The index Dr depends on the power, flow and many other values, such as, the distribution of the flow axial and radial neutronic, the temperature of the feeding water, the fraction of holes and other thermohydraulic and nuclear parameters. A simple relationship of Dr is derived leaving of the pattern reduced of March-Leuba, where three independent variables are had that are the power, the flow and a parameter that it contains the rest of the phenomenology, that is to say all the other quantities that affect the value of Dr. This relationship developed work presently and verified its prediction with data of start-up of commercial reactors could be used for the design of a practical procedure practice of start-up, what would support to the operator to prevent this type of events of uncertainty. (Author)

  20. SUN-RAH: a nucleoelectric BWR university simulator based in reduced order models; SUN-RAH: simulador universitario de nucleoelectrica BWR basado en modelos de orden reducido

    Energy Technology Data Exchange (ETDEWEB)

    Morales S, J.B.; Lopez R, A.; Sanchez B, A.; Sanchez S, R.; Hernandez S, A. [DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: jms0620@yahoo.com

    2003-07-01

    The development of a simulator that allows to represent the dynamics of a nucleo electric central, with nuclear reactor of the BWR type, using reduced order models is presented. These models present the characteristics defined by the dominant poles of the system (1) and most of those premature operation transitories in a power station can be reproduced with considerable fidelity if the models are identified with data of plant or references of a code of better estimate like RAMONA, TRAC (2) or RELAP. The models of the simulator are developments or own simplifications starting from the physical laws and retaining the main terms. This work describes the objective of the project and the general specifications of the University student of Nucleo electric simulator with Boiling Water Reactor type (SUN-RAH) as well as the finished parts that fundamentally are the nuclear reactor, the one of steam supply (NSSS), the plant balance (BOP), the main controllers of the plant and the implemented graphic interfaces. The pendent goals as well as the future developments and applications of SUN-RAH are described. (Author)

  1. Chondrosarcoma: A Rare Misfortune in Aging Human Cartilage? The Role of Stem and Progenitor Cells in Proliferation, Malignant Degeneration and Therapeutic Resistance

    Directory of Open Access Journals (Sweden)

    Karen A. Boehme

    2018-01-01

    Full Text Available Unlike other malignant bone tumors including osteosarcomas and Ewing sarcomas with a peak incidence in adolescents and young adults, conventional and dedifferentiated chondrosarcomas mainly affect people in the 4th to 7th decade of life. To date, the cell type of chondrosarcoma origin is not clearly defined. However, it seems that mesenchymal stem and progenitor cells (MSPC in the bone marrow facing a pro-proliferative as well as predominantly chondrogenic differentiation milieu, as is implicated in early stage osteoarthritis (OA at that age, are the source of chondrosarcoma genesis. But how can MSPC become malignant? Indeed, only one person in 1,000,000 will develop a chondrosarcoma, whereas the incidence of OA is a thousandfold higher. This means a rare coincidence of factors allowing escape from senescence and apoptosis together with induction of angiogenesis and migration is needed to generate a chondrosarcoma. At early stages, chondrosarcomas are still assumed to be an intermediate type of tumor which rarely metastasizes. Unfortunately, advanced stages show a pronounced resistance both against chemo- and radiation-therapy and frequently metastasize. In this review, we elucidate signaling pathways involved in the genesis and therapeutic resistance of chondrosarcomas with a focus on MSPC compared to signaling in articular cartilage (AC.

  2. Experimental study for quantative aging evaluation of epoxy liner in BWR nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Na, H. S. [KEPRI, Taejon (Korea, Republic of); Song, Y. C.; Kim, N. Y. [Korea Univ. of Educational Technology, Chonnon (Korea, Republic of)

    2001-10-01

    The purpose of this study is an experimental approach to quantitatively evaluate the aging status of epoxy coating onto containment structure in BWR nuclear power plant. Based on accelerated aging experiment for 64 days, adhesion test was performed to evaluate an physical bonding. To compare with adhesion data, both impedance data by UT and data by thermal gravimetric analysis were obtained during experiment. At almost 50% of adhesion force decrease, it was identified that aging phenamena of epoxy such as pine hole, blistering was discovered. Coating to establish aging status of epoxy, relations among three kinds of different data were analyze. By compatibility of these data, physical aging situation of as-built epoxy coating was figured out. The possibility to develop new methodology of time-dependent aging status on epoxy coating was identified.

  3. Development of a parametric containment event tree model for a severe BWR accident

    Energy Technology Data Exchange (ETDEWEB)

    Okkonen, T. [OTO-Consulting Ay, Helsinki (Finland)

    1995-04-01

    A containment event tree (CET) is built for analysis of severe accidents at the TVO boiling water reactor (BWR) units. Parametric models of severe accident progression and fission product behaviour are developed and integrated in order to construct a compact and self-contained Level 2 PSA model. The model can be easily updated to correspond to new research results. The analyses of the study are limited to severe accidents starting from full-power operation and leading to core melting, and are focused mainly on the use and effects of the dedicated severe accident management (SAM) systems. Severe accident progression from eight plant damage states (PDS), involving different pre-core-damage accident evolution, is examined, but the inclusion of their relative or absolute probabilities, by integration with Level 1, is deferred to integral safety assessments. (33 refs., 5 figs., 7 tabs.).

  4. Non-local two phase flow momentum transport in S BWR

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa P, G.; Salinas M, L.; Vazquez R, A., E-mail: gepe@xanum.uam.mx [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Area de Ingenieria en Recursos Energeticos, Apdo. Postal 55-535, 09340 Ciudad de Mexico (Mexico)

    2015-09-15

    The non-local momentum transport equations derived in this work contain new terms related with non-local transport effects due to accumulation, convection, diffusion and transport properties for two-phase flow. For instance, they can be applied in the boundary between a two-phase flow and a solid phase, or in the boundary of the transition region of two-phase flows where the local volume averaging equations fail. The S BWR was considered to study the non-local effects on the two-phase flow thermal-hydraulic core performance in steady-state, and the results were compared with the classical local averaging volume conservation equations. (Author)

  5. Modeling of BWR core meltdown accidents - for application in the MELRPI. MOD2 computer code

    Energy Technology Data Exchange (ETDEWEB)

    Koh, B R; Kim, S H; Taleyarkhan, R P; Podowski, M Z; Lahey, Jr, R T

    1985-04-01

    This report summarizes improvements and modifications made in the MELRPI computer code. A major difference between this new, updated version of the code, called MELRPI.MOD2, and the one reported previously, concerns the inclusion of a model for the BWR emergency core cooling systems (ECCS). This model and its computer implementation, the ECCRPI subroutine, account for various emergency injection modes, for both intact and rubblized geometries. Other changes to MELRPI deal with an improved model for canister wall oxidation, rubble bed modeling, and numerical integration of system equations. A complete documentation of the entire MELRPI.MOD2 code is also given, including an input guide, list of subroutines, sample input/output and program listing.

  6. Application of passive auto catalytic recombiner (PAR) for BWR plants in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kobayashi, K.; Murano, K. [Tokyo Electric Power Co., Inc. (Japan); Yamanari, S. [Hitachi Cable, Ltd., Tokyo (Japan); Yamamoto, Y. [Toshiba Corp., Yokohama (Japan)

    2001-07-01

    The passive auto-catalytic recombiner (PAR), which can recombine flammable gases such as hydrogen and oxygen with each other to avoid an explosion in case of a loss-of-coolant accident (LOCA), installed in the primary containment vessel does not require a power supply or dynamic equipment, while the existing flammability gas control system (FCS) of most BWRs as an outer loop of the primary containment vessel needs them to make flammable gases circulate through blowers and heaters in the system. PAR offers a number of advantages over existing FCS, such as high reliability, low cost due to much smaller amount of materials needed, good maintainability, good operability in case of a LOCA, and smaller space for installation. An experimental study has been carried out for the purpose of solving the problems of applying PAR to Japanese BWR plants instead of existing FCS, in which we grasped the basic characteristics of PAR. (author)

  7. An efficient modeling method for thermal stratification simulation in a BWR suppression pool

    Energy Technology Data Exchange (ETDEWEB)

    Haihua Zhao; Ling Zou; Hongbin Zhang; Hua Li; Walter Villanueva; Pavel Kudinov

    2012-09-01

    The suppression pool in a BWR plant not only is the major heat sink within the containment system, but also provides major emergency cooling water for the reactor core. In several accident scenarios, such as LOCA and extended station blackout, thermal stratification tends to form in the pool after the initial rapid venting stage. Accurately predicting the pool stratification phenomenon is important because it affects the peak containment pressure; and the pool temperature distribution also affects the NPSHa (Available Net Positive Suction Head) and therefore the performance of the pump which draws cooling water back to the core. Current safety analysis codes use 0-D lumped parameter methods to calculate the energy and mass balance in the pool and therefore have large uncertainty in prediction of scenarios in which stratification and mixing are important. While 3-D CFD methods can be used to analyze realistic 3D configurations, these methods normally require very fine grid resolution to resolve thin substructures such as jets and wall boundaries, therefore long simulation time. For mixing in stably stratified large enclosures, the BMIX++ code has been developed to implement a highly efficient analysis method for stratification where the ambient fluid volume is represented by 1-D transient partial differential equations and substructures such as free or wall jets are modeled with 1-D integral models. This allows very large reductions in computational effort compared to 3-D CFD modeling. The POOLEX experiments at Finland, which was designed to study phenomena relevant to Nordic design BWR suppression pool including thermal stratification and mixing, are used for validation. GOTHIC lumped parameter models are used to obtain boundary conditions for BMIX++ code and CFD simulations. Comparison between the BMIX++, GOTHIC, and CFD calculations against the POOLEX experimental data is discussed in detail.

  8. Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steels in BWR Environments

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Chopra, O. K. [Argonne National Lab. (ANL), Argonne, IL (United States); Gruber, Eugene E. [Argonne National Lab. (ANL), Argonne, IL (United States); Shack, William J. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2010-06-01

    The internal components of light water reactors are exposed to high-energy neutron irradiation and high-temperature reactor coolant. The exposure to neutron irradiation increases the susceptibility of austenitic stainless steels (SSs) to stress corrosion cracking (SCC) because of the elevated corrosion potential of the reactor coolant and the introduction of new embrittlement mechanisms through radiation damage. Various nonsensitized SSs and nickel alloys have been found to be prone to intergranular cracking after extended neutron exposure. Such cracks have been seen in a number of internal components in boiling water reactors (BWRs). The elevated susceptibility to SCC in irradiated materials, commonly referred to as irradiation-assisted stress corrosion cracking (IASCC), is a complex phenomenon that involves simultaneous actions of irradiation, stress, and corrosion. In recent years, as nuclear power plants have aged and irradiation dose increased, IASCC has become an increasingly important issue. Post-irradiation crack growth rate and fracture toughness tests have been performed to provide data and technical support for the NRC to address various issues related to aging degradation of reactor-core internal structures and components. This report summarizes the results of the last group of tests on compact tension specimens from the Halden-II irradiation. The IASCC susceptibility of austenitic SSs and heat-affected-zone (HAZ) materials sectioned from submerged arc and shielded metal arc welds was evaluated by conducting crack growth rate and fracture toughness tests in a simulated BWR environment. The fracture and cracking behavior of HAZ materials, thermally sensitized SSs and grain-boundary engineered SSs was investigated at several doses (≤3 dpa). These latest results were combined with previous results from Halden-I and II irradiations to analyze the effects of neutron dose, water chemistry, alloy compositions, and welding and processing conditions on IASCC

  9. CSAU methodology and results for an ATWS event in a BWR using information theory methods

    Energy Technology Data Exchange (ETDEWEB)

    Munoz-Cobo, J.L., E-mail: jlcobos@iqn.upv.es [Universitat Politècnica de València, Thermal-Hydraulics and Nuclear Engineering Group (TIN), Institute for Energy Engineering (IEE), Valencia (Spain); Escrivá, A., E-mail: aescriva@iqn.upv.es [Universitat Politècnica de València, Thermal-Hydraulics and Nuclear Engineering Group (TIN), Institute for Energy Engineering (IEE), Valencia (Spain); Mendizabal, R., E-mail: rmsanz@csn.es [Consejo de Seguridad Nuclear, 28040 Madrid (Spain); Pelayo, F., E-mail: fpl@csn.es [Consejo de Seguridad Nuclear, 28040 Madrid (Spain); Melara, J., E-mail: jls@iberdrola.es [IBERINCO, IBERDROLA Ingeniería y Construcción, Madrid (Spain)

    2014-10-15

    Highlights: • We apply the CSAU methodology to an ATWS in a BWR using information theory methods. • We show how to perform the selection of the most influential inputs on the critical safety parameter. • We apply the maximum entropy principle to get the input parameter distribution. • We examine the maximum relative entropy principle to update the input parameter PDF. • We quantify the uncertainty of the critical safety parameter using order statistics and information theory. - Abstract: This paper shows an application of the CSAU methodology to an ATWS in a BWR reactor, when the temperature of the suppression pool is taken as the critical safety parameter. The method combines CSAU methodology with recent techniques of information theory. In this paper we use auxiliary tools to help in the evaluation and improvement of the parameters distribution that enter in the elements II and III of CSAU based methodologies. These tools have been implemented in two FORTRAN programs: GEDIPA (Generation of the Parameter Distribution) and UNTHERCO (Uncertainty in Thermal Hydraulic Codes). The first one analyzes the information data available on a given parameter or parameters with the goal to know all the information about the probability distribution function of these parameters. The second apply information theory methods, as the maximum entropy principle (MEP) and the maximum relative entropy Principle (MREP), in order to build conservative distribution functions for the parameters from the available data. Also, the distribution function of a given parameter can be updated using the MREP principle when new information is provided. UNTHERCO performs the MONTECARLO sampling for a given set of parameters when the distribution function of these parameters is previously known. If the distribution of a parameter is unknown, then, the MEP is applied to deduce the distribution function for this parameter.

  10. Characteristics of debris in the lower head of a BWR in different severe accident scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Phung, Viet-Anh, E-mail: vaphung@kth.se; Galushin, Sergey, E-mail: galushin@kth.se; Raub, Sebastian, E-mail: raub@kth.se; Goronovski, Andrei, E-mail: andreig@kth.se; Villanueva, Walter, E-mail: walterv@kth.se; Kööp, Kaspar, E-mail: kaspar@safety.sci.kth.se; Grishchenko, Dmitry, E-mail: dmitry@safety.sci.kth.se; Kudinov, Pavel, E-mail: pavel@safety.sci.kth.se

    2016-08-15

    Highlights: • Station blackout scenario with delayed recovery of safety systems in a Nordic BWR is considered. • Genetic algorithm and random sampling methods are used to explore accident scenario domain. • Main groups of scenarios are identified. • Ranges and distributions of characteristics of debris bed in the lower head are determined. - Abstract: Nordic boiling water reactors (BWRs) adopt ex-vessel debris cooling to terminate severe accident progression. Core melt released from the vessel into a deep pool of water is expected to fragment and form a coolable debris bed. Characteristics of corium melt ejection from the vessel determine conditions for molten fuel–coolant interactions (FCI) and debris bed formation. Non-coolable debris bed or steam explosion can threaten containment integrity. Vessel failure and melt ejection mode are determined by the in-vessel accident progression. Characteristics (such as mass, composition, thermal properties, timing of relocation, and decay heat) of the debris bed formed in the process of core relocation into the vessel lower plenum define conditions for the debris reheating, remelting, melt-vessel structure interactions, vessel failure and melt release. Thus core degradation and relocation are important sources of uncertainty for the success of the ex-vessel accident mitigation strategy. The goal of this work is improve understanding how accident scenario parameters, such as timing of failure and recovery of different safety systems can affect characteristics of the debris in the lower plenum. Station blackout scenario with delayed power recovery in a Nordic BWR is considered using MELCOR code. The recovery timing and capacity of safety systems were varied using genetic algorithm (GA) and random sampling methods to identify two main groups of scenarios: with relatively small (<20 tons) and large (>100 tons) amount of relocated debris. The domains are separated by the transition regions, in which relatively small

  11. Analysis of the performance of fuel cells BWR with a single enrichment and radial distribution of enrichments; Analisis del desempeno de celdas combustibles BWR con un solo enriquecimiento y con distribucion radial de enriquecimientos

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez, J. A.; Vargas, S.; Alonso, G.; Del Valle, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. Lindavista, Mexico D. F. 07738 (Mexico); Xolocostli M, J. V. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: govaj666@hotmail.com

    2008-07-01

    The efficient use of the fuel is one of the objectives in the assemblies design of type BWR. The present tendency in the assemblies design of type BWR is through a radial distribution of enrichments. The present work has like object showing the because of this decision, for what a comparison of the neutronic performance of two fuel cells with the same enrichment average but one of them with radial distribution of enrichment and the other with a single enrichment equal to the average. The cells were analyzed with the CASMO-4 code and the obtained results of the behavior of the neutron flow and the power sustain the because of the radial distribution of enrichments. (Author)

  12. Proliferation of myogenic stem cells in human skeletal muscle in response to low-load resistance training with blood-flow restriction

    DEFF Research Database (Denmark)

    Nielsen, Jakob Lindberg; Aagaard, Per; Bech, Rune Dueholm

    2012-01-01

    Low-load resistance training with blood-flow restriction has been shown to elicit substantial increases in muscle mass and muscle strength; however the effect on myogenic stem cells (MSC) and myonuclei number remains unexplored. Ten male subjects (22.8±2.3 yrs) performed 4 sets of knee extensor...... exercise (20% 1RM) to concentric failure during blood-flow restriction (BFR) of the proximal thigh (100 mmHg), while eight work-matched controls (21.9±3.0 yrs) trained without BFR (CON). 23 training sessions were performed within 19 days. Maximal isometric knee extensor strength (MVC) was examined pre...... and post training, while muscle biopsies were obtained at baseline (Pre), after 8 days intervention (Mid8) and 3 (Post3) and 10 days (Post10) post training to examine changes in myofibre area (MFA), MSC and myonuclei number. MVC increased by 7.1% (Post5) and 10.6% (Post12) (P...

  13. Comparative analysis of results between CASMO, MCNP and Serpent for a suite of Benchmark problems on BWR reactors; Analisis comparativo de resultados entre CASMO, MCNP y SERPENT para una suite de problemas Benchmark en reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Xolocostli M, J. V.; Vargas E, S.; Gomez T, A. M. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Reyes F, M. del C.; Del Valle G, E., E-mail: vicente.xolocostli@inin.gob.mx [IPN, Escuela Superior de Fisica y Matematicas, UP - Adolfo Lopez Mateos, Edif. 9, 07738 Mexico D. F. (Mexico)

    2014-10-15

    In this paper a comparison is made in analyzing the suite of Benchmark problems for reactors type BWR between CASMO-4, MCNP6 and Serpent code. The Benchmark problem consists of two different geometries: a fuel cell of a pin and assembly type BWR. To facilitate the study of reactors physics in the fuel pin their nuclear characteristics are provided to detail, such as burnt dependence, the reactivity of selected nuclide, etc. With respect to the fuel assembly, the presented results are regarding to infinite multiplication factor for burning different steps and different vacuum conditions. Making the analysis of this set of Benchmark problems provides comprehensive test problems for the next fuels generation of BWR reactors with high extended burned. It is important to note that when making this comparison the purpose is to validate the methodologies used in modeling for different operating conditions, if the case is of other BWR assembly. The results will be within a range with some uncertainty, considering that does not depend on code that is used. Escuela Superior de Fisica y Matematicas of Instituto Politecnico Nacional (IPN (Mexico) has accumulated some experience in using Serpent, due to the potential of this code over other commercial codes such as CASMO and MCNP. The obtained results for the infinite multiplication factor are encouraging and motivate the studies to continue with the generation of the X S of a core to a next step a respective nuclear data library is constructed and this can be used by codes developed as part of the development project of the Mexican Analysis Platform of Nuclear Reactors AZTLAN. (Author)

  14. Pre-study of dynamic loads on the internals caused by a large pipe break in a BWR; Foerstudie av stroemningsinducerade laster paa interndelar vid brott i huvudcirkulationskretsarna i BWR

    Energy Technology Data Exchange (ETDEWEB)

    Marcinkiewicz, Jerzy; Lindgren, Anders [Det Norske Veritas Nuclear Technology AB, Stockholm (Sweden)

    2002-12-01

    Det Norske Veritas Nuclear Technology has performed a literature study of dynamic load on a BWR (Boiling Water Reactor) internals caused by a large pipe break. The goal of the study was to improve the knowledge about the physics of phenomena occurring in the RPV (Reactor Pressure Vessel) after pipe break in the main circulation system and also to make a review of calculation methods, models and computer programs including their capabilities when calculating the dynamic loads. The report presents description of relevant parts of a BWR, initial and boundary conditions, and phenomena determining the loads - rapid depressurization and propagation of pressure wave (including none-equilibrium). Furthermore, the report generally describes possible methodologies for calculating the dynamic loads on internals after the pipe break and the experiences from calculations the dynamic loads with different methods (computer programs) including comparisons with experimental data. Fluid-Structure Interaction methodology and its importance for calculation of dynamic loads on reactor internals is discussed based on experimental data. A very intensive research program for studying and calculating the dynamic loads on internals after pipe breaks has been performed in USA and Germany during the seventies and the eighties. Several computer programs have been developed and a number of large-scale experiments have been performed to calibrate the calculation methods. In spite of the fact that all experiments were performed for PWR several experiences should be valid also for BWR. These experiences, connected mainly to capabilities of computer programs calculating dynamic loads, are discussed in the report.

  15. Experimental results and analysis of core physics experiments, FUBILA, for high burn-up BWR full MOX cores

    Energy Technology Data Exchange (ETDEWEB)

    Yamamoto, T.; Kikuchi, S.; Kawashima, K.; Kamimura, K. [Japan Nuclear Energy Safety Organization, 3-17-1, Toranomon, Minato-ku, Tokyo 105-0001 (Japan)

    2006-07-01

    JNES has been performing MOX core physics experiments, FUBILA, in the EOLE critical facility of the CEA Cadarache center with collaboration of a French Consortium (CEA and COGEMA). The experiments have been designed to obtain the core physics data of operating conditions of full MOX BWR cores consisting of high burn up BWR MOX assemblies. The experiments consisting of seven different core configurations started from January 2005 and will be completed by August 2006. Theoretical analysis of the experimental data has been also carried out using a deterministic code, SRAC, and a continuous energy Monte Carlo calculation code, MVP, with major nuclear data libraries, JENDL-3.3, 3.2, ENDF/B-VI and JEFF-3.1 for the first critical core. (authors)

  16. BWR water chemistry guidelines and PWR primary water chemistry guidelines in Japan – Purpose and technical background

    Energy Technology Data Exchange (ETDEWEB)

    Kawamura, Hirotaka, E-mail: kawamuh@criepi.denken.or.jp [Central Research Institute of Electric Power Industry (Japan); Hirano, Hideo [Central Research Institute of Electric Power Industry (Japan); Katsumura, Yousuke [University of Tokyo (Japan); Uchida, Shunsuke [Tohoku University (Japan); Mizuno, Takayuki [Mie University (Japan); Kitajima, Hideaki; Tsuzuki, Yasuo [Japan Nuclear Safety Institute (Japan); Terachi, Takumi [Institute of Nuclear Safety System, Inc. (Japan); Nagase, Makoto; Usui, Naoshi [Hitachi-GE Nuclear Energy, Ltd. (Japan); Takagi, Junichi; Urata, Hidehiro [Toshiba Corporation (Japan); Shoda, Yasuhiko; Nishimura, Takao [Mitsubishi Heavy Industry, Ltd. (Japan)

    2016-12-01

    Highlights: • Framework of BWR/PWR water chemistry Guidelines in Japan are presented. • Guideline necessity, definitions, philosophy and technical background are mentioned. • Some guideline settings for control parameters and recommendations are explaines. • Chemistry strategy is also mentioned. - Abstract: After 40 years of light water reactor (LWR) operations in Japan, the sustainable development of water chemistry technologies has aimed to ensure the highest coolant system component integrity and fuel reliability performance for maintaining LWRs in the world; additionally, it aimed to achieve an excellent dose rate reduction. Although reasonable control and diagnostic parameters are utilized by each boiling water reactor (BWR) and pressurized water reactor (PWR) owner, it is recognized that specific values are not shared among everyone involved. To ensure the reliability of BWR and PWR operation and maintenance, relevant members of the Atomic Energy Society of Japan (AESJ) decided to establish guidelines for water chemistry. The Japanese BWR and PWR water chemistry guidelines provide strategies to improve material and fuel reliability performance as well as to reduce dosing rates. The guidelines also provide reasonable “control values”, “diagnostic values” and “action levels” for multiple parameters, and they stipulate responses when these levels are exceeded. Specifically, “conditioning parameters” are adopted in the Japanese PWR primary water chemistry guidelines. Good practices for operational conditions are also discussed with reference to long-term experience. This paper presents the purpose, technical background and framework of the preliminary water chemistry guidelines for Japanese BWRs and PWRs. It is expected that the guidelines will be helpful as an introduction to achieve safety and reliability during operations.

  17. Parametric study of the potential for BWR ECCS strainer blockage due to LOCA generated debris. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Zigler, G.; Brideau, J.; Rao, D.V.; Shaffer, C.; Souto, F.; Thomas, W. [Science and Engineering Associates, Inc., Albuquerque, NM (United States)

    1995-10-01

    This report documents a plant-specific study for a BWR/4 with a Mark I containment that evaluated the potential for LOCA generated debris and the probability of losing long term recirculation capability due ECCS pump suction strainer blockage. The major elements of this study were: (1) acquisition of detailed piping layouts and installed insulation details for a reference BWR; (2) analysis of plant specific piping weld failure probabilities to estimate the LOCA frequency; (3) development of an insulation and other debris generation and drywell transport models for the reference BWR; (4) modeling of debris transport in the suppression pool; (5) development of strainer blockage head loss models for estimating loss of NPSH margin; (6) estimation of core damage frequency attributable to loss of ECCS recirculation capability following a LOCA. Elements 2 through 5 were combined into a computer code, BLOCKAGE 2.3. A point estimate of overall DEGB pipe break frequency (per Rx-year) of 1.59E-04 was calculated for the reference plant, with a corresponding overall ECCS loss of NPSH frequency (per Rx-year) of 1.58E-04. The calculated point estimate of core damage frequency (per Rx-year) due to blockage related accident sequences for the reference BWR ranged from 4.2E-06 to 2.5E-05. The results of this study show that unacceptable strainer blockage and loss of NPSH margin can occur within the first few minutes after ECCS pumps achieve maximum flows when the ECCS strainers are exposed to LOCA generated fibrous debris in the presence of particulates (sludge, paint chips, concrete dust). Generic or unconditional extrapolation of these reference plant calculated results should not be undertaken.

  18. Animation of Antimicrobial Resistance

    Medline Plus

    Full Text Available ... how antimicrobial resistance both emerges and proliferates among bacteria. Over time, the use of antimicrobial drugs will result in the development of resistant strains of bacteria, complicating clinician's efforts to select the appropriate antimicrobial ...

  19. Animation of Antimicrobial Resistance

    Medline Plus

    Full Text Available ... Medicine (CVM) produced a nine-minute animation explaining how antimicrobial resistance both emerges and proliferates among bacteria. ... concept more understandable to non-scientists by showing how bacterial antimicrobial resistance can develop and spread. All ...

  20. Sustainability of CD24 expression, cell proliferation and migration, cisplatin-resistance, and caspase-3 expression during mesenchymal-epithelial transition induced by the removal of TGF-β1 in A549 lung cancer cells.

    Science.gov (United States)

    Kim, Seong-Kwan; Park, Jin-A; Zhang, Dan; Cho, Sang-Hyun; Yi, Hee; Cho, Soo-Min; Chang, Byung-Joon; Kim, Jin-Suk; Shim, Jae-Han; Abd El-Aty, A M; Shin, Ho-Chul

    2017-08-01

    Epithelial-mesenchymal transition (EMT) is a notable mechanism underlying cancer cell metastasis. Transforming growth factor β1 (TGF-β1) has been used to induce EMT; however, there is a lack of information regarding the role of TGF-β1 in mesenchymal-epithelial transition (MET). In the present study, EMT was induced in A549 lung cancer cells using TGF-β1 (TGF-β1-treated group) and MET was induced sequentially from the TGF-β1-treated group by removing the TGF-β1 (MET/return group). Untreated A549 lung cancer cells were used as a control. Characteristic features, including cancer stem cell markers [cluster of differentiation (CD)24, CD44 and CD133], cell proliferation and migration and diverse intracellular mechanisms, were observed in all groups. Using western blot analysis, the TGF-β1-treated group demonstrated increased vimentin and reduced E-cadherin expression, whereas the MET/return group demonstrated the opposite trend. Among cancer stem cell markers, the population of CD24low cells was reduced in the TGF-β1-treated group. Furthermore, the G2/M phase cell cycle population, cisplatin-sensitivity, and cell proliferation and migration ability were increased in the TGF-β1-treated group. These features were unaltered in the MET/return group when compared to the TGF-β1-treated group. Immunoblotting revealed an increase in the levels of SMAD3, phosphorylated SMAD3, phosphorylated extracellular signal-regulated kinase and caspase-3, and a decrease in active caspase-3 levels in the TGF-β1-treated group. Increased caspase-3 and reduced active caspase-3 levels were observed in the MET/return group, similar to those in the TGF-β1-treated group; however, levels of other signalling proteins were unchanged compared with the control group. EMT induced by TGF-β1 was not preserved; however, stemness-associated properties (CD24 expression, caspase-3 expression, cell proliferation and cisplatin-resistance) were sustained following removal of TGF-β1.

  1. Evaluation of the cracking by stress corrosion in nuclear reactor environments type BWR; Evaluacion del agrietamiento por corrosion bajo esfuerzo en ambientes de reactores nucleares tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Arganis J, C. R.

    2010-07-01

    The stress corrosion cracking susceptibility was studied in sensitized, solution annealed 304 steel, and in 304-L welded with a heat treatment that simulated the radiation induced segregation, by the slow strain rate test technique, in a similar environment of a boiling water reactor (BWR), 288 C, 8 MPa, low conductivity and a electrochemical corrosion potential near 200 mV. vs. standard hydrogen electrode (She). The electrochemical noise technique was used for the detection of the initiation and propagation of the cracking. The steels were characterized by metallographic studies with optical and scanning electronic microscopy and by the electrochemical potentiodynamic reactivation of single loop and double loop. In all the cases, the steels present delta ferrite. The slow strain rate tests showed that the 304 steel in the solution annealed condition is susceptible to transgranular stress corrosion cracking (TGSCC), such as in a normalized condition showed granulated. In the sensitized condition the steel showed intergranular stress corrosion cracking, followed by a transition to TGSCC. The electrochemical noise time series showed that is possible associated different time sequences to different modes of cracking and that is possible detect sequentially cracking events, it is means, one after other, supported by the fractographic studies by scanning electron microscopy. The parameter that can distinguish between the different modes of cracking is the re passivation rate, obtained by the current decay rate -n- in the current transients. This is due that the re passivation rate is a function of the microstructure and the sensitization. Other statistic parameters like the localized index, Kurtosis, Skew, produce results that are related with mixed corrosion. (Author)

  2. 3D simulation of a core operation cycle of a BWR using Serpent; Simulacion 3D de un ciclo de operacion del nucleo de un BWR usando SERPENT

    Energy Technology Data Exchange (ETDEWEB)

    Barrera Ch, M. A.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. Lindavista, 07738 Ciudad de Mexico (Mexico); Gomez T, A. M., E-mail: rionchez@icloud.com [ININ, Departamento de Sistemas Nucleares, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2016-09-15

    This work had the main goal to develop a methodology to obtain the length of an operating cycle of the core of a BWR under different operating states using the Serpent code. The reactor core modeled in Serpent is composed of 444 fuel assemblies (120 with fresh fuels and 324 fuels from previous cycles), 109 cruciform control rods and light water as moderator and coolant. Once the core of the reactor was modeled in Serpent (Three-dimensional) without considering the cruciform control rods, a simulation was carried out with different steps of burning in the operational state with the average values of the fuel temperature (900 K), moderator temperature (600 K) and voids fraction equal to 0.4. In addition, the thermal power considered was 2017 MWt. This operational state was chosen because a previous analysis (not shown in this work) was carried out in 4 types of control cells. The first and second control cell has all of its natural uranium fuel pellets, with control rod and without control rod respectively. The third and fourth control cell types have various types of enrichment, both natural uranium and gadolinium in their fuel pellets, with control rod and without control rod. The conclusion of this previous analysis was that the behavior of the effective multiplication factor along the fuel burnout within the four control cell types was almost unaffected by the fuel temperature but was affected by the voids fraction. Thus, for this operating cycle in the operating state defined above, its length was 14,63052 GW t/Tm. In addition, at the end of this cycle, the decay heat obtained was equal to 116.71 MWt and the inventory of the most important isotopes to be considered was obtained, such as some isotopes of uranium, neptune, plutonium, americium and curio. (Author)

  3. Verification of the Advanced Nodal Method on BWR Core Analyses by Whole-Core Heterogeneous Transport Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Shinya Kosaka

    2000-11-12

    Recent boiling water reactor (BWR) core and fuel designs have become more sophisticated and heterogeneous to improve fuel cycle cost, thermal margin, etc. These improvements, however, tend to lead to a strong interference effect among fuel assemblies, and it my cause some inaccuracies in the BWR core analyses by advanced nodal codes. Furthermore, the introduction of mixed-oxide (MOX) fuel will lead to a much stronger interference effect between MOX and UO{sub 2} fuel assemblies. However, the CHAPLET multiassembly characteristics transport code was developed recently to solve two-dimensional cell-heterogeneous whole-core problems efficiently, and its results can be used as reference whole-core solutions to verify the accuracy of nodal core calculations. In this paper, the results of nodal core calculations were compared with their reference whole-core transport solutions to verify their accuracy (in k{sub eff}, assembly power and pin power via pin power reconstruction) of the advanced nodal method on both UO{sub 2} and MOX BWR whole-core analyses. Especially, it was investigated if there were any significant differences in the accuracy between MOX and UO{sub 2} results.

  4. Impact of modeling Choices on Inventory and In-Cask Criticality Calculations for Forsmark 3 BWR Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Martinez-Gonzalez, Jesus S. [Univ. Politecnica de Madrid (Spain); Ade, Brian J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bowman, Stephen M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauld, Ian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ilas, Germina [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Marshall, William BJ J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    Simulation of boiling water reactor (BWR) fuel depletion poses a challenge for nuclide inventory validation and nuclear criticality safety analyses. This challenge is due to the complex operating conditions and assembly design heterogeneities that characterize these nuclear systems. Fuel depletion simulations and in-cask criticality calculations are affected by (1) completeness of design information, (2) variability of operating conditions needed for modeling purposes, and (3) possible modeling choices. These effects must be identified, quantified, and ranked according to their significance. This paper presents an investigation of BWR fuel depletion using a complete set of actual design specifications and detailed operational data available for five operating cycles of the Swedish BWR Forsmark 3 reactor. The data includes detailed axial profiles of power, burnup, and void fraction in a very fine temporal mesh for a GE14 (10×10) fuel assembly. The specifications of this case can be used to assess the impacts of different modeling choices on inventory prediction and in-cask criticality, specifically regarding the key parameters that drive inventory and reactivity throughout fuel burnup. This study focused on the effects of the fidelity with which power history and void fraction distributions are modeled. The corresponding sensitivity of the reactivity in storage configurations is assessed, and the impacts of modeling choices on decay heat and inventory are addressed.

  5. Development of a fuzzy logic method to build objective functions in optimization problems: application to BWR fuel lattice design

    Energy Technology Data Exchange (ETDEWEB)

    Martin-del-Campo, C.; Francois, J.L.; Barragan, A.M. [Universidad Nacional Autonoma de Mexico - Facultad de Ingenieria (Mexico); Palomera, M.A. [Universidad Nacional Autonoma de Mexico - Instituto de Investigaciones en Matematicas Aplicadas y Sistema, Mexico, D. F. (Mexico)

    2005-07-01

    In this paper we develop a methodology based on the use of the Fuzzy Logic technique to build multi-objective functions to be used in optimization processes applied to in-core nuclear fuel management. As an example, we selected the problem of determining optimal radial fuel enrichment and gadolinia distributions in a typical 'Boiling Water Reactor (BWR)' fuel lattice. The methodology is based on the use of the mathematical capability of Fuzzy Logic to model nonlinear functions of arbitrary complexity. The utility of Fuzzy Logic is to map an input space into an output space, and the primary mechanism for doing this is a list of if-then statements called rules. The rules refer to variables and adjectives that describe those variables and, the Fuzzy Logic technique interprets the values in the input vectors and, based on the set of rules assigns values to the output vector. The methodology was developed for the radial optimization of a BWR lattice where the optimization algorithm employed is Tabu Search. The global objective is to find the optimal distribution of enrichments and burnable poison concentrations in a 10*10 BWR lattice. In order to do that, a fuzzy control inference system was developed using the Fuzzy Logic Toolbox of Matlab and it has been linked to the Tabu Search optimization process. Results show that Tabu Search combined with Fuzzy Logic performs very well, obtaining lattices with optimal fuel utilization. (authors)

  6. Increased levels of peroxisome proliferator-activated receptor gamma, coactivator 1 alpha (PGC-1alpha) improve lipid utilisation, insulin signalling and glucose transport in skeletal muscle of lean and insulin-resistant obese Zucker rats.

    Science.gov (United States)

    Benton, C R; Holloway, G P; Han, X-X; Yoshida, Y; Snook, L A; Lally, J; Glatz, J F C; Luiken, J J F P; Chabowski, A; Bonen, A

    2010-09-01

    Reductions in peroxisome proliferator-activated receptor gamma, coactivator 1 alpha (PGC-1alpha) levels have been associated with the skeletal muscle insulin resistance. However, in vivo, the therapeutic potential of PGC-1alpha has met with failure, as supra-physiological overexpression of PGC-1alpha induced insulin resistance, due to fatty acid translocase (FAT)-mediated lipid accumulation. Based on physiological and metabolic considerations, we hypothesised that a modest increase in PGC-1alpha levels would limit FAT upregulation and improve lipid metabolism and insulin sensitivity, although these effects may differ in lean and insulin-resistant muscle. Pgc-1alpha was transfected into lean and obese Zucker rat muscles. Two weeks later we examined mitochondrial biogenesis, intramuscular lipids (triacylglycerol, diacylglycerol, ceramide), GLUT4 and FAT levels, insulin-stimulated glucose transport and signalling protein phosphorylation (thymoma viral proto-oncogene 2 [Akt2], Akt substrate of 160 kDa [AS160]), and fatty acid oxidation in subsarcolemmal and intermyofibrillar mitochondria. Electrotransfection yielded physiologically relevant increases in Pgc-1alpha (also known as Ppargc1a) mRNA and protein ( approximately 25%) in lean and obese muscle. This induced mitochondrial biogenesis, and increased FAT and GLUT4 levels, insulin-stimulated glucose transport, and Akt2 and AS160 phosphorylation in lean and obese animals, while bioactive intramuscular lipids were only reduced in obese muscle. Concurrently, PGC-1alpha increased palmitate oxidation in subsarcolemmal, but not in intermyofibrillar mitochondria, in both groups. In obese compared with lean animals, the PGC-1alpha-induced improvement in insulin-stimulated glucose transport was smaller, but intramuscular lipid reduction was greater. Increases in PGC-1alpha levels, similar to those that can be induced by physiological stimuli, altered intramuscular lipids and improved fatty acid oxidation, insulin signalling

  7. Proliferation Vulnerability Red Team report

    Energy Technology Data Exchange (ETDEWEB)

    Hinton, J.P.; Barnard, R.W.; Bennett, D.E. [and others

    1996-10-01

    This report is the product of a four-month independent technical assessment of potential proliferation vulnerabilities associated with the plutonium disposition alternatives currently under review by DOE/MD. The scope of this MD-chartered/Sandia-led study was limited to technical considerations that could reduce proliferation resistance during various stages of the disposition processes below the Stored Weapon/Spent Fuel standards. Both overt and covert threats from host nation and unauthorized parties were considered. The results of this study will be integrated with complementary work by others into an overall Nonproliferation and Arms Control Assessment in support of a Secretarial Record of Decision later this year for disposition of surplus U.S. weapons plutonium.

  8. Technical Basis for Peak Reactivity Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, William BJ J [ORNL; Ade, Brian J [ORNL; Bowman, Stephen M [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL; Mertyurek, Ugur [ORNL; Radulescu, Georgeta [ORNL

    2015-01-01

    Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate application of burnup credit for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase (1) investigates applicability of peak reactivity methods currently used in spent fuel pools (SFPs) to storage and transportation systems and (2) evaluates validation of both reactivity (keff) calculations and burnup credit nuclide concentrations within these methods. The second phase will focus on extending burnup credit beyond peak reactivity. This paper documents the first phase, including an analysis of lattice design parameters and depletion effects, as well as both validation components. Initial efforts related to extended burnup credit are discussed in a companion paper. Peak reactivity analyses have been used in criticality analyses for licensing of BWR fuel in SFPs over the last 20 years. These analyses typically combine credit for the gadolinium burnable absorber present in the fuel with a modest amount of burnup credit. Gadolinium burnable absorbers are used in BWR assemblies to control core reactivity. The burnable absorber significantly reduces assembly reactivity at beginning of life, potentially leading to significant increases in assembly reactivity for burnups less than 15–20 GWd/MTU. The reactivity of each fuel lattice is dependent on gadolinium loading. The number of gadolinium-bearing fuel pins lowers initial lattice reactivity, but it has a small impact on the burnup and reactivity of the peak. The gadolinium concentration in each pin has a small impact on initial lattice reactivity but a significant effect on the reactivity of the peak and the burnup at which the peak occurs. The importance of the lattice parameters and depletion conditions are primarily determined by their impact on the gadolinium depletion. Criticality code validation for BWR burnup

  9. Analyses of Instability Events in the Peach Bottom-2 BWR Using Thermal-Hydraulic and 3D Neutron Kinetic Coupled Codes Technique

    Directory of Open Access Journals (Sweden)

    Antonella Lombardi Costa

    2008-01-01

    Full Text Available Boiling water reactor (BWR instabilities may occur when, starting from a stable operating condition, changes in system parameters bring the reactor towards an unstable region. In order to design more stable and safer core configurations, experimental and theoretical studies about BWR stability have been performed to characterise the phenomenon and to predict the conditions for its occurrence. In this work, contributions to the study of BWR instability phenomena are presented. The RELAP5/MOD3.3 thermal-hydraulic (TH system code and the PARCS-2.4 3D neutron kinetic (NK code were coupled to simulate BWR transients. Different algorithms were used to calculate the decay ratio (DR and the natural frequency (NF from the power oscillation predicted by the transient calculations as two typical parameters used to provide a quantitative description of instabilities. The validation of the code model set up for the Peach Bottom Unit 2 BWR plant is performed against low-flow stability tests (LFSTs. The four series of LFST have been performed during the first quarter of 1977 at the end of cycle 2 in Pennsylvania. The tests were intended to measure the reactor core stability margins at the limiting conditions used in design and safety analyses.

  10. Experimental investigation of mixing of non-isothermal water streams at BWR operating conditions

    Energy Technology Data Exchange (ETDEWEB)

    Bergagio, Mattia, E-mail: bergagio@kth.se [AlbaNova University Center, Nuclear Reactor Technology Division, Department of Physics, Royal Institute of Technology, 106 91 Stockholm (Sweden); Anglart, Henryk, E-mail: henryk@kth.se [AlbaNova University Center, Nuclear Reactor Technology Division, Department of Physics, Royal Institute of Technology, 106 91 Stockholm (Sweden); Institute of Heat Engineering, Warsaw University of Technology, 21/25 Nowowiejska Street, 00-665 Warsaw (Poland)

    2017-06-15

    Highlights: • Temperatures are measured in the presence of mixing at BWR operating conditions. • The thermocouple support is moved along a pattern to extend the measurement region. • Uncertainty of 1.58 K for temperatures acquired at 1000 Hz. • Momenta of the hot streams and thermal stratification affect the data examined. • Unconventional spectral analysis is required to further study the data collected. - Abstract: In this experimental investigation, wall surface temperatures have been measured during mixing of three water streams in the annular gap between two coaxial stainless-steel tubes. The inner tube, with an outer diameter of 35 mm and a thickness of 5 mm, holds six K-type, ungrounded thermocouples with a diameter of 0.5 mm, which measured surface temperatures with a sampling rate of either 100 Hz or 1000 Hz. The tube was rotated from 0 to 360° and moved in a range of 387 mm in the axial direction to allow measurements of surface temperatures in the whole mixing region. The outer tube has an inner diameter of 80 mm and a thickness of 10 mm to withstand a water pressure of 9 MPa. A water stream at a temperature of either 333 K or 423 K and a Reynolds number between 1657 and 8410 rose vertically in the annular gap and mixed with two water streams at a temperature of 549 K and a Reynolds number between 3.56 × 10{sup 5} and 7.11 × 10{sup 5}. These two water streams entered the annulus radially on the same axial level, 180° apart. Water pressure was kept at 7.2 MPa. Temperature recordings were performed at five axial and eight azimuthal locations, for each set of boundary conditions. Each recording lasted 120 s to provide reliable data on the variance, intermittency and frequency of the surface temperature time series at hand. Thorough calculations indicate that the uncertainty in the measured temperature is of 1.58 K. The mixing region extends up to 0.2 m downward of the hot inlets. In most cases, measurements indicate non-uniform mixing in the

  11. Cracks propagation by stress corrosion cracking in conditions of Boiling Water Reactor (BWR); Propagacion de grietas por corrosion bajo esfuerzo en condiciones de reactor de agua hirviente (BWR)

    Energy Technology Data Exchange (ETDEWEB)

    Fuentes C, P

    2003-07-01

    This work presents the results of the assays carried out in the Laboratory of Hot Cells of the National Institute of Nuclear Research (ININ) to a type test tube Compact Tension (CT), built in steel austenitic stainless type 304L, simulating those conditions those that it operates a Boiling Water Reactor (BWR), at temperature 288 C and pressure of 8 MPa, to determine the speed to which the cracks spread in this material that is of the one that different components of a reactor are made, among those that it highlights the reactor core vessel. The application of the Hydrogen Chemistry of the Water is presented (HWC) that is one alternative to diminish the corrosion effect low stress in the component, this is gets controlling the quantity of oxygen and of hydrogen as well as the conductivity of the water. The rehearsal is made following the principles of the Mechanics of Elastic Lineal Fracture (LEFM) that considers a crack of defined size with little plastic deformation in the tip of this; the measurement of crack advance is continued with the technique of potential drop of direct current of alternating signal, this is contained inside the standard Astm E-647 (Method of Test Standard for the Measurement of Speed of Growth of Crack by fatigue) that is the one that indicates us as carrying out this test. The specifications that should complete the test tubes that are rehearsed as for their dimensions, it forms, finish and determination of mechanical properties (tenacity to the fracture mainly) they are contained inside the norm Astm E-399, the one which it is also based on the principles of the fracture mechanics. The obtained results were part of a database to be compared with those of other rehearsals under different conditions, Normal Chemistry of the Water (NWC) and it dilutes with high content of O{sub 2}; to determine the conditions that slow more the phenomena of stress corrosion cracking, as well as the effectiveness of the used chemistry and of the method of

  12. Radial optimization of a BWR fuel cell using genetic algorithms; Optimizacion radial de una celda de combustible BWR usando algoritmos geneticos

    Energy Technology Data Exchange (ETDEWEB)

    Martin del Campo M, C.; Carmona H, R.; Oropeza C, I.P. [UNAM, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)]. e-mail: cmcm@fi-b.unam.mx

    2006-07-01

    The development of the application of the Genetic Algorithms (GA) to the optimization of the radial distribution of enrichment in a cell of fuel of a BWR (Boiling Water Reactor) is presented. The optimization process it was ties to the HELIOS simulator, which is a transport code of neutron simulation of fuel cells that has been validated for the calculation of nuclear banks for BWRs. With heterogeneous radial designs can improve the radial distribution of the power, for what the radial design of fuel has a strong influence in the global design of fuel recharges. The optimum radial distribution of fuel bars is looked for with different enrichments of U{sup 235} and contents of consumable poison. For it is necessary to define the representation of the solution, the objective function and the implementation of the specific optimization process to the solution of the problem. The optimization process it was coded in 'C' language, it was automated the creation of the entrances to the simulator, the execution of the simulator and the extraction, in the exit of the simulator, of the parameters that intervene in the objective function. The objective function includes four parameters: average enrichment of the cell, average gadolinia concentration of the cell, peak factor of radial power and k-infinite multiplication factor. To be able to calculate the parameters that intervene in the objective function, the one evaluation process of GA was ties to the HELIOS code executed in a Compaq Alpha workstation. It was applied to the design of a fuel cell of 10 x 10 that it can be employee in the fuel assemble designs that are used at the moment in the Laguna Verde Nucleo electric Central. Its were considered 10 different fuel compositions which four contain gadolinia. Three heuristic rules that consist in prohibiting the placement of bars with gadolinia in the ends of the cell, to place the compositions with the smallest enrichment in the corners of the cell and to fix

  13. Effects of axial higher mode on core stability of natural circulation BWR

    Energy Technology Data Exchange (ETDEWEB)

    Inada, F.; Furuya, M.; Yasuo, A. [Central Research Institute of Electric Power Industry, Tokyo (Japan)

    2003-07-01

    The effect of the chimney on the core stability of natural circulation BWR was investigated using linear stability analysis. The point kinetics model was used for estimating void-reactivity feedback. Drift-flux model was used to evaluate boiling two-phase flow. When chimney was short, the 1st mode was dominant in the case of low power as well as high power. Higher inlet subcooling and higher power could lead to destabilizing effect, and in some case core instability could occur. On the other hand, when chimney was long, it was found that higher harmonics of void fraction perturbation in the chimney could be important. In the case of low power, the 1st mode was less stable, and lower power could lead to the destabilizing effect. In the case of higher power, the higher mode was dominant rather than the 1st mode, and higher inlet subcooling could lead to destabilizing effect, but it was still stable. Sensitivity of stability in the case of long chimney was smaller than that of short chimney, and the instability phenomenon was not generated easily in the case of the long chimney.

  14. Potential for containment leak paths through electrical penetration assemblies under severe accident conditions. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Sebrell, W.

    1983-07-01

    The leakage behavior of containments beyond design conditions and knowledge of failure modes is required for evaluation of mitigation strategies for severe accidents, risk studies, emergency preparedness planning, and siting. These studies are directed towards assessing the risk and consequences of severe accidents. An accident sequence analysis conducted on a Boiling Water Reactor (BWR), Mark I (MK I), indicated very high temperatures in the dry-well region, which is the location of the majority of electrical penetration assemblies. Because of the high temperatures, it was postulated in the ORNL study that the sealants would fail and all the electrical penetration assemblies would leak before structural failure would occur. Since other containments had similar electrical penetration assemblies, it was concluded that all containments would experience the same type of failure. The results of this study, however, show that this conclusion does not hold for PWRs because in the worst accident sequence, the long time containment gases stabilize to 350/sup 0/F. BWRs, on the other hand, do experience high dry-well temperatures and have a higher potential for leakage.

  15. AREVA NP burnup credit investigation on irradiated MOX fuel within the REBUS BWR programme

    Energy Technology Data Exchange (ETDEWEB)

    Alander, Alexandra; Misu, Stefan; Timm, Wolf [AREVA, AREVA NP, Erlangen (Germany); Thareau, Sebastien [AREVA, AREVA NP, Paris (France)

    2008-07-01

    The present paper summarizes a criticality and burnup credit investigation carried out using the 2D spectral codes CASMO-4 and APOLLO2-A. Fission rate distributions and multiplication factors, for UOX and MOX configurations, are calculated as well as the reactivity effect caused by burnup on a selection of irradiated MOX fuel assemblies. 3D core criticality calculations were carried out with the Monte Carlo transport code MOCA and the deterministic transport code VARIANT (a nodal code developed by ANL) using CASMO-4 generated cross section libraries. Calculations were compared to experimental data from the critical facility VENUS in the context of the REBUS BWR Programme. The results confirm that the spectral codes CASMO-4 and APOLLO2-A are well suited to calculate fission rates, multiplication factors and reactivity effects. It is also found that the calculated burnup reactivity effect, using CASMO-4 generated cross sections and the best 3D method (MOCA), is underestimated by merely 7% compared to the experimental value, which can mainly be attributed to the simplifications done in order to model the critical configurations with reasonable efforts. (authors)

  16. THERMAL EVALUATION OF THE CONCEPTUAL 24 BWR UCF TUBE BASKET DESIGN DISPOSAL CONTAINER

    Energy Technology Data Exchange (ETDEWEB)

    T.L. Lotz

    1995-12-18

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) as specified in the Waste Package Implementation Plan (pp. 4-8,4-11,4-24,5-1, and 5-13; Ref. 5.10) and Waste Package Plan (pp. 3-15,3-17, and 3-24; Ref. 5.9). The design data request addressed herein is: (1) Characterize the conceptual 24 boiling water reactor (BWR) uncanistered fuel (UCF) waste package (WP) to show that the design is feasible for use in the MGDS environment. The purpose of this analysis is to respond to a concern that the long-term disposal thermal issues for the UCF waste package do not preclude UCF waste package compatibility with the MGDS. The objective of this analysis is to provide thermal parameter information for the conceptual UCF WP design under nominal MGDS repository conditions. The results are intended to show that the design has a reasonable chance to meet the MGDS design requirements for normal MGDS operation and to provide the required guidance to determining the major design issues for future design efforts. Future design efforts will focus on UCF design changes as further design and operations information becomes available.

  17. Analysis of fission product transport under terminated LOCA conditions. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Gieseke, J.A.; Baybutt, P.; Jordan, H.; Denning, R.S.; Wooton, R.O.

    1977-12-30

    An analytical model was developed to allow preditions of the source term to the containment as dependent on release from the fuel pins and deposition within the primary system. The model was developed into a flexible computer code adaptable to various geometrical arrangements and flow paths. The calculational framework was established in such a way to permit, in principle, the determination of particle and vapor transport and deposition in a general system of control volumes connected by fluid flow in an arbitrary way. This framework requires as input the rate of fission product release to the primary flow, as a function of time for each vapor and particulate species to be considered, and a complete, time dependent, thermal-hydraulic description of the system. TRAP-PWR, TRAP-BWR and TRACK computer codes are specializations, described in detail in the text of this report, of the general framework. The nature of these specializations depends strongly on the degree of detail with which the transporting fluid medium is modeled.

  18. The integrity of NSSS and containment during extended station blackout for Kuosheng BWR plant

    Energy Technology Data Exchange (ETDEWEB)

    Hsu, Keng-Hsien; Yuann, Yng-Ruey; Lin, Ansheng [Atomic Energy Council, Taoyuan City, Taiwan (China). Inst. of Nuclear Energy Research

    2017-11-15

    The Fukushima Daiichi accident occurring on March 11, 2011, reveals that Station Blackout (SBO) may last longer than 8 h. However, the original design may not have sufficient capacity to cope with a SBO for more than 8 h. In view of this, Taiwan Power Company has initiated several enhancements to mitigate the severity of the extended SBO. Based on the improved plant configuration, a SBO coping analysis is performed in this study to assess whether the Kuosheng BWR plant has sufficient capability to cope with SBO for 24 h with respect to maintaining the integrity of the reactor core and containment. The analyses in the Nuclear Steam Supply System (NSSS) and the containment are based on the RETRAN-3D and GOTHIC models, respectively. The flow conditions calculated by RETRAN-3D during the event are retrieved and input to the GOTHIC containment model to determine the containment pressure and temperature response. These boundary conditions include SRV flow rate, SRV flow enthalpy, and total reactor coolant system leakage flow rate.

  19. Standard technical specifications General Electric plants, BWR/6. Volume 1, Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-04-01

    This report documents the results of the combined effort of the NRC and the industry to produce improved Standard Technical Specifications (STS), Revision 1 for General Electric BWR/6 Plants. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved STS or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS.

  20. Standard technical specifications: General Electric plants, BWR/4. Volume 1, Revision 1: Specifications

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-04-01

    This report documents the results of the combined effort of the NRC and the industry to produce improved Standard Technical Specifications (STS), Revision 1 for General Electric BWR/4 Plants. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved STS or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS.

  1. HORIZONTAL LIFTING OF 5 DHLW/DOE LONG, 12-PWR LONG AND 24-BWR WASTE PACKAGES

    Energy Technology Data Exchange (ETDEWEB)

    V. de la Brosse

    2001-05-17

    The objective of this calculation was to determine the structural response of a 12-Pressurized Water Reactor (PWR) Long, a 24-Boiling Water Reactor (BWR) and a 5-Defense High Level Waste/Department of Energy (DHLW/DOE)--Long spent nuclear fuel waste packages lifted in a horizontal position. The scope of this calculation was limited to reporting the calculation results in terms of maximum stress intensities in the trunnion collar sleeves. In addition, the maximum stress intensities in the inner and outer shells of the waste packages were presented for illustrative purposes. The information provided by the sketches (Attachments I, II and III) is that of the potential design of the types of waste packages considered in this calculation, and all obtained results are valid for these designs only. This calculation is associated with the waste package design and was performed by the Waste Package Design Section in accordance with the ''Technical work plan for: Waste Package Design Description for LA'' (Ref. 7). AP-3.12Q, Calculations (Ref. 13), was used to perform the calculation and develop the document.

  2. 3D analysis methods - Study and seminar[BWR safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Daaviittila, A. [Valtion Teknillinen Tutkimuskeskus (Finland)

    2003-10-01

    The first part of the report results from a study that was performed as a Nordic co-operation activity with active participation from Studsvik Scandpower and Westinghouse Atom in Sweden, and VTT in Finland. The purpose of the study was to identify and investigate the effects rising from using the 3D transient com-puter codes in BWR safety analysis, and their influence on the transient analysis methodology. One of the main questions involves the critical power ratio (CPR) calculation methodology. The present way, where the CPR calculation is per-formed with a separate hot channel calculation, can be artificially conservative. In the investigated cases, no dramatic minimum CPR effect coming from the 3D calculation is apparent. Some cases show some decrease in the transient change of minimum CPR with the 3D calculation, which confirms the general thinking that the 1D calculation is conservative. On the other hand, the observed effect on neutron flux behaviour is quite large. In a slower transient the 3D effect might be stronger. The second part of the report is a summary of a related seminar that was held on the 3D analysis methods. The seminar was sponsored by the Reactor Safety part (NKS-R) of the Nordic Nuclear Safety Research Programme (NKS). (au)

  3. The Physical Mechanism of Core-Wide and Local Instabilities at the Forsmark-1 BWR

    Energy Technology Data Exchange (ETDEWEB)

    Analytis, G. Th. [Paul Scherrer Inst., CH-5232 Villigen PSI (Switzerland). Lab. for Thermohydraulics; Hennig, D. [Paul Scherrer Inst., CH-5232 Villigen PSI (Switzerland). Lab. for Reactor Physics and Systems Engineering; Karlsson, J. K.-H. [Chalmers University of Technology, Gothenburg (Sweden)

    1998-10-01

    During the last 15 years, the problem of BWR instabilities has attracted the attention of a number of researchers. From the theoretical point of view, one would be interested in physically understanding the mechanisms responsible for the in- and out-of-phase core wide power oscillations observed at certain operating points of the power-flow map in different BWRs. From the practical point of view, one must try to avoid these `incidents` since either locally, or globally, the power may substantially exceed the prescribed levels. In this work, we shall use RAMONA3-12 and analyse a rather unusual instability incident at Forsmark-1 in which in addition to the core-wide fundamental spatial mode oscillation, there were local large amplitude power oscillations at different radial positions in the core. We were able to reproduce these unusual experimental findings by assuming that there are large amplitude Density Wave Oscillations (DWOs) in different bundles, induced by the fact that these bundles were not seated properly into the lower fuel support plate. (author) 17 refs., 12 figs.

  4. Study on nuclear physics of high burn-up full MOX-BWR core

    Energy Technology Data Exchange (ETDEWEB)

    Shirakawa, Toshihisa; Okubo, Tsutomu; Ochiai, Masa-aki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-08-01

    In this report, neutronics study of full Mixed-oxide (MOX) high burn-up BWR core is presented. Our design goals are about 3-year cycle length, four-batch refueling scheme and more than 100GWd/t fuel discharge burn-up. Base core configuration is 1,350MWe US version of ABWR with 9 x 9 type fuel assembly. Investigation of the reactor core has been carried out by arranging Gd{sub 2}O{sub 3} contents in fuel rods and changing water to fuel volume ratio (V{sub m}/V{sub f}) through the number of water rods or adjustment of fuel clad diameter. JAERI`s general purpose neutronics code system SRAC95 was used for two dimensional XY fuel assembly cell neutronics calculations. Calculated cases are for a comparatively high moderated fuel assembly with 9 water rods, a fuel assembly without water rods and a comparatively low moderated fuel assembly without water rods and with larger fuel clad diameter. All these 3 cases seem to achieve our design goals mentioned above. For the last case, three dimensional core burn-up calculation was performed by this code system. This case seems to attain a low linear power density and the operation with all control rod out. (author)

  5. Identification and assessment of containment and release management strategies for a BWR Mark I containment

    Energy Technology Data Exchange (ETDEWEB)

    Lin, C.C.; Lehner, J.R. (Brookhaven National Lab., Upton, NY (United States))

    1991-09-01

    This report identifies and assesses accident management strategies which could be important for preventing containment failure and/or mitigating the release of fission products during a severe accident in a BWR plant with a Mark 1 type of containment. Based on information available from probabilistic risk assessments and other existing severe accident research, and using simplified containment and release event trees, the report identifies the challenges a Mark 1 containment could face during the course of a severe accident, the mechanisms behind these challenges, and the strategies that could be used to mitigate the challenges. A safety objective tree is developed which provides the connection between the safety objectives, the safety functions, the challenges, and the strategies. The strategies were assessed by applying them to certain severe accident sequence categories which have one or more of the following characteristics: have high probability of core damage or high consequences, lead to a number of challenges, and involve the failure of multiple systems. 59 refs., 55 figs., 27 tabs.

  6. Experimental investigation of BWR Suppression Pool stratification during RCIC system operation

    Energy Technology Data Exchange (ETDEWEB)

    Solom, Matthew, E-mail: msolom@sandia.gov [Sandia National Laboratories, MS-0748, P.O. Box 5800, Albuquerque, NM 87185-0748 (United States); Vierow Kirkland, Karen [Department of Nuclear Engineering, Texas A& M University, MS 3133, College Station, TX 77843-3133 (United States)

    2016-12-15

    Highlights: • An experiment at Texas A&M University explored extended RCIC System operations. • Thermal stratification in Suppression Pools was found to develop and later disappear. • Greater containment pressure led to much greater vertical thermal stratification. - Abstract: In Boiling Water Reactor (BWR) nuclear power plants with the Mark I containment, the condition of the Suppression Pool can be a large influence on overall plant safety. When the Reactor Core Isolation Cooling (RCIC) System is operating, steam from the reactor drives the RCIC turbine and is then exhausted to the Suppression Pool. When subcooled, the pool can readily condense the steam, warming it up in the process. However, if hot spots or thermal stratification appear, this can limit the Suppression Pool’s ability to perform its safety functions, and can be a limiting factor for RCIC System operation. In order to better understand the RCIC system and its true limits of long-term operation, an experimental model of the system was constructed at the Laboratory for Nuclear Heat Transfer Systems at Texas A&M University (TAMU). These tests provide confirmation of thermal stratification in the Suppression Pool from RCIC System operations, and show a significant degree of dependence on pressure in the airspace above the pool. In the TAMU facility, vertical thermal stratification was limited to 21 °C when fully vented to atmospheric pressure, while pre-pressurization led to stratification well in excess of 60 °C.

  7. Genusa Bepu methodologies for the safety analysis of BWRs; Metodologias Bepu de Genusa para el analisis de seguridad de reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Trueba, M.; Garcia, J.; Goodson, C.; Ibarra, L.

    2016-08-01

    This article describes the BEPU methodologies developed by General Electric-Hitachi (GEH) for the evaluation of the BWR reactor safety analysis based on the TRACG best-estimate code. These methodologies are applicable to a wide range of events, operational transients (AOO), anticipated transients without scram (ATWS), loss of coolant accidents (LOCA) and instability events; to different BWR types operating commercially. General Electric (GE( designs and other vendors, including Generation III+ESBWR; to the new operation strategies, and to all types of BWR fuel. Their application achieves, among other benefits, a better understanding of the overall plant response and an improvement in margins to the operating limits; thus, the increase of flexibility in reactor operation and reduction in generation costs. (Author)

  8. CFD Simulation of rigid venting of the containment of a BWR-5 Mark-II reactor; Simulacion CFD de los venteos rigidos de la contencion de un reactor BWR-5 Mark-II

    Energy Technology Data Exchange (ETDEWEB)

    Galindo G, I. F.; Vazquez B, A. K.; Velazquez E, L. [Instituto Nacional de Electricidad y Energias Limpias, Reforma 113, Col. Palmira, 62490 Cuernavaca, Morelos (Mexico); Tijerina S, F.; Tapia M, R., E-mail: francisco.tijerina@cfe.gob.mx [CFE, Central Nucleoelectrica Laguna Verde, Carretera Federal Cardel-Nautla Km 42.5, 91476 Municipio Alto Lucero, Veracruz (Mexico)

    2016-09-15

    In conditions of prolonged loss of external energy or a severe accident, venting to the atmosphere is an alternative to prevent overpressure and release of fission products from the primary containment of a nuclear reactor. Due to the importance of flow determination through rigid vents, a computational fluid dynamics (CFD) model is proposed to verify the capacity of rigid vents in the primary containment of a boiling water reactor (BWR) under different operating conditions (pressure, temperature and compositions of the fluids). The model predicts and provides detailed information on variables such as mass flow and velocity of the venting gases. In the proposed model the primary containment gas is vented to the atmosphere via rigid vents (pipes) from the dry and wet pit. Is assumed that the container is pressurized because is in a defined scenario, and at one point the venting is open and the gas released into the atmosphere. The objective is to characterize the flow and validate the CFD model for the overpressure conditions that occur in an accident such as a LOCA, Sbo, etc. The model is implemented with Ansys-Fluent general-purpose CFD software based on the geometry of the venting ducts of the containment of a BWR. The model is developed three-dimensional and resolves at steady state for compressible flow and includes the effects of the turbulence represented by the Reynolds stress model. The CFD results are compared with the values of a one-dimensional and isentropic model for compressible flow. The relative similarity of results leads to the conclusion that the proposed CFD model can help to predict the rigid venting capacity of the containment of a BWR, however more information is required for full validation of the proposed model. (Author)

  9. Benchmark calculation for radioactivity inventory using MAXS library based on JENDL-4.0 and JEFF-3.0/A for decommissioning BWR plants

    Directory of Open Access Journals (Sweden)

    Tanaka Ken-ichi

    2016-01-01

    Full Text Available We performed benchmark calculation for radioactivity activated in a Primary Containment Vessel (PCV of a Boiling Water Reactor (BWR by using MAXS library, which was developed by collapsing with neutron energy spectra in the PCV of the BWR. Radioactivities due to neutron irradiation were measured by using activation foil detector of Gold (Au and Nickel (Ni at thirty locations in the PCV. We performed activation calculations of the foils with SCALE5.1/ORIGEN-S code with irradiation conditions of each foil location as the benchmark calculation. We compared calculations and measurements to estimate an effectiveness of MAXS library.

  10. miR-19b enhances proliferation and apoptosis resistance via the EGFR signaling pathway by targeting PP2A and BIM in non-small cell lung cancer.

    Science.gov (United States)

    Baumgartner, Ulrich; Berger, Fabienne; Hashemi Gheinani, Ali; Burgener, Sabrina Sofia; Monastyrskaya, Katia; Vassella, Erik

    2018-02-19

    Epidermal growth factor receptor (EGFR) mutations enable constitutive active downstream signaling of PI3K/AKT, KRAS/ERK and JAK/STAT pathways, and promote tumor progression by inducing uncontrolled proliferation, evasion of apoptosis and migration of non-small cell lung cancer (NSCLC). In addition, such EGFR mutations increase the susceptibility of patients with NSCLC to tyrosine kinase inhibitor (TKI) therapy, but treated patients will invariably relapse with resistant disease. A global understanding of underlying molecular mechanisms of EGFR signaling may improve the management of NSCLC patients. microarray analysis was performed to identify PI3K/AKT-regulated miRNAs. Phosphoproteomic analysis and cell based assays were performed using NSCLC cell lines lentivirally transduced with anti-miR or miR overexpressing constructs. Here, we show that 17 miRNAs including members of the miR-17~ 92 cluster are dysregulated following PI3K/AKT inhibition of EGFR mutant NSCLC cells. Bioinformatics analysis revealed that dysregulated miRNAs act in a concerted manner to enhance the activity of the EGFR signaling pathway. These findings were closely mirrored by attenuation of miR-17~ 92 family member miR-19b in NSCLC cell lines which resulted in reduced phosphorylation of ERK, AKT and STAT and effector proteins in EGFR mutant NSCLC cells. Consistent with this finding, cell cycle progression, clonogenic growth and migration were reduced and apoptosis was enhanced. Co-treatment of NSCLC cells with the tyrosine kinase inhibitor (TKI) gefitinib and anti-miR-19b construct reduced migration and clonogenic growth in a synergistic manner suggesting that EGFR and miR-19b act together to control oncogenic processes. Serine/threonine phosphatase PP2A subunit PPP2R5E and BCL2L11 encoding BIM were identified as major targets of miR-19b by target validation assays. Consistent with this finding, PP2A activity was strongly enhanced in NSCLC transduced with anti-miR-19b construct, but not in

  11. Changes in the cellular microRNA profile by the intracellular expression of HIV-1 Tat regulator: A potential mechanism for resistance to apoptosis and impaired proliferation in HIV-1 infected CD4+ T cells.

    Science.gov (United States)

    Sánchez-Del Cojo, María; López-Huertas, María Rosa; Díez-Fuertes, Francisco; Rodríguez-Mora, Sara; Bermejo, Mercedes; López-Campos, Guillermo; Mateos, Elena; Jiménez-Tormo, Laura; Gómez-Esquer, Francisco; Díaz-Gil, Gema; Alcamí, José; Coiras, Mayte

    2017-01-01

    HIV-1 induces changes in the miRNA expression profile of infected CD4+ T cells that could improve viral replication. HIV-1 regulator Tat modifies the cellular gene expression and has been appointed as an RNA silencing suppressor. Tat is a 101-residue protein codified by two exons that regulates the elongation of viral transcripts. The first exon of Tat (amino acids 1-72) forms the transcriptionally active protein Tat72, but the presence of the second exon (amino acids 73-101) results in a more competent regulatory protein (Tat101) with additional functions. Intracellular, full-length Tat101 induces functional and morphological changes in CD4+ T cells that contribute to HIV-1 pathogenesis such as delay in T-cell proliferation and protection against FasL-mediated apoptosis. But the precise mechanism by which Tat produces these changes remains unknown. We analyzed how the stable expression of intracellular Tat101 and Tat72 modified the miRNA expression profile in Jurkat cells and if this correlated with changes in apoptotic pathways and cell cycle observed in Tat-expressing cells. Specifically, the enhanced expression of hsa-miR-21 and hsa-miR-222 in Jurkat-Tat101 cells was associated with the reduced expression of target mRNAs encoding proteins related to apoptosis and cell cycle such as PTEN, PDCD4 and CDKN1B. We developed Jurkat cells with stable expression of hsa-miR-21 or hsa-miR-222 and observed a similar pattern to Jurkat-Tat101 in resistance to FasL-mediated apoptosis, cell cycle arrest in G2/M and altered cell morphology. Consequently, upregulation of hsa-miR-21 and hsa-miR-222 by Tat may contribute to protect against apoptosis and to anergy observed in HIV-infected CD4+ T cells.

  12. Nimesulide, a cyclooxygenase-2 selective inhibitor, suppresses obesity-related non-alcoholic fatty liver disease and hepatic insulin resistance through the regulation of peroxisome proliferator-activated receptor γ.

    Science.gov (United States)

    Tsujimoto, Shunsuke; Kishina, Manabu; Koda, Masahiko; Yamamoto, Yasutaka; Tanaka, Kohei; Harada, Yusuke; Yoshida, Akio; Hisatome, Ichiro

    2016-09-01

    Cyclooxygenase (COX)-2 selective inhibitors suppress non-alcoholic fatty liver disease (NAFLD); however, the precise mechanism of action remains unknown. The aim of this study was to examine how the COX-2 selective inhibitor nimesulide suppresses NAFLD in a murine model of high-fat diet (HFD)‑induced obesity. Mice were fed either a normal chow diet (NC), an HFD, or HFD plus nimesulide (HFD-nime) for 12 weeks. Body weight, hepatic COX-2 mRNA expression and triglyceride accumulation were significantly increased in the HFD group. Triglyceride accumulation was suppressed in the HFD-nime group. The mRNA expression of hepatic peroxisome proliferator-activated receptor γ (PPARγ) and the natural PPARγ agonist 15-deoxy-Δ12,14-prostaglandin J2 (15d‑PGJ2) were significantly increased in the HFD group and significantly suppressed in the HFD-nime group. Glucose metabolism was impaired in the HFD group compared with the NC group, and it was significantly improved in the HFD-nime group. In addition, the plasma insulin levels in the HFD group were increased compared with those in the NC group, and were decreased in the HFD-nime group. These results indicate that HFD-induced NAFLD is mediated by the increased hepatic expression of COX-2. We suggest that the production of 15d-PGJ2, which is mediated by COX-2, induces NAFLD and hepatic insulin resistance by activating PPARγ. Furthermore, the mRNA expression of tissue inhibitor of metalloproteinases-1 (TIMP‑1), procollagen-1 and monocyte chemoattractant protein-1 (MCP-1), as well as the number of F4/80-positive hepatic (Kupffer) cells, were significantly increased in the HFD group compared with the NC group, and they were reduced by nimesulide. In conclusion, COX-2 may emerge as a molecular target for preventing the development of NAFLD and insulin resistance in diet-related obesity.

  13. Deposition characteristics of iodine from the off-air of the nuclear auxiliaries of a BWR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Patzelt, A.

    1983-12-01

    Contrary to the present opinion, the penetrating iodine fraction in the off-air from the auxiliaries of a BWR reactor may be higher than the values communicated in the relevant literature. This seems to be a result of internal processes in the systems in which ion exchanger resins are used. In these systems, filters made of a combination of impregnated and activated carbon have proved to be a better solution as experiments have shown. Systems of this type are recommended for general used.

  14. Analysis of dose rates received around the storage pool for irradiated control rods in a BWR nuclear power plant.

    Science.gov (United States)

    Ródenas, J; Abarca, A; Gallardo, S

    2011-08-01

    BWR control rods are activated by neutron reactions in the reactor. The dose produced by this activity can affect workers in the area surrounding the storage pool, where activated rods are stored. Monte Carlo (MC) models for neutron activation and dose assessment around the storage pool have been developed and validated. In this work, the MC models are applied to verify the expected reduction of dose when the irradiated control rod is hanged in an inverted position into the pool. 2010 Elsevier Ltd. All rights reserved.

  15. Overview of New Tools to Perform Safety Analysis: BWR Station Black Out Test Case

    Energy Technology Data Exchange (ETDEWEB)

    D. Mandelli; C. Smith; T. Riley; J. Nielsen; J. Schroeder; C. Rabiti; A. Alfonsi; Cogliati; R. Kinoshita; V. Pasucci; B. Wang; D. Maljovec

    2014-06-01

    Dynamic Probabilistic Risk Assessment (DPRA) methodologies couple system simulator codes (e.g., RELAP, MELCOR) with simulation controller codes (e.g., RAVEN, ADAPT). While system simulator codes accurately model system dynamics deterministically, simulation controller codes introduce both deterministic (e.g., system control logic, operating procedures) and stochastic (e.g., component failures, parameter uncertainties) elements into the simulation. Typically, a DPRA is performed by: 1) sampling values of a set of parameters from the uncertainty space of interest (using the simulation controller codes), and 2) simulating the system behavior for that specific set of parameter values (using the system simulator codes). For complex systems, one of the major challenges in using DPRA methodologies is to analyze the large amount of information (i.e., large number of scenarios ) generated, where clustering techniques are typically employed to allow users to better organize and interpret the data. In this paper, we focus on the analysis of a nuclear simulation dataset that is part of the Risk Informed Safety Margin Characterization (RISMC) Boiling Water Reactor (BWR) station blackout (SBO) case study. We apply a software tool that provides the domain experts with an interactive analysis and visualization environment for understanding the structures of such high-dimensional nuclear simulation datasets. Our tool encodes traditional and topology-based clustering techniques, where the latter partitions the data points into clusters based on their uniform gradient flow behavior. We demonstrate through our case study that both types of clustering techniques complement each other in bringing enhanced structural understanding of the data.

  16. Noise analysis of Forsmark 1 data to investigate BWR core local instability

    Energy Technology Data Exchange (ETDEWEB)

    Oguma, R. [GSE Power Systems AB, Nykoeping (Sweden)

    1998-04-01

    BWR core local instability was experienced at Forsmark 1 (F1) during reactor operation in cycle 16. The event has been studied by applying noise analysis and stability calculations to get insight into the event as well as to identify the cause of local instability. The present report is concerned with noise analysis of data collected during start-up in cycle 17. The results of the current study indicates: The F1 core is quite stable in cycle 17. The max. decay ratio (DR) value of 0.37 was obtained from the stability evaluation of an APRM (average power range monitor) and LPRM (local power range monitor) signals measured at 66% (APRM) of reactor power and 4252 Kg/s (SA-HC) of core flow. Compared with the power profile in cycle 17 (as well as in reactor F2), the core in cycle 16 had an extreme power profile with high power and bottom-shifted axial peak in the core periphery esp. at the four quadrant corners. Such a profile decreases the stability margin in the region. It is a common observation that the DR obtained from APRM tends to be higher than that from LPRM if the global instability mechanism is dominant in the core, and vice versa. The comparison of global and local DR values should be an effective method for detecting local instability during the reactor operation. In order to detect the local instability it is important to evaluate the core stability with sufficient number of LPRMs so as to cover the whole core cross section together with APRMs 8 refs, 16 figs

  17. Parametric studies on containment thermal hydraulic loads during high pressure melt ejection in a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Silde, A.; Lindholm, I. [VTT Energy, Espoo (Finland)

    1997-12-01

    The containment thermal hydraulic loads during high pressure melt ejection in a Nordic BWR are studied parametrically with the CONTAIN and the MELCOR codes. The work is part of the Nordic RAK-2 project. The containment analyses were divided into two categories according to composition of the discharged debris: metallic and oxidic debris cases. In the base case with highly metallic debris, all sources from the reactor coolant system to the containment were based on the MELCOR/BH calculation. In the base case with the oxidic debris, the source data was specified assuming that {approx} 15% of the whole core material inventory and 34,000 kg of saturated water was discharged from the reactor pressure vessel (RPV) during 30 seconds. In this case, the debris consisted mostly of oxides. The highest predicted containment pressure peaks were about 8.5 bar. In the scenarios with highly metallic debris source, very high gas temperature of about 1900 K was predicted in the pedestal, and about 1400 K in the upper drywell. The calculations with metallic debris were sensititive to model parameters, like the particle size and the parameters, which control the chemical reaction kinetics. In the scenarios with oxidic debris source, the predicted pressure peaks were comparable to the cases with the metallic debris source. The maximum gas temperatures (about 450-500 K) in the containment were, however, significantly lower than in the respective metallic debris case. The temperatures were also insensitive to parametric variations. In addition, one analysis was performed with the MELCOR code for benchmarking of the MELCOR capabilities against the more detailed CONTAIN code. The calculations showed that leak tightness of the containment penetrations could be jeopardized due to high temperature loads, if a high pressure melt ejection occurred during a severe accident. Another consequence would be an early containment venting. (au). 28 refs.

  18. Development and Assessment of CTF for Pin-resolved BWR Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Salko, Robert K [ORNL; Wysocki, Aaron J [ORNL; Collins, Benjamin S [ORNL; Avramova, Maria [North Carolina State University (NCSU), Raleigh; Gosdin, Chris [Pennsylvania State University

    2017-01-01

    CTF is the modernized and improved version of the subchannel code, COBRA-TF. It has been adopted by the Consortium for Advanced Simulation for Light Water Reactors (CASL) for subchannel analysis applications and thermal hydraulic feedback calculations in the Virtual Environment for Reactor Applications Core Simulator (VERA-CS). CTF is now jointly developed by Oak Ridge National Laboratory and North Carolina State University. Until now, CTF has been used for pressurized water reactor modeling and simulation in CASL, but in the future it will be extended to boiling water reactor designs. This required development activities to integrate the code into the VERA-CS workflow and to make it more ecient for full-core, pin resolved simulations. Additionally, there is a significant emphasis on producing high quality tools that follow a regimented software quality assurance plan in CASL. Part of this plan involves performing validation and verification assessments on the code that are easily repeatable and tied to specific code versions. This work has resulted in the CTF validation and verification matrix being expanded to include several two-phase flow experiments, including the General Electric 3 3 facility and the BWR Full-Size Fine Mesh Bundle Tests (BFBT). Comparisons with both experimental databases is reasonable, but the BFBT analysis reveals a tendency of CTF to overpredict void, especially in the slug flow regime. The execution of these tests is fully automated, analysis is documented in the CTF Validation and Verification manual, and the tests have become part of CASL continuous regression testing system. This paper will summarize these recent developments and some of the two-phase assessments that have been performed on CTF.

  19. Recriticality calculation with GENFLO code for the BWR core after steam explosion in the lower head

    Energy Technology Data Exchange (ETDEWEB)

    Miettinen, J. [VTT Processes (Finland)

    2002-12-01

    Recriticality of the partially degraded BWR core has been studied by assuming a severe accident phase during which the fuel rods are still intact but the control rods have experienced extensive damage. Previous NKS and EU projects have studied the same case assuming reflooding by the ECCS system In the present study it was assumed that coolant enters the core due to melt-coolant interaction in the lower plenum. In the first case specified the relocation and fragmentation of the molten control rod metal causes the level swell in the core but no steam explosion. In the second case a steam explosion in the lower head was assumed. I n the first case a prompt recriticality peak can occur, but after the peak no semistable power generation remains. In the second case the consequence of the slug entrance into the core is so violent that the fuel disintegration and melting during the first power peak may occur. After the large power peak water is rapidly pushed back from the core and no semistable power generation maintains. The fuel disintegration studies have been based on a coarse assumption that the acceptable local energy addition into the fresh fuel may be 170 cal/g, but with increasing burn-up it can be as low as 60-70 cal/g. In the level swell variations the maximum energy addition was between these limits, but in most of the steam explosion variations much above these limits. Additional variation of the assumptions related to the neutronics demonstrated that for the converged analysis result some interactions would be useful with respect to the boundary conditions and neutronic options.

  20. Design of a mixed recharge with MOX assemblies of greater relation of moderation for a BWR reactor; Diseno de una recarga mixta con ensambles MOX de mayor relacion de moderacion para un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J.R.; Alonso V, G.; Palacios H, J. [ININ, Carretera Mexico-Toluca Km. 36.5, 52045 Estado de Mexico (Mexico)]. e-mail: jrrs@nuclear.inin.mx

    2004-07-01

    The study of the fuel of mixed oxides of uranium and plutonium (MOX) it has been topic of investigation in many countries of the world and those are even discussed in many places the benefits of reprocessing the spent fuel to extract the plutonium created during the irradiation of the fuel in the nuclear power reactors. At the moment those reactors that have been loaded partially with MOX fuel, are mainly of the type PWR where a mature technology has been achieved in some countries like they are France, Belgium and England, however the experience with reactors of the type BWR is more limited and it is continued studying the best way to introduce this type of fuel in BWRs, one of the main problems to introduce MOX in reactors BWR is the neutronic design of the same one, existing different concepts to introduce the plutonium in the assemblies of fuel and one of them is the one of increasing the relationship of moderation of the assemble. In this work a MOX fuel assemble design is presented and the obtained results so far in the ININ. These results indicate that the investigated concept has some exploitable advantages in the use of the MOX fuel. (Author)

  1. Development of the radiation models of a BWR type reactor and it facility in the SUN-RAH; Desarrollo de modelos de radiacion de un reactor tipo BWR y su instalacion en el SUN-RAH

    Energy Technology Data Exchange (ETDEWEB)

    Barron A, I. [Facultad de Ingenieria, UNAM, 04510 Mexico D.F. (Mexico)]. e-mail: isbarron@yahoo.com.mx

    2005-07-01

    This work about generation models, transport in processes and radioactive contamination of areas of a BWR central, is an amplification to the project developed in the UNAM to have a support tool in subjects or electric generation courses. It is planned about the implementation of models of radiation generation in a BWR type reactor for complement the functions developed in the University Simulator of Nucleo electric- Boiling water reactor (SUN-RAH) which it has been implemented in Simulink of MatLab and it has a model for the dynamics of one nucleo electric central that presents the main characteristics of the reactor vessel, the recirculation system, steam lines, turbines, generator, condensers and feeding water, defined by the main processes that intervene in the generation of energy of these plants. By this way the radiation monitoring systems for area and process, operate simultaneously with the processes of energy generation, with that is possible to observe the changes that present with respect to the operation conditions of the plant, and likewise to appreciate the radiation transport process through the components of the reactor, steam lines and turbines, for different operation conditions and possible faults that they could be presented during the reactor operation. (Author)

  2. Connected analysis nuclear-thermo-hydraulic of parallel channels of a BWR reactor using distributed computation; Analisis acoplado nuclear-termohidraulico de canales paralelos de un reactor BWR empleando computacion distribuida

    Energy Technology Data Exchange (ETDEWEB)

    Campos Gonzalez, Rina Margarita

    2007-07-15

    This work consists of the integration of three models previously developed which are described widely in Literature: model of the thermo-hydraulic channel, model of the modal neutronic and the model of the recirculation bows. The tool used for this connection of models is the PVM system, Parallel Virtual Machine that allowed paralleling the model by means of the concept of distributed computation. The purpose of making this connection of models is the one of obtaining a more complete tool than better represents the real configuration and the phenomenology of the nucleus of a BWR reactor, thus obtaining better results. In addition to maintaining the flexibility to improve the resulting model at any time, since the very complex or sophisticated models are difficult to improve being impossible to modify the equations they use and can include variables that are not of primary importance in the tackled problem or that mask relations among variables due to the excess of results. Also maintaining the flexibility for adding component of models or systems of the BWR reactor, all of this following the modeling needs. The Swedish Ringhals power plant was chosen to characterize the resulting connected model for counting on a Stability Benchmark that offers the opportunity to count on real plant data. Besides that in case 9 of cycle 14 of this Benchamark oscillations outside phase appeared, which are from great interest because the detection systems that register the average of the power of the nucleus do not detect them. Additionally in this work the model of the recirculation bows as an independent module is obtained in an individual way, since this model belongs to another work and works connected to the reactor vessel. The model of the recirculation bows is able to model several transients of interest, as it is shown in the Appendix A of this work, among which are found the tripping of recirculation pumps or the transference at low or high velocity of them. The scope of the

  3. Application of the MOVE algorithm for the identification of reduced order models of a core of a BWR type reactor; Aplicacion del algoritmo MOVE para la identificacion de modelos de orden reducido del nucleo de un reactor tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Victoria R, M.A.; Morales S, J.B. [UNAM, DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: angelvr@gmail.com

    2005-07-01

    Presently work is applied the modified algorithm of the ellipsoid of optimal volume (MOVE) to a reduced order model of 5 differential equations of the core of a boiling water reactor (BWR) with the purpose of estimating the parameters that model the dynamics. The viability is analyzed of carrying out an analysis that calculates the global dynamic parameters that determine the stability of the system and the uncertainty of the estimate. The modified algorithm of the ellipsoid of optimal volume (MOVE), is a method applied to the parametric identification of systems, in particular to the estimate of groups of parameters (PSE for their initials in English). It is looked for to obtain the ellipsoid of smaller volume that guarantees to contain the real value of the parameters of the model. The PSE MOVE is a recursive identification method that can manage the sign of noise and to ponder it, the ellipsoid represents an advantage due to its easy mathematical handling in the computer, the results that surrender are very useful for the design of Robust Control since to smaller volume of the ellipsoid, better is in general the performance of the system to control. The comparison with other methods presented in the literature to estimate the reason of decline (DR) of a BWR is presented. (Author)

  4. Proliferation Security Initiative (PSI)

    National Research Council Canada - National Science Library

    Squassoni, Sharon

    2005-01-01

    President Bush announced the Proliferation Security Initiative (PSI) on May 31, 2003. Since then, 16 nations have pledged their cooperation in interdicting shipments of weapons of mass destruction-related...

  5. Development of a methodology of analysis of instabilities in BWR reactors; Desarrollo de una metodologia de analisis de inestabilidades en reactores PWR

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Fenoll, M.; Abarca, A.; Barrachina, T.; Miro, R.; Verdu, G.

    2012-07-01

    This paper presents a methodology of analysis of the reactors instabilities of BWR type. This methodology covers of modal analysis of the point operation techniques of signal analysis and simulation of transients, through 3D Coupled RELAP5/PARCSv2.7 code.

  6. Data report of BWR post-CHF tests. Transient core thermal-hydraulic test program. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Iguchi, Tadashi; Itoh, Hideo; Kiuchi, Toshio; Watanabe, Hironori; Kimura, Mamoru; Anoda, Yoshinari [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-03-01

    JAERI has been performing transient core thermal-hydraulic test program. In the program, authors performed BWR/ABWR DBE simulation tests with a test facility, which can simulate BWR/ABWR transients. The test facility has a 4 x 4 bundle core simulator with 15-rod heaters and one non-heated rod. Through the tests, authors quantified the thermal safety margin for core cooling. In order to quantify the thermal safety margin, authors collected experimental data on post-CHF. The data are essential for the evaluation of clad temperature transient when core heat-up occurs during DBEs. In comparison with previous post-CHF tests, present experiments were performed in much wider experimental condition, covering high clad temperature, low to high pressure and low to high mass flux. Further, data at wider elevation (lower to higher elevation of core) were obtained in the present experiments, which make possible to discuss the effect of axial position on thermal-hydraulics, while previous works usually discuss the thermal-hydraulics at the position where the first heat-up occurs. This data report describes test procedure, test condition and major experimental data of post-CHF tests. (author)

  7. An assessment of BWR (boiling water reactor) Mark-II containment challenges, failure modes, and potential improvements in performance

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, D.L.; Jones, K.R.; Dallman, R.J. (EG and G Idaho, Inc., Idaho Falls, ID (USA)); Wagner, K.C. (Science Applications International Corp., Albuquerque, NM (USA))

    1990-07-01

    This report assesses challenges to BWR Mark II containment integrity that could potentially arise from severe accidents. Also assessed are some potential improvements that could prevent core damage or containment failure, or could mitigate the consequences of such failure by reducing the release of fission products to the environment. These challenges and improvements are analyzed via a limited quantitative risk/benefit analysis of a generic BWR/4 reactor with Mark II containment. Point estimate frequencies of the dominant core damage sequences are obtained and simple containment event trees are constructed to evaluate the response of the containment to these severe accident sequences. The resulting containment release modes are then binned into source term release categories, which provide inputs to the consequence analysis. The output of the consequences analysis is used to construct an overall base case risk profile. Potential improvements and sensitivities are evaluated by modifying the event tree spilt fractions, thus generating a revised risk profile. Several important sensitivity cases are examined to evaluate the impact of phenomenological uncertainties on the final results. 75 refs., 25 figs., 65 tabs.

  8. Technical Specification action statements requiring shutdown. A risk perspective with application to the RHR/SSW systems of a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Mankamo, T. [Avaplan Oy, Espoo (Finland); Kim, I.S.; Samanta, P.K. [Brookhaven National Lab., Upton, NY (United States)

    1993-11-01

    When safety systems fail during power operation, the limiting conditions for operation (LCOs) and associated action statements of technical specifications typically require that the plant be shut down within the limits of allowed outage time (AOT). However, when a system needed to remove decay heat, such as the residual heat removal (RHR) system, is inoperable or degraded, shutting down the plant may not necessarily be preferable, from a risk perspective, to continuing power operation over a usual repair time, giving priority to the repairs. The risk impact of the basic operational alternatives, i.e., continued operation or shutdown, was evaluated for failures in the RHR and standby service water (SSW) systems of a boiling-water reactor (BWR) nuclear power plant. A complete or partial failure of the SSW system fails or degrades not only the RHR system but other front-line safety systems supported by the SSW system. This report presents the methodology to evaluate the risk impact of LCOs and associated AOT; the results of risk evaluation from its application to the RHR and SSW systems of a BWR; the findings from the risk-sensitivity analyses to identify alternative operational policies; and the major insights and recommendations to improve the technical specifications action statements.

  9. Gemcitabine resistance in breast cancer cells regulated by PI3K/AKT-mediated cellular proliferation exerts negative feedback via the MEK/MAPK and mTOR pathways

    Directory of Open Access Journals (Sweden)

    Yang XL

    2014-06-01

    ability of 231/Gem cells. Western blot analysis showed that treatment with a PI3K/AKT inhibitor decreased the expression levels of p-AKT, p-MEK, p-mTOR, and p-P70S6K; however, treatments with either MEK/MAPK or mTOR inhibitor significantly increased p-AKT expression. Thus, our data suggest that gemcitabine resistance in breast cancer cells is mainly mediated by activation of the PI3K/AKT signaling pathway. This occurs through elevated expression of p-AKT protein to promote cell proliferation and is negatively regulated by the MEK/MAPK and mTOR pathways. Keywords: chemoresistance, gemcitabine, breast cancer

  10. Mitigation strategies of intergranular corrosion in systems of reactors of water boiling (BWR). Combined action of the chemistry of the hydrogen and the oxygen; Estrategias de mitigacion de la corrosion intergranular en sistemas de reactores de agua en ebullicion (BWR). Accion combinada de la quimica del hidrogeno y del oxigeno

    Energy Technology Data Exchange (ETDEWEB)

    Verdugo, M.

    2015-07-01

    Inter-Granular Stress Corrosion cracking (IGSCC) in austenitic stainless steel and in austenitic nickel-based alloys has been the subject of many studies the aim of which was to resolve one of the main problems faced by BWR nuclear power plants since the 1960s. This corrosion phenomenon is the result of the combined action of three factors: sensitization of the material, high local stresses and an aggressive medium. This paper deals with these factors separately and analyzes the oxidative chemistry of BWR reactors (aggressivity of the medium) as one the main causes if IGSCC. (Author)

  11. Cell proliferation in carcinogenesis

    Energy Technology Data Exchange (ETDEWEB)

    Cohen, S.M.; Ellwein, L.B. (Univ. of Nebraska Medical Center, Omaha (USA))

    1990-08-31

    Chemicals that induce cancer at high doses in animal bioassays often fail to fit the traditional characterization of genotoxins. Many of these nongenotoxic compounds (such as sodium saccharin) have in common the property that they increase cell proliferation in the target organ. A biologically based, computerized description of carcinogenesis was used to show that the increase in cell proliferation can account for the carcinogenicity of nongenotoxic compounds. The carcinogenic dose-response relationship for genotoxic chemicals (such as 2-acetylaminofluorene) was also due in part to increased cell proliferation. Mechanistic information is required for determination of the existence of a threshold for the proliferative (and carcinogenic) response of nongenotoxic chemicals and the estimation of risk for human exposure.

  12. Development of an Input Model to MELCOR 1.8.5 for the Oskarshamn 3 BWR

    Energy Technology Data Exchange (ETDEWEB)

    Nilsson, Lars [Lentek, Nykoeping (Sweden)

    2006-05-15

    An input model has been prepared to the code MELCOR 1.8.5 for the Swedish Oskarshamn 3 Boiling Water Reactor (O3). This report describes the modelling work and the various files which comprise the input deck. Input data are mainly based on original drawings and system descriptions made available by courtesy of OKG AB. Comparison and check of some primary system data were made against an O3 input file to the SCDAP/RELAP5 code that was used in the SARA project. Useful information was also obtained from the FSAR (Final Safety Analysis Report) for O3 and the SKI report '2003 Stoerningshandboken BWR'. The input models the O3 reactor at its current state with the operating power of 3300 MW{sub th}. One aim with this work is that the MELCOR input could also be used for power upgrading studies. All fuel assemblies are thus assumed to consist of the new Westinghouse-Atom's SVEA-96 Optima2 fuel. MELCOR is a severe accident code developed by Sandia National Laboratory under contract from the U.S. Nuclear Regulatory Commission (NRC). MELCOR is a successor to STCP (Source Term Code Package) and has thus a long evolutionary history. The input described here is adapted to the latest version 1.8.5 available when the work began. It was released the year 2000, but a new version 1.8.6 was distributed recently. Conversion to the new version is recommended. (During the writing of this report still another code version, MELCOR 2.0, has been announced to be released within short.) In version 1.8.5 there is an option to describe the accident progression in the lower plenum and the melt-through of the reactor vessel bottom in more detail by use of the Bottom Head (BH) package developed by Oak Ridge National Laboratory especially for BWRs. This is in addition to the ordinary MELCOR COR package. Since problems arose running with the BH input two versions of the O3 input deck were produced, a NONBH and a BH deck. The BH package is no longer a separate package in the new 1

  13. Direct torus venting analysis for Chinshan BWR-4 plant with MARK-I containment

    Energy Technology Data Exchange (ETDEWEB)

    Yuann, Yng-Ruey, E-mail: ryyuann@iner.gov.tw

    2017-03-15

    Highlights: • Study the effectiveness of Direct Torus Venting System (DTVS) during extended SBO of 24 h for Chinshan MARK-I plant. • Containment response is analyzed by GOTHIC based on boundary conditions from RETRAN calculation. • Analyses are performed with and without DTVS, respectively. • Suppression pool is sub-divided and thermal stratification is observed. - Abstract: The Chinshan plant, owned by Taiwan Power Company, has twin units of BWR-4 reactor and MARK-I containment. Both units have been operating at rated core thermal power of 1840 MWt. The existing Direct Torus Venting System (DTVS) is the main system used for venting the containment during the extended station blackout event. The purpose of this paper is to study the effects of the DTVS venting on the response of the containment pressure and temperature. The reactor is depressurized by manually opening the safety relief valves (SRVs) during the SBO, which causes the mass and energy to be discharged into and heat up the suppression pool. The RETRAN model is used to calculate the Nuclear Steam Supply System (NSSS) response and generate the SRV blowdown conditions, including SRV pressure, enthalpy, and mass flow rate. These conditions are then used as the time-dependent boundary conditions for the GOTHIC code to calculate the containment pressure and temperature response. The DTVS model is established in the GOTHIC model based on the venting size, venting piping loss, venting initiation time, and venting source. The lumped volume model, 1-D coarse-mesh model, and 3-D coarse-mesh model are considered in the torus volume. The calculation is first done without DTVS venting to establish a reference basis. Then a case with DTVS available is performed. Comparison of the two cases shows that the existing DTVS design is effective in mitigating the severity of the containment pressure and temperature transients. The results also show that the 1-D coarse-mesh model may not be appropriate since a

  14. BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

    Energy Technology Data Exchange (ETDEWEB)

    Bevard, Bruce Balkcom [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mertyurek, Ugur [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Scaglione, John M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-10-01

    utilized or referenced, justification has been provided as to why the data can be utilized for BWR fuel.

  15. Optimization of axial enrichment distribution for BWR fuels using scoping libraries and block coordinate descent method

    Energy Technology Data Exchange (ETDEWEB)

    Tung, Wu-Hsiung, E-mail: wstong@iner.gov.tw; Lee, Tien-Tso; Kuo, Weng-Sheng; Yaur, Shung-Jung

    2017-03-15

    Highlights: • An optimization method for axial enrichment distribution in a BWR fuel was developed. • Block coordinate descent method is employed to search for optimal solution. • Scoping libraries are used to reduce computational effort. • Optimization search space consists of enrichment difference parameters. • Capability of the method to find optimal solution is demonstrated. - Abstract: An optimization method has been developed to search for the optimal axial enrichment distribution in a fuel assembly for a boiling water reactor core. The optimization method features: (1) employing the block coordinate descent method to find the optimal solution in the space of enrichment difference parameters, (2) using scoping libraries to reduce the amount of CASMO-4 calculation, and (3) integrating a core critical constraint into the objective function that is used to quantify the quality of an axial enrichment design. The objective function consists of the weighted sum of core parameters such as shutdown margin and critical power ratio. The core parameters are evaluated by using SIMULATE-3, and the cross section data required for the SIMULATE-3 calculation are generated by using CASMO-4 and scoping libraries. The application of the method to a 4-segment fuel design (with the highest allowable segment enrichment relaxed to 5%) demonstrated that the method can obtain an axial enrichment design with improved thermal limit ratios and objective function value while satisfying the core design constraints and core critical requirement through the use of an objective function. The use of scoping libraries effectively reduced the number of CASMO-4 calculation, from 85 to 24, in the 4-segment optimization case. An exhausted search was performed to examine the capability of the method in finding the optimal solution for a 4-segment fuel design. The results show that the method found a solution very close to the optimum obtained by the exhausted search. The number of

  16. Analysis of core degradation and relocation phenomena and scenarios in a Nordic-type BWR

    Energy Technology Data Exchange (ETDEWEB)

    Galushin, Sergey, E-mail: galushin@kth.se; Kudinov, Pavel, E-mail: pkudinov@kth.se

    2016-12-15

    Highlights: • A data base of the debris properties in lower plenum generated using MELCOR code. • The timing of safety systems has significant effect on the relocated debris properties. • Loose coupling between core relocation and vessel failure analyses was established. - Abstract: Severe Accident Management (SAM) in Nordic Boiling Water Reactors (BWR) employs ex-vessel cooling of core melt debris. The melt is released from the failed vessel and poured into a deep pool of water located under the reactor. The melt is expected to fragment, quench, and form a debris bed, coolable by a natural circulation and evaporation of water. Success of the strategy is contingent upon melt release conditions from the vessel and melt-coolant interaction that determine (i) properties of the debris bed and its coolability (ii) potential for energetic melt-coolant interactions (steam explosions). Risk Oriented Accident Analysis Methodology (ROAAM+) framework is currently under development for quantification of the risks associated with formation of non-coolable debris bed and occurrence of steam explosions, both presenting a credible threats to containment integrity. The ROAAM+ framework consist of loosely coupled models that describe each stage of the accident progression. Core relocation analysis framework provides initial conditions for melt vessel interaction, vessel failure and melt release frameworks. The properties of relocated debris and melt release conditions, including in-vessel and ex-vessel pressure, lower drywell pool depth and temperature, are sensitive to the accident scenarios and timing of safety systems recovery and operator actions. This paper illustrates a methodological approach and relevant data for establishing a connection between core relocation and vessel failure analysis in ROAAM+ approach. MELCOR code is used for analysis of core degradation and relocation phenomena. Properties of relocated debris are obtained as functions of the accident scenario

  17. RELAP4/MOD5: a computer program for transient thermal-hydraulic analysis of nuclear reactors and related systems. User's manual. Volume III. Checkout applications. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    Obenchain, C. F.; Ramsthaler, J. H.; Eales, E. P.; Charlton, T. R.; Childs, F. W.; Giles, M. M.; Good, E. G.; Gruen, G. E.; Guttman, J.; Johnsen, G. W.; Katsma, K. R.; Keeler, C. D.; Lawford, T. W.; Mohr, C. M.; Singer, G. L.; Townsend, W. C.

    1976-09-01

    Checkout problems presented include the following: PWR large cold leg break; PWR small cold leg break; PWR intermediate sized cold leg break; BWR large recirculation line break; BWR small recirculation line break; INEL Semiscale small cold leg break; INEL LOFT large cold leg break and INEL Semiscale large cold leg break. Also included is Update 2 of the RELAP 4/M0D5 code.

  18. Simulations of ex-vessel fuel coolant interactions in a Nordic BWR using MC3D code

    Energy Technology Data Exchange (ETDEWEB)

    Thakre, S.; Ma, W. [Royal Institute of Technology, KTH. Div. of Nuclear Power Safety, Stockholm (Sweden)

    2013-08-15

    Nordic Boiling Water Reactors (BWRs) employ a drywell cavity flooding technique as a nuclear severe accident management strategy. In case of core melt accident where the reactor pressure vessel will fail and the melt will eject from the lower head and fall into a water pool, may be in the form of a continuous jet. It is assumed that the melt jet will fragment, quench and form a coolable debris bed into the water pool. The melt interaction with a water pool may cause an energetic steam explosion which creates a potential risk towards the integrity of containment, leading to fission products release into the atmosphere. The results of the APRI-7 project suggest that the significant damage to containment structures by steam explosion cannot be ruled according to the state-of-the-art knowledge about corresponding accident scenario. In the follow-up project APRI-8 (2012-2016) one of the goals of the KTH research is to resolve the steam explosion energetics (SEE) issue, developing a risk-oriented framework for quantifying conditional threats to containment integrity for a Nordic type BWR. The present study deals with the premixing and explosion phase calculations of a Nordic BWR dry cavity, using MC3D, a multiphase CFD code for fuel coolant interactions. The main goal of the study is the assessment of pressure buildup in the cavity and the impact loading on the side walls. The conditions for the calculations are used from the SERENA-II BWR case exercise. The other objective was to do the sensitivity analysis of the parameters in modeling of fuel coolant interactions, which can help to reduce uncertainty in assessment of steam explosion energetics. The results show that the amount of liquid melt droplets in the water (region of void<0.6) is maximum even before reaching the jet at the bottom. In the explosion phase, maximum pressure is attained at the bottom and the maximum impulse on the wall is at the bottom of the wall. The analysis is carried out using two different

  19. Estimation of the coolant flow through a natural circulation BWR fuel channel applying and equivalent electrical model

    Energy Technology Data Exchange (ETDEWEB)

    Valle H, J.; Morales S, J. B. [UNAM, DEPFI, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, Campus Morelos en IMTA, Jiutepec, Morelos (Mexico); Espinosa P, G., E-mail: julfi_ig@yahoo.com.m [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Av. San Rafael Atlixco No. 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)

    2010-10-15

    This work presents the design and implementation of an advanced controller for a reduced order model of a BWR reactor core cooled by natural circulating water, which allows real time estimates of coolant flows through fuel assemblies about standard neutron flux strings. Nuclear power plants with boiling water reactors control individual fuel assembly coolant flows by forced circulation using external or internal water pumps and different core support plate orifices. These two elements reduce flow dependency on local channel pressure drops. In BWR reactors using only natural circulation coolant flows, these two elements are not available and therefore individual channel coolant flows are highly dependent in local conditions, such as power distributions and local pressure drops. Therefore it is expected that grater uncertainties in these variables be used during safety, fuel management and other analysis, which in turns may lead to increased operation penalties, such as tighter operating limits. The objective of this work is to asses by computer simulations means to reduce uncertainties in the measurement of fuel assembly coolant flows and eventually the associated penalties. During coolant phase transitions, pressure drops and local power may alter local natural circulation through fuel assemblies and flow estimates can be helped or not by control rod moves. This work presents the construction of an optimal controller for a core flow estimator based on a reduced order model of the coolant going though the reactor vessel components and nuclear core. This model is to be driven by plant signals from standard BWR instrumentation in order to estimate the coolant flows in selected fuel assemblies about a LPRM string. For this purpose an equivalent electrical model has been mathematically developed and numerically tested. The power-flow maps of typical BRW are used as steady state references for this equivalent model. Once these were fully reproduced for steady state

  20. Recent trends in the mitigation of the IGSCC through modifications in the water chemistry of BWR reactors; Tendencias recientes en la mitigacion del IGSCC mediante modificaciones en la quimica del agua de reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Diaz S, A.; Robles, E.F. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2003-07-01

    During the last years, the Nuclear Power stations had been that to adequate or to modify the parameters and operational conditions, attempting to maintain and to safeguard the integrity and functionality of its components and systems, as well as the personnel safety involved in its operation. In a Boiling water reactor (BWR), the chemical control of the water, constitutes one of the fundamental aspects to get a sure and reliable operation, having as main objectives: (a) The protection of the reactor vessel, of the structural materials of the same one and of the pipes and components of those recirculation systems against the Intergranular stress corrosion phenomena (IGSCC); (b) To guarantee the integrity of the nuclear fuel minimizing the corrosion phenomena in the fuel elements; and (c) The reduction of the operational dose of the personnel involved directly in the operation and maintenance by means of the control of the activated corrosion products. (Author)

  1. A methodology for obtaining the control rod patterns in a BWR using genetic algorithms; Una metodologia para obtener los patrones de barras de control en un BWR usando algoritmos geneticos

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz S, J.J.; Montes T, J.L. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico); Requena R, I. [Universidad de Granada, 18071 Granada (Spain)]. e-mail: jjortiz@nuclear.inin.mx

    2003-07-01

    In this work the GACRP system based on the genetic algorithms technique for the obtaining of the drivers of control bars in a BWR reactor is presented. This methodology was applied to a transition cycle and a one of balance of the Laguna Verde nuclear power station (CNLV). For each one of the studied cycles, it was executed the methodology with a fixed length of the cycle and it was compared the effective multiplication factor of neutrons at the end of the cycle that it is obtained with the proposed drivers of control bars and the multiplication factor of neutrons obtained by means of a Haling calculation. It was found that it is possible to extend several days the length of both cycles with regard to the one Haling calculation. (Author)

  2. Impact analysis of modifying the composition of the nuclear fuel of a BWR with beryllium oxide; Analisis del impacto de modificar la composicion del combustible nuclear de un BWR con oxido de berilio

    Energy Technology Data Exchange (ETDEWEB)

    Gallardo V, J. M.; Morales S, J. B., E-mail: euqrop@hotmail.com [UNAM, Facultad de Ingenieria, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2013-10-15

    The beryllium oxide (Be O) presents excellent physical properties, especially its high thermal conductivity that contrasts clearly with that of the uranium dioxide (UO{sub 2}) used at the present as fuel in a great number of nuclear plants. The present work models a nuclear reactor cooled by light water in boiling with two external recirculation loops (BWR/5) using the code for the transitory analysis and postulated accidents Trac-B F1, implementing a UO{sub 2} mixture and different fractions of Be O, with the objective of improving the thermal conductivity of the fuel. The numeric results and the realized analyses indicate that when adding a fraction in volume of 10% the central temperature decreases in 30.4% in stationary state, while during the large break loss of coolant accident the peak cladding temperature diminishes in 7%. Although the real interaction of the mixture has not been determined experimentally, the obtained results are promising. (Author)

  3. MCTP, a code for the thermo-mechanical analysis of a fuel rod of BWR type reactors (Neutron part); MCTP, un codigo para el analisis termo-mecanico de una barra combustible de reactores tipo BWR (Parte Neutronica)

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez L, H.; Ortiz V, J. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: hhl@nuclear.inin.mx

    2003-07-01

    In the National Institute of Nuclear Research of Mexico a code for the thermo-mechanical analysis of the fuel rods of the BWR type reactors of the Nucleo electric Central of Laguna Verde is developed. The code solves the diffusion equation in cylindrical coordinates with several energy groups. The code, likewise, calculates the temperature distribution and power distribution in those fuel rods. The code is denominated Multi groups With Temperatures and Power (MCTP). In the code, the energy with which the fission neutrons are emitted it is divided in six groups. They are also considered the produced perturbations by the changes in the temperatures of the materials that constitute the fuel rods, the content of fission products, the uranium consumption and in its case the gadolinium, as well as the plutonium production. In this work there are present preliminary results obtained with the code, using data of operation of the Nucleo electric Central of Laguna Verde. (Author)

  4. BUTREN-RC an hybrid system for the recharges optimization of nuclear fuels in a BWR; BUTREN-RC un sistema hibrido para la optimizacion de recargas de combustible nuclear en un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz S, J.J.; Castillo M, J.A. [ININ, Carretera Mexico-Toluca Km. 36.5, 52045 Estado de Mexico (Mexico); Valle G, E. del [IPN, ESFM, 07738 Mexico D.F. (Mexico)

    2004-07-01

    The obtained results with the hybrid system BUTREN-RC are presented that obtains recharges of nuclear fuel for a BWR type reactor. The system has implemented the methods of optimization heuristic taboo search and neural networks. The optimization it carried out with the technique of taboo search, and the neural networks, previously trained, were used to predict the behavior of the recharges of fuel, in substitution of commercial codes of reactor simulation. The obtained recharges of nuclear fuel correspond to 5 different operation cycles of the Laguna Verde Nuclear Power plant, Veracruz in Mexico. The obtained results were compared with the designs of this cycles. The energy gain with the recharges of fuel proposals is of approximately 4.5% with respect to those of design. The time of compute consumed it was considerably smaller that when a commercial code for reactor simulation is used. (Author)

  5. Evaluation of the radial design of fuel cells in an operation cycle of a BWR reactor; Evaluacion del diseno radial de celdas de combustible en un ciclo de operacion de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez C, J.; Martin del Campo M, C. [Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Facultad de Ingenieria, UNAM, Paseo Cuauhnahuac 8532, Jiutepec, Morelos (Mexico)]. e-mail: jgco@ver.megared.net.mx

    2003-07-01

    This work is continuation of one previous in the one that the application of the optimization technique called Tabu search to the radial design of fuel cells of boiling water reactors (BWR, Boiling Water Reactor) is presented. The objective function used in the optimization process only include neutron parameters (k-infinite and peak of radial power) considering the cell at infinite media. It was obtained to reduce the cell average enrichment completing the characteristics of reactivity of an original cell. The objective of the present work is to validate the objective function that was used for the radial design of the fuel cell (test cell), analyzing the operation of a one cycle of the reactor in which fuels have been fresh recharged that contain an axial area with the nuclear database of the cell designed instead of the original cell. For it is simulated it with Cm-Presto the cycle 10 of the reactor operation of the Unit 1 of the Nuclear Power station of Laguna Verde (U1-CNLV). For the cycle evaluation its were applied so much the simulation with the Haling strategy, as the simulation of the one cycle with control rod patterns and they were evaluated the energy generation and several power limits and reactivity that are used as design parameters in fuel reloads of BWR reactors. The results at level of an operation cycle of the reactor, show that the objective function used in the optimization and radial design of the cell is adequate and that it can induce to one good use of the fuel. (Author)

  6. BWR simulation in a stationary state for the evaluation of fuel cell design; Simulacion de un reactor BWR en estado estacionario para la evaluacion del diseno de celdas de combustible

    Energy Technology Data Exchange (ETDEWEB)

    Montes T, J. L.; Ortiz S, J. J.; Perusquia del C, R.; Castillo M, A., E-mail: joseluis.montes@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    In this paper the simulation of a BWR in order to evaluate the performance of a set of fuel assemblies under stationary state in three dimensions (3-D) is presented. 15 cases selected from a database containing a total of 18225 cases are evaluated. The main selection criteria were based on the results of the design phase of the power cells in two dimensions (2-D) and 3-D initial study. In 2-D studies the parameters that were used to qualify and select the designs were basically the local power peaking factor and neutron multiplication factor of each fuel cell. In the initial 3-D study variables that defined the quality of results, and from which the selection was realized, are the margins to thermal limits of reactor operation and the value of the effective multiplication factor at the end of cycle operation. From the 2-D and 3-D results of the studies described a second 3-D study was realized, where the optimizations of the fuel reload pattern was carried out. The results presented in this paper correspond to this second 3-D study. It was found that the designs of the fuel cell they had a similar behavior to those provided by the fuel supplier of reference BWR. Particularly it noted the impact of reload pattern on the cold shut down margin. An estimate of the operation costs of reference cycle analyzed with each one designed reload batch was also performed. As a result a positive difference (gain) up to 10,347 M/US D was found. (Author)

  7. Simulation in 3 dimensions of a cycle 18 months for an BWR type reactor using the Nod3D program; Simulacion en 3 dimensiones de un ciclo de 18 meses para un reactor BWR usando el programa Nod3D

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez, N.; Alonso, G. [ININ, A.P. 18-1027, 11801 Mexico D.F. (Mexico)]. E-mail: nhm@nuclear.inin.mx; Valle, E. del [IPN, ESFM, 07738 Mexico D.F. (Mexico)

    2004-07-01

    The development of own codes that you/they allow the simulation in 3 dimensions of the nucleus of a reactor and be of easy maintenance, without the consequent payment of expensive use licenses, it can be a factor that propitiates the technological independence. In the Department of Nuclear Engineering (DIN) of the Superior School of Physics and Mathematics (ESFM) of the National Polytechnic Institute (IPN) a denominated program Nod3D has been developed with the one that one can simulate the operation of a reactor BWR in 3 dimensions calculating the effective multiplication factor (kJJ3, as well as the distribution of the flow neutronic and of the axial and radial profiles of the power, inside a means of well-known characteristics solving the equations of diffusion of neutrons numerically in stationary state and geometry XYZ using the mathematical nodal method RTN0 (Raviart-Thomas-Nedelec of index zero). One of the limitations of the program Nod3D is that it doesn't allow to consider the burnt of the fuel in an independent way considering feedback, this makes it in an implicit way considering the effective sections in each step of burnt and these sections are obtained of the code Core Master LEND. However even given this limitation, the results obtained in the simulation of a cycle of typical operation of a reactor of the type BWR are similar to those reported by the code Core Master LENDS. The results of the keJ - that were obtained with the program Nod3D they were compared with the results of the code Core Master LEND, presenting a difference smaller than 0.2% (200 pcm), and in the case of the axial profile of power, the maxim differs it was of 2.5%. (Author)

  8. Calculation of releases of radioactive materials in gaseous and liquid effluents from boiling water reactors (BWR-GALE Code)

    Energy Technology Data Exchange (ETDEWEB)

    Bangart, R.L.; Bell, L.G.; Boegli, J.S.; Burke, W.C.; Lee, J.Y.; Minns, J.L.; Stoddart, P.G.; Weller, R.A.; Collins, J.T.

    1978-12-01

    The calculational procedures described in the report reflect current NRC staff practice. The methods described will be used in the evaluation of applications for construction permits and operating licenses docketed after January 1, 1979, until this NUREG is revised as a result of additional staff review. The BWR-GALE (Boiling Water Reactor Gaseous and Liquid Effluents) Code is a computerized mathematical model for calculating the release of radioactive material in gaseous and liquid effluents from boiling water reactors (BWRs). The calculations are based on data generated from operating reactors, field tests, laboratory tests, and plant-specific design considerations incorporated to reduce the quantity of radioactive materials that may be released to the environment.

  9. Safety and licensing issues that are being addressed by the Power Burst Facility test programs. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    McCardell, R.K.; MacDonald, P.E.

    1980-01-01

    This paper presents an overview of the results of the experimental program being conducted in the Power Burst Facility and the relationship of these results to certain safety and licensing issues. The safety issues that were addressed by the Power-Cooling-Mismatch, Reactivity Initiated Accident, and Loss of Coolant Accident tests, which comprised the original test program in the Power Burst Facility, are discussed. The resolution of these safety issues based on the results of the thirty-six tests performed to date, is presented. The future resolution of safety issues identified in the new Power Burst Facility test program which consists of tests which simulate BWR and PWR operational transients, anticipated transients without scram, and severe fuel damage accidents, is described.

  10. A reactor noise analysis methodology for BWR core stability evaluation: application and assessment to Leibstadt stability tests

    Energy Technology Data Exchange (ETDEWEB)

    Dokhane, A. [Ecole Polytechnique Federale de Lausanne, CH-1015 Lausanne (Switzerland); Dokhane, A.; Ferroukhi, H.; Zimmermann, M.A. [Paul Scherrer Institut, Lab. for Reactor Physics and Systems Behavior, CH-5232 Villigen PSI (Switzerland); Aguirre, C. [Kernkraftwerk Leibstadt, CH-5325 Leibstadt (Switzerland)

    2005-07-01

    As a first step towards establishing a best-estimate methodology for the evaluation of BWR core stability parameters, i.e. the decay ratio and resonance frequency, from measured reactor noise signals, a systematic approach has recently been developed and adopted at PSI for the analysis of the Swiss BWRs. The aim is to evaluate stability tests in a consistent manner at any operating condition in the power/flow map and for any operating cycle. This methodology principally consists in a general reactor noise analysis where as much as possible information recorded during the tests is investigated prior to determining 'core representative' stability parameters along with an associated uncertainty range. A central part in this approach is that a time series analysis of all measured neutron flux signals, rather than only one or few signals, is performed. In addition, for each signal, three different model-order optimization methods are systematically employed to take into account the sensitivity upon the model-order. The adopted methodology is then applied to the evaluation of the core stability measurements performed at the Kernkraftwerk Leibstadt nuclear power plant, Switzerland, during cycles 10, 13 and 19. A total of 28 tests are hence analyzed. This is primarily done in order to obtain a broad range of tests to serve as basis for the validation of the coupled neutronic/thermal-hydraulic codes used at PSI for BWR stability calculations. In addition, in order to assess the results obtained with the current methodology, a comparative study has been carried out with respect to results from previously developed and applied procedures. The results show a good agreement between the current method and the other methods. (authors)

  11. JPRS Report, Proliferation Issues

    Science.gov (United States)

    1992-05-27

    JPRS-TND-92-016 27 MAY 1992 JPRS Repor Proliferation Issues ÄBpxovea tcz pursue ieiaaM| Ipfe. fötmbuasa OsüoaÜBd .^L ■ — —— au »** 19980112...6 MOROCCO Berrada on Proposed Nuclear Power Plant [ MAROC SOIR 22 Apr] 6 JPRS-TND-92-016 27 May 1992 2 CENTRAL EURASIA Proposals on...three days of talks here on normalizing relations with Japan, which were largely stalemated over Tokyo’s demand for Pyongyang’s assurance that it did

  12. Design and optimization of a fuel reload of BWR with plutonium and minor actinides; Diseno y optimizacion de una recarga de combustible de BWR con plutonio y actinidos menores

    Energy Technology Data Exchange (ETDEWEB)

    Guzman A, J. R.; Francois L, J. L.; Martin del Campo M, C.; Palomera P, M. A. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, Jiutepec, Morelos 62550 (Mexico)]. e-mail: maestro_juan_rafael@hotmail.com

    2008-07-01

    In this work is designed and optimized a pattern of fuel reload of a boiling water reactor (BWR), whose fuel is compound of uranium coming from the enrichment lines, plutonium and minor actinides (neptunium, americium, curium); obtained of the spent fuel recycling of reactors type BWR. This work is divided in two stages: in the first stage a reload pattern designs with and equilibrium cycle is reached, where the reload lot is invariant cycle to cycle. This reload pattern is gotten adjusting the plutonium content of the assembly for to reach the length of the wished cycle. Furthermore, it is necessary to increase the concentration of boron-10 in the control rods and to introduce gadolinium in some fuel rods of the assembly, in order to satisfy the margin approach of out. Some reactor parameters are presented: the axial profile of power average of the reactor core, and the axial and radial distribution of the fraction of holes, for the one reload pattern in balance. For the design of reload pattern codes HELIOS and CM-PRESTO are used. In the second stage an optimization technique based on genetic algorithms is used, along with certain obtained heuristic rules of the engineer experience, with the intention of optimizing the reload pattern obtained in the first stage. The objective function looks for to maximize the length of the reactor cycle, at the same time as that they are satisfied their limits related to the power and the reactor reactivity. Certain heuristic rules are applied in order to satisfy the recommendations of the fuel management: the strategy of the control cells core, the strategy of reload pattern of low leakage, and the symmetry of a quarter of nucleus. For the evaluation of the parameters that take part in the objective function it simulates the reactor using code CM-PRESTO. Using the technique of optimization of the genetic algorithms an energy of the cycle of 10834.5 MW d/tHM is obtained, which represents 5.5% of extra energy with respect to the

  13. Proliferation: myth or reality?; La proliferation: mythe ou realite?

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-07-01

    This article analyzes the proliferation approach, its technical condition and political motivation, and the share between the myth (political deception, assumptions and extrapolations) and the reality of proliferation. Its appreciation is complicated by the irrational behaviour of some political actors and by the significant loss of the non-use taboo. The control of technologies is an important element for proliferation slowing down but an efficient and autonomous intelligence system remains indispensable. (J.S.)

  14. Actinides record, power calculations and activity for present isotopes in the spent fuel of a BWR; Historial de actinidos y calculos de potencia y actividad para isotopos presentes en el combustible gastado de un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Enriquez C, P.; Ramirez S, J. R.; Lucatero, M. A., E-mail: pastor.enriquez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-10-15

    The administration of spent fuel is one of the more important stages of the nuclear fuel cycle, and this has become a problem of supreme importance in countries that possess nuclear reactors. Due to this in this work, the study on the actinides record and present fission products to the discharge of the irradiated fuel in a light water reactor type BWR is shown, to quantify the power and activity that emit to the discharge and during the cooling time. The analysis was realized on a fuel assembly type 10 x 10 with an enrichment average of 3.69 wt % in U-235 and the assembly simulation assumes four cycles of operation of 18 months each one and presents an exposition of 47 G Wd/Tm to the discharge. The module OrigenArp of the Scale 6 code is the computation tool used for the assembly simulation and to obtain the results on the actinides record presents to the fuel discharge. The study covers the following points: a) Obtaining of the plutonium vector used in the fuel production of mixed oxides, and b) Power calculation and activity for present actinides to the discharge. The results presented in this work, correspond at the same time immediate of discharge (0 years) and to a cooling stage in the irradiated fuel pool (5 years). (Author)

  15. Optimal estimate of the coolant flow in the assemblies of a BWR of natural circulation in real time; Estimacion optima del flujo de refrigerante en los ensambles de un BWR de circulacion natural en tiempo real

    Energy Technology Data Exchange (ETDEWEB)

    Valle H, J.; Morales S, J. B. [UNAM, Facultad de Ingenieria, Division de Estudios de Posgrado, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, Col. Progreso, 62550 Jiutepec, Morelos (Mexico); Espinosa P, G., E-mail: julfi_jg@yahoo.com.mx [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)

    2012-10-15

    The present work exposes the design and the implementation of an advanced controller that allows estimating the coolant flow in the fuel assemblies of a BWR reactor of natural circulation in real time. To be able to reduce the penalizations that are established in the calculations of the operation limits due to the magnitude of the uncertainties in the coolant flows of a natural circulation reactor, is the objective of this research. In this work the construction of the optimal controller that allows estimating the coolant flows in a fuel channels group of the reactor is shown, as well as the operation of this applied to a reduced order model that simulates the dynamics of a natural circulation reactor. The controller design required of an estimator of the valuation variables not directly of the plant and of the estimates use of the local distributions of the coolant flow. The controller construction of the estimator was based mathematically in the filter Kalman whose algorithm allows to be carried out an advanced control of the system. To prove the estimator operation was development a simplified model that reproduces the basic dynamics of the flowing coolant in the reactor, which works as observer of the system, this model is coupled by means of the estimator controller to a detail model of the plant. The results are presented by means of graphics of the interest variables and the estimate flow, and they are documented in the chart that is presented at the end of this article. (Author)

  16. Solution to the transport equation with anisotropic dispersion in a BWR type assembly using the AZTRAN code; Solucion de la ecuacion de transporte con dispersion anisotropica en un ensamble tipo BWR usando el codigo AZTRAN

    Energy Technology Data Exchange (ETDEWEB)

    Chepe P, M. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, San Rafael Atlixco No. 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico); Xolocostli M, J. V.; Gomez T, A. M. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Del Valle G, E., E-mail: liaison.web@gmail.com [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. San Pedro Zacatenco, 07730 Ciudad de Mexico (Mexico)

    2016-09-15

    Due to the current computing power, the deterministic codes for analyzing nuclear reactors that have been used for several years are becoming more relevant, since much more precise solution techniques can be used; the last century would have been very difficult, since memory and processor capacities were very limited or had high prices on the components. In this work we analyze the effect of the anisotropic dispersion of the effective dispersion section, compared to the isotropic dispersion. The anisotropy implementation was carried out in the AZTRAN transport code, which is part of the AZTLAN platform for nuclear reactors analysis (in development). The AZTRAN code solves the Boltzmann transport equation in one, two and three dimensions at steady state, using the multi-group technique for energy discretization, the RTN-0 nodal method in spatial discretization and for angular discretization the discrete ordinates without considering anisotropy originally. The effect of the anisotropy dispersion on the effective multiplication factor and the axial and radial power on a fuel assembly BWR type are analyzed. (Author)

  17. Nuclear fuel activity with minor actinides after their useful life in a BWR; Actividad del combustible nuclear con actinidos menores despues de su vida util en un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Martinez C, E.; Ramirez S, J. R.; Alonso V, G., E-mail: eduardo.martinez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2016-09-15

    Nuclear fuel used in nuclear power reactors has a life cycle, in which it provides energy, at the end of this cycle is withdrawn from the reactor core. This used fuel is known as spent nuclear fuel, a strong problem with this fuel is that when the fuel was irradiated in a nuclear reactor it leaves with an activity of approximately 1.229 x 10{sup 15} Bq. The aim of the transmutation of actinides from spent nuclear fuel is to reduce the activity of high level waste that must be stored in geological repositories and the lifetime of high level waste; these two achievements would reduce the number of necessary repositories, as well as the duration of storage. The present work is aimed at evaluating the activity of a nuclear fuel in which radioactive actinides could be recycled to remove most of the radioactive material, first establishing a reference of actinides production in the standard nuclear fuel of uranium at end of its burning in a BWR, and a fuel rod design containing 6% of actinides in an uranium matrix from the enrichment tails is proposed, then 4 standard uranium fuel rods are replaced by 4 actinide bars to evaluate the production and transmutation of the same, finally the reduction of actinide activity in the fuel is evaluated. (Author)

  18. Simulation of the flow obstruction of a jet pump in a BWR reactor with the code RELAP/SCDAPSIM; Simulacion de la obstruccion de flujo de una bomba jet en un reactor BWR con el codigo RELAP/SCDAPSIM

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas V, J.; Filio L, C., E-mail: jaime.cardenas@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose M. Barragan 779, Col. Narvarte, 03020 Ciudad de Mexico (Mexico)

    2016-09-15

    This work simulates the flow obstruction of a jet pump in one of the recirculation loops of a nuclear power plant with a reactor of type BWR at 100% of operating power, in order to analyze the behavior of the total flow of the refrigerant passing through the reactor core, the total flow in each recirculation loop of the reactor, together with the 10 jet pumps of each loop. The behavior of the power and the reactivity insertion due to the change of the refrigerant flow pattern is also analyzed. The simulation was carried out using the RELAP/SCDAPSIM version 3.5 code, using a reactor model with 10 jet pumps in each recirculation loop and a core consisting of 6 radial zones and 25 axial zones. The scenario postulates the flow obstruction in a jet pump in a recirculation loop A when the reactor operates at 100% rated power, causing a change in the total flow of refrigerant in the reactor core, leading to a decrease in power. Once the reactor conditions are established to its new power, the operator tries to recover the nominal power using the flow control valve of the recirculation loop A, opening stepwise as a strategy to safely recover the reactor power. In this analysis is assumed that the intention of the nuclear plant operator is to maintain the operation of the reactor during the established cycle. (Author)

  19. Calculation of the neutron flux and fluence in the covering of the nucleus and the vessel of a BWR; Calculo del flujo neutronico y fluencia en la envolvente del nucleo y la vasija de un reactor nuclear BWR

    Energy Technology Data Exchange (ETDEWEB)

    Martinez C, E.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, U. P. Adolfo Lopez Mateos, Col. Lindavista, 07738 Mexico D. F. (Mexico); Longoria G, L. C., E-mail: evalle@esfm.ipn.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2011-11-15

    One of the main objectives related with the safety in any nuclear power plant, including the nuclear power plant of Laguna Verde, is to guarantee the structural integrity of the pressure vessel of the reactor. To identify and quantifying the damage caused be neutron irradiation in the vessel of any nuclear reactor, is necessary to know as much the neutron flux as the fluence that it has been receiving during their time of operation life, since the observables damages by means of tests mechanics are products of micro-structural effects, induced by neutron irradiation, therefore, is important the study and prediction of the neutron flux to have a better knowledge of the damage that are receiving these materials. In our calculation the code DORT was used, which solves the transport equation in discreet coordinates and in two dimensions (x-y, r-{theta} and r-z), in accord to the regulator guide, it requires to make and approach of the neutron flux in three dimensions by means of the Synthesis Method. Whit this method is possible to achieve a representation of the flux in 3D combining or synthesizing the calculated fluxes by DORT code in r-{theta}, r-z and r. In this work the application of the Synthesis Method is presented, according to the Regulator Guide 1.190, to determine the fluxes 3D in the interns of a BWR using three different space meshes. (Author)

  20. Thermomechanical analysis of a fuel rod in a BWR reactor using the FUELSIM code; Analisis termomecanico de una barra de combustible de un reactor BWR utilizando el codigo FUELSIM

    Energy Technology Data Exchange (ETDEWEB)

    Pantoja C, R. [Escuela Superior de Fisica y Matematicas, Departamento de Ingenieria Nuclear, IPN, Av. Instituto Politecnico Nacional s/n, Col. San Pedro Zacatenco, 07738 Mexico, D. F. (Mexico); Ortiz V, J.; Araiza M, E. [ININ, Departamento de Sistemas Nucleares, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)], e-mail: rapaca78@yahoo.com.mx

    2009-10-15

    The thermomechanical behaviour of a fuel rod exposed to irradiation is a complex process in which are coupled great quantity of interrelated physical-chemical phenomena, for that analysis of rod performance in the core of a nuclear power reactor is realized generally with computation codes that integrate several phenomena expected during the time life of fuel rod in the core. An application of this type of thermomechanical codes is to predict, inside certain reliability margin, the design parameters that would be required to adjust, in order to get a better economy or rod performance, for a systematic approach to the fuel design optimization. FUELSIM is a thermomechanical code based on the models of FRAPCON code, which was developed under auspice of Nuclear Regulatory Commission of USA. FUELSIM allows iterative calculations like part of its programming structure, allowing search of extreme cases of behaviour, probabilistic analysis (or statistical), parametric analysis (or sensibility) and also can include as entrance data to the uncertainties associated with production data, code parameters and associated models. In this work is reported a first analysis of thermomechanical performance of a typical fuel rod used in a BWR 5/6. Results of maximum temperatures are presented in the fuel center and of axial deformation, for the 10 axial nodes in that the active longitude of fuel rod was divided. (Author)

  1. Comparison of results for burning with BWR reactors CASMO and SCALE 6.2 (TRITON / NEWT); Comparacion de los resultados de quemado para reactores BWR con CASMO y SCALE 6.2 (TRITON/NEWT)

    Energy Technology Data Exchange (ETDEWEB)

    Mesado, C.; Miro, R.; Barrachina, T.; Verdu, G.

    2014-07-01

    In this paper we compare the results from two codes burned, CASMO and SCALE 6.2 (TRITON). To do this, is simulated all segments corresponding to a boiling water reactor (BWR) using both codes. In addition, to account for different working points, simulations changing the instantaneous variables, these are repeated: void fractions (6 points), fuel temperature (6 points) and control rods (two points), with a total of 72 possible combinations of different instantaneous variables for each segment. After all simulations are completed for each segment, we can reorder the obtained cross sections, as SCALE CASMO both, to create a library of compositions nemtab format. This format is accepted by the neutronic code of nodal diffusion, PARCS v2.7. Finally compares the results obtained with PARCS and with the SIMULATE3 -SIMTAB methodology to level of full reactor. Also, we have made use of the KENO-VI and MCDANCOFF modules belonging to SCALE. The first is a Monte Carlo transport code with which you can validate the value of the multiplier, the second has been used to obtain values of Dancoff factor and increase the accuracy of model SCALE. (Author)

  2. Simulation of the automatic depressurization system (Ads) for a boiling water reactor (BWR) based on RELAP; Simulacion del sistema de despresurizacion automatica (ADS) para un reactor de agua en ebullicion (BWR) basado en RELAP

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez G, C.; Chavez M, C., E-mail: ces.raga@gmail.com [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2012-10-15

    The automatic depressurization system (Ads) of the boiling water reactor (BWR) like part of the emergency cooling systems is designed to liberate the vapor pressure of the reactor vessel, as well as the main vapor lines. At the present time in the Engineering Faculty, UNAM personnel works in the simulation of the Laguna Verde reactor based on the nuclear code RELAP/SCADAP and in the incorporation to the same of the emergency cooling systems. The simulation of the emergency cooling systems began with the inclusion of two hydrodynamic volumes, one source and another drain, and the incorporation of the initiation logic for each emergency system. In this work is defined and designed a simplified model of Ads of the reactor, considering a detail level based on the main elements that compose it. As tool to implement the proposed model, the RELAP code was used. The simulated main functions of Ads are centered in the quick depressurization of the reactor by means of the vapor discharge through the relief/safety valves to the suppression pool, and, in the event of break of the main vapor line, the reduction of the vessel pressure operates for that the cooling systems of the core to low pressure (Lpcs and Lpci) they can begin their operation. (Author)

  3. Analysis of core stability measurement data of advanced 9 x 9 fuel assembly in a BWR core

    Energy Technology Data Exchange (ETDEWEB)

    Tsuda, Katsuhiro; Itami, Akira; Kubo, Yuichiro; Shakudo, Taketomi [Nuclear Fuel Industries Ltd., Tokyo (Japan); Kreuter, D.; Anegawa, Takafumi; Kitamura, Hideya; Ishikawa, Masumi

    1997-05-01

    The core stability measurements were taken during the cycle-9 startup of the 1,300 MWe BWR, Kernkraftwerk Kruemmel (KKK). The core contained advanced 9 x 9 type high burn-up design reload fuel with a higher enrichment than current 8 x 8 fuel. A design feature of the advanced 9 x 9 fuel assembly (FA) is a large square water channel for enhanced neutron moderation. The measurement data as a function of core flow and power showed almost the same stability characteristics as those of the past measurement during the cycle-3 startup of the KKK core with the 8 x 8 FA. The local power range monitors (LPRM) detected neutron flux oscillations in both core-wide in-phase and half-core out-of-phase modes. The frequency-domain stability analysis using the STAIF-PK code well reproduced the measurement result that the onset of unstable operation in KKK first occurs when about half of the reactor internal pumps are operating and the other half are stopped. The stability performance of the advanced 9 x 9 FA in the core was compared with the 8 x 8 FA by a design parameter analysis with respect to thermal-hydraulic and neutronic design. It has been demonstrated by the analysis that the stability performance of the advanced 9 x 9 FA is comparable with current 8 x 8 FA. (author)

  4. OECD/NEA burnup credit criticality benchmarks phase IIIA: Criticality calculations of BWR spent fuel assemblies in storage and transport

    Energy Technology Data Exchange (ETDEWEB)

    Okuno, Hiroshi; Naito, Yoshitaka [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Ando, Yoshihira [Toshiba Corp., Kawasaki, Kanagawa (Japan)

    2000-09-01

    The report describes the final results of Phase IIIA Benchmarks conducted by the Burnup Credit Criticality Calculation Working Group under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development (OECD/NEA). The benchmarks are intended to confirm the predictive capability of the current computer code and data library combinations for the neutron multiplication factor (k{sub eff}) of a layer of irradiated BWR fuel assembly array model. In total 22 benchmark problems are proposed for calculations of k{sub eff}. The effects of following parameters are investigated: cooling time, inclusion/exclusion of FP nuclides and axial burnup profile, and inclusion of axial profile of void fraction or constant void fractions during burnup. Axial profiles of fractional fission rates are further requested for five cases out of the 22 problems. Twenty-one sets of results are presented, contributed by 17 institutes from 9 countries. The relative dispersion of k{sub eff} values calculated by the participants from the mean value is almost within the band of {+-}1%{delta}k/k. The deviations from the averaged calculated fission rate profiles are found to be within {+-}5% for most cases. (author)

  5. Fission product transport analysis: Task 2. Quarterly progress report, April--June 1977. [PWR; BWR; primary system reflooding following LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Gieseke, J.A.; Jordan, H.; Baybutt, P.; Wooton, R.O.; Denning, R.S.

    1977-09-30

    Continuing activities associated with modeling the transport and deposition of fission products within PWR and BWR primary systems during the reflood time period following a terminated LOCA are reported. The original scope of the overall project has been expanded to include consideration of conditions consistent with those leading to postulated core meltdown situations. Initial tasks on this continuation study involve limited improvements to the TRAP codes and performance of sensitivity analyses for terminated LOCA's plus the evaluation of thermal-hydraulic conditions for use in evaluating fission product transport and deposition in postulated meltdown situations. Beyond these initial tasks, emphasis will be concentrated solely on meltdown analyses. Major efforts within the past quarter have included completion in final form of the informal interim report on analyses for terminated LOCA conditions, initiation of improvement for the TRAP codes to include additional fission product deposition mechanisms, consideration of suitable methods for performing the sensitivity analyses, specification of ranges for variables to be covered in the sensitivity analyses, and initiation of efforts to specify flow conditions to be assumed in future development of analysis procedures applicable to meltdown conditions.

  6. Effect of Nitrogen Addition in 304 L Stainless Steel on the IGSCC Crack Growth Rate in Simulated BWR Environment

    Science.gov (United States)

    Roychowdhury, S.; Kain, V.; Prasad, R. C.

    Intergranular Stress Corrosion Cracking (IGSCC) in austenitic Stainless Steels (SS) in Boiling Water Reactor (BWR) operating conditions have been reported worldwide. Nitrogen containing Stainless Steel is used in BWRs and it can affect IGSCC behavior. In this investigation type 304L stainless steel with two different levels of nitrogen was evaluated in the sensitized and non-sensitised strain-hardened condition. Experiments were carried out in high temperature water with controlled dissolved oxygen. In the sensitised condition, the Crack Growth Rate (CGR) reduced and in the non-sensitised strain-hardened condition the CGR increased with increase in nitrogen level in SS. Transmission electron microscopic (TEM) investigations of the as-rolled SS and the SS after tensile testing at 288 °C indicated that rolling resulted in higher grain boundary strain which is a possible cause for higher CGR in the SS with higher nitrogen. Nitrogen did not have a noticeable effect on the deformation mechanism, for the SS after tensile testing at 288 °C, and the dislocation structures observed were similar for both the SS.

  7. Methods and results of a PSA level 2 for a German BWR of the 900 MWe class

    Energy Technology Data Exchange (ETDEWEB)

    Loffler, H.; Sonnenkalb, M. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Koln (Germany)

    2006-07-01

    On behalf of the federal Ministry for Environment, Nature Conservation and Reactor Safety (BMU) GRS has performed a PSA level 2 for a BWR type 69 NPP of the 900 MWe class, equipped with a N{sub 2} inerted steel containment and a pressure suppression system. Integral deterministic accident analyses have been performed with the computer code MELCOR 1.8.5. Additional analyses have been done for those events and phenomena which are not or not sufficiently covered by MELCOR. The probabilistic event tree analysis begins with the core damage states received from PSA level 1, and it ends with the definition of release categories and the determination of their frequencies. Uncertainties about the frequency of core damage states and about events during the accident progression are taken into account by means of Monte Carlo simulations. If there is a core damage state there is a high probability (>50 %) for a very high and rapid release of radionuclides into the environment. This high conditional probability is due to the very low probability to retain a partly destroyed core inside the reactor pressure vessel (RPV) and because the containment almost certainly fails at the bottom of the control rod drives room after melt release from the failed RPV. (authors)

  8. Histone deacetylase inhibitors repress chondrosarcoma cell proliferation.

    Science.gov (United States)

    Zhu, Jiaxue; Gu, Jianhua; Ma, Jie; Xu, Zhixing; Tao, Hairong

    2015-01-01

    Due to the high resistance to conventional therapy, there is still no convincingly effective treatment for chondrosarcoma. As a promising new treatment strategy, histone deacetylase inhibitors (HDACIs) have been reported to induce cell arrest, apoptosis and differentiation in some kinds of malignancies, but how HDACi exert their effects on chondrosarcoma is not well understood yet. We investigated the effects of HDACIs trichostatin A (TSA) and sodium valproate (VPA) on chondrosarcoma cells in vitro and in vivo. The cell proliferation and cell cycle were examined in two chondrosarcoma cell lines, SW1353 and JJ012, by MTS and flow cytometry assays, respectively. The in vivo effects of HDACIs were investigated by assessing the chondrosarcoma growth in a mouse xenograft model. Our results showed that TSA and VPA significantly repressed the proliferation of chondrosarcoma cells in a concentration-dependent manner. Flow cytometry indicated that TSA arrested the cell cycle in G2/M phase and VPA arrested the cell cycle in G1 phase. The tumor growth was markedly suppressed in mice treated with TSA and VPA. HDACIs significantly repress the proliferation of chondrosarcoma cells in vitro and in vivo. Our findings imply that HDACIs may provide a novel therapeutic target for the treatment of chondrosarcoma.

  9. Utility of Social Modeling for Proliferation Assessment - Preliminary Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Coles, Garill A.; Gastelum, Zoe N.; Brothers, Alan J.; Thompson, Sandra E.

    2009-06-01

    This Preliminary Assessment draft report will present the results of a literature search and preliminary assessment of the body of research, analysis methods, models and data deemed to be relevant to the Utility of Social Modeling for Proliferation Assessment research. This report will provide: 1) a description of the problem space and the kinds of information pertinent to the problem space, 2) a discussion of key relevant or representative literature, 3) a discussion of models and modeling approaches judged to be potentially useful to the research, and 4) the next steps of this research that will be pursued based on this preliminary assessment. This draft report represents a technical deliverable for the NA-22 Simulations, Algorithms, and Modeling (SAM) program. Specifically this draft report is the Task 1 deliverable for project PL09-UtilSocial-PD06, Utility of Social Modeling for Proliferation Assessment. This project investigates non-traditional use of social and cultural information to improve nuclear proliferation assessment, including nonproliferation assessment, proliferation resistance assessments, safeguards assessments and other related studies. These assessments often use and create technical information about the State’s posture towards proliferation, the vulnerability of a nuclear energy system to an undesired event, and the effectiveness of safeguards. This project will find and fuse social and technical information by explicitly considering the role of cultural, social and behavioral factors relevant to proliferation. The aim of this research is to describe and demonstrate if and how social science modeling has utility in proliferation assessment.

  10. JPRS report. Proliferation issues

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1992-07-10

    This report contains foreign media information on issues related to worldwide proliferation and transfer activities in nuclear, chemical, and biological weapons, including delivery systems and the transfer of weapons-relevant technologies. Foreign Broadcast Information Service (FBIS) and Joint Publications Research Service (JPRS) publications contain political, military, economic, environmental, and sociological news, commentary, and other information, as well as scientific and technical data and reports. All information has been obtained from foreign radio and television broadcasts, news agency transmissions, newspapers, books, and periodicals. Items generally are processed from the first or best available sources. It should not be inferred that they have been disseminated only in the medium, in the language, or to the area indicated. Items from foreign language sources are translated; those from English-language sources are transcribed. Except for excluding certain diacritics, FBIS renders personal names and place-names in accordance with the romanization systems approved for U.S. Government publications by the U.S. Board of Geographic Names.

  11. JPRS report. Proliferation issues

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1992-03-13

    This report contains foreign media information on issues related to worldwide proliferation and transfer activities in nuclear, chemical, and biological weapons, including delivery systems and the transfer of weapons-relevant technologies. Foreign Broadcast Information Service (FBIS) and Joint Publications Research Service (JPRS) publications contain political, military, economic, environmental, and sociological news, commentary, and other information, as well as scientific and technical data and reports. All information has been obtained from foreign radio and television broadcasts. news agency transmissions, newspapers, books, and periodicals. Items generally are processed from the first or best available sources. It should not be inferred that they have been disseminated only in the medium, in the language, or to the area indicated. Items from foreign language sources are translated; those from English-language sources are transcribed. Except for excluding certain diacritics, FBIS renders personal names and place-names in accordance with the romanization systems approved for U.S. Government publications by the U.S. Board of Geographic Names.

  12. JPRS report. Proliferation issues

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1992-07-16

    This report contains foreign media information on issues related to worldwide proliferation and transfer activities in nuclear, chemical, and biological weapons, including delivery systems and the transfer of weapons-relevant technologies. Foreign Broadcast Information Service (FBIS) and Joint Publications Research Service (JPRS) publications contain political, military, economic, environmental, and sociological news, commentary, and other information, as well as scientific and technical data and reports. All information has been obtained from foreign radio and television broadcasts, news agency transmissions, newspapers, books, and periodicals. Items generally are processed from the first or best available sources. It should not be inferred that they have been disseminated only in the medium, in the language, or to the area indicated. Items from foreign language sources are translated; those from English-language sources are transcribed. Except for excluding certain diacritics, FBIS renders personal names and place-names in accordance with the romanization systems approved for U.S. Government publications by the U.S. Board of Geographic Names.

  13. JPRS report. Proliferation issues

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1992-02-21

    This report contains foreign media information on issues related to worldwide proliferation and transfer activities in nuclear, chemical, and biological weapons, including delivery systems and the transfer of weapons-relevant technologies. Foreign Broadcast Information Service (FBIS) and Joint Publications Research Service (JPRS) publications contain political, military, economic, environmental, and sociological news, commentary, and other information, as well as scientific and technical data and reports. All information has been obtained from foreign radio and television broadcasts, news agency transmissions, newspapers, books, and periodicals. Items generally are processed from the first or best available sources. It should not be inferred that they have been disseminated only in the medium, in the language, or to the area indicated. Items from foreign language sources are translated; those from English-language sources are transcribed. Except for excluding certain diacritics, FBIS renders personal names and place-names in accordance with the romanization systems approved for U.S. Government publications by the U.S. Board of Geographic Names.

  14. Simplified system for the pressure control of a Nucleo electric central of the BWR type; Sistema simplificado para el control de presion de una central Nucleoelectrica del tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez J, J. [FI-UNAM, DEPFI Campus Morelos, Jiutepec, Morelos (Mexico)

    2003-07-01

    One of the main preoccupations of the electric power generator stations is the appropriate operation of the same ones. The operators must be qualified to respond in an adequate way and to be able to take to these power stations to an optimal, sure and stable operation condition under any circumstance. The Laboratory of Analysis in Nuclear Reactors Engineering (LAIRN) of the Engineering Faculty of UNAM (Fl) in collaboration with the International Atomic Energy Agency (IAEA), it develops an interactive classroom simulator in which simulations of the phenomena which take place in a nuclear power station are executed. The classroom simulator bases its operation on specialized nuclear codes feeding interactive graphic unfolding with those that it is possible to make a monitoring, supervision and control of the behavior of the power station under any operation regime, either in normal operation, transitory events or postulated accident sequence. The development of this classroom simulator includes a modular and re configurable structure. Due to it is indispensable to count with a higher inter activity with the system it is included the simulation of the control system of the plant and inside the same, one of those more important it is the reactor pressure control system. The present work describes the conceptual design and the used methodology for the development and implementation in the simulator of a simplified model of the pressure control system for a BWR generic central. The reach of the development will allow to accomplish the necessary tests to demonstrate that this has an adequate performance according to the carried out simplifications. (Author)

  15. Stability analysis of a recycling circuit of a BWR type reactor. Theoretical study; Analisis de estabilidad de un circuito de recirculacion de un reactor del tipo BWR. Estudio teorico

    Energy Technology Data Exchange (ETDEWEB)

    Salinas H, J.G.; Espinosa P, G. [Universidad Autonoma Metropolitana-Iztapalapa, 09000 Mexico D.F. (Mexico); Gonzalez M, V.M. [Comision Nacional de Seguridad Nuclear y Salvaguardias, 04000 Mexico D.F. (Mexico)

    2000-07-01

    The Technology, Regulation and Services Management of the National Commission of Nuclear Safety and Safeguards financed and in coordinate form with the I.P.H. Department of the Metropolitan Autonomous-Iztapalapa University developed the present project with the purpose of studying the effect of the recycling system on the linear stability of a BWR reactor whose reference central is the Laguna Verde power station. The present project forms part of a work series focused to the linear stability of the nuclear reactor of the Unit 1 at Laguna Verde power station. The components of the recycling system considered for the study of stability are the recycling external circuit (recycling pumps, valves) and the internal circuit (downcomer, jet pumps, lower full, driers, separators). The mathematical model is obtained applying mass balances and movement quantity in each one of the mentioned circuits. With respect to the nucleus model two regions are considered, the first one is made of a flow in one phase and the second one of a flow in two phases. For modelling the biphasic region it is considered homogenous flow. Generally it is studied the system behavior in the frequency domain starting from the transfer function applied to four operational states which correspond to the lower stability zone in the map power-flow of the Unit 1 of Laguna Verde power station. The Nyquist diagrams corresponding to each state as well as their characteristic frequency were determined. The results show that exists a very clear dependence of the power-flow relation on the stability of the system. It was found that the boiling length is an important parameter for the linear stability of the system. The obtained results show that the characteristic frequencies in unstability zones are similar to the reported data of the Unit 1 of the Laguna Verde power station in the event of power oscillations carried out in January 1995. (Author)

  16. Study of environmental noise in a BWR plant like the Nuclear Power Plant Laguna Verde; Estudio de ruido ambiental en una planta BWR como la Central Nuclear Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Tijerina S, F.; Cruz G, M.; Amador C, C., E-mail: francisco.tijerina@cfe.gob.mx [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Subgerencia de Ingenieria, Carretera Cardel-Nautla Km. 42.5, Alto Lucero, Veracruz (Mexico)

    2013-10-15

    In all industry type the health costs generated by the noise are high, because the noise can cause nuisance and to harm the capacity to work when causing tension and to perturb the concentration, and in more severe cases to reach to lose the sense of the hearing in the long term. The noise levels in the industry have been designated for the different types of use like residential, commercial, and industrial and silence areas. The noise can cause accidents when obstructing the communications and alarm signs. For this reason the noise should be controlled and mitigated, at a low level as reasonably is possible, taking into account that the noise is an acoustic contamination. The present study determines a bases line of the environmental noise levels in a nuclear power plant BWR-5 as Laguna Verde, (like reference) to be able to determine and to give pursuit to the possible solutions to eliminate or to limit the noise level in the different job areas. The noise levels were registered with a meter of integrative noise level (sonometer) and areas of noise exposure levels mapping the general areas in the buildings were established, being the registered maximum level of 96.94 dba in the building of the Reactor-elevation 0.65 m under the operation conditions of Extended Power Up rate (EPU) of 120% PTN. Knowing that the exposition to noises and the noise dose in the job place can influence in the health and in the safety of the workers, are extensive topics that they should be analyzed for separate as they are: to) the effects in the health of the exposure to the noise, b) how measuring the noise, c) the methods and technologies to combat and to control the noise in the industry by part of engineering area and d) the function of the industrial safety bodies as delegates of the health and safety in the task against the noise in the job. (author)

  17. Parametric and experimentally informed BWR Severe Accident Analysis Utilizing FeCrAl - M3FT-17OR020205041

    Energy Technology Data Exchange (ETDEWEB)

    Ott, Larry J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howell, Michael [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-08-01

    Iron-chromium-aluminum (FeCrAl) alloys are being considered as advanced fuel cladding concepts with enhanced accident tolerance. At high temperatures, FeCrAl alloys have slower oxidation kinetics and higher strength compared with zirconium-based alloys. FeCrAl could be used for fuel cladding and spacer or mixing vane grids in light water reactors and/or as channel box material in boiling water reactors (BWRs). There is a need to assess the potential gains afforded by the FeCrAl accident-tolerant-fuel (ATF) concept over the existing zirconium-based materials employed today. To accurately assess the response of FeCrAl alloys under severe accident conditions, a number of FeCrAl properties and characteristics are required. These include thermophysical properties as well as burst characteristics, oxidation kinetics, possible eutectic interactions, and failure temperatures. These properties can vary among different FeCrAl alloys. Oak Ridge National Laboratory has pursued refined values for the oxidation kinetics of the B136Y FeCrAl alloy (Fe-13Cr-6Al wt %). This investigation included oxidation tests with varying heating rates and end-point temperatures in a steam environment. The rate constant for the low-temperature oxidation kinetics was found to be higher than that for the commercial APMT FeCrAl alloy (Fe-21Cr-5Al-3Mo wt %). Compared with APMT, a 5 times higher rate constant best predicted the entire dataset (root mean square deviation). Based on tests following heating rates comparable with those the cladding would experience during a station blackout, the transition to higher oxidation kinetics occurs at approximately 1,500°C. A parametric study varying the low-temperature FeCrAl oxidation kinetics was conducted for a BWR plant using FeCrAl fuel cladding and channel boxes using the MELCOR code. A range of station blackout severe accident scenarios were simulated for a BWR/4 reactor with Mark I containment. Increasing the FeCrAl low-temperature oxidation rate

  18. PMP27 PROMOTES PEROXISOMAL PROLIFERATION

    NARCIS (Netherlands)

    MARSHALL, PA; KRIMKEVICH, YI; LARK, RH; DYER, JM; VEENHUIS, M; GOODMAN, JM; Krimkevich, Yelena I.; Lark, Richard H.; Dyer, John M.; Goodman, Joel M.

    Peroxisomes perform many essential functions in eukaryotic cells. The weight of evidence indicates that these organelles divide by budding from preexisting peroxisomes. This process is not understood at the molecular level. Peroxisomal proliferation can be induced in Saccharomyces cerevisiae by

  19. Cell Proliferation and Cytotoxicity Assays.

    Science.gov (United States)

    Adan, Aysun; Kiraz, Yağmur; Baran, Yusuf

    Cell viability is defined as the number of healthy cells in a sample and proliferation of cells is a vital indicator for understanding the mechanisms in action of certain genes, proteins and pathways involved cell survival or death after exposing to toxic agents. Generally, methods used to determine viability are also common for the detection of cell proliferation. Cell cytotoxicity and proliferation assays are generally used for drug screening to detect whether the test molecules have effects on cell proliferation or display direct cytotoxic effects. Regardless of the type of cell-based assay being used, it is important to know how many viable cells are remaining at the end of the experiment. There are a variety of assay methods based on various cell functions such as enzyme activity, cell membrane permeability, cell adherence, ATP production, co-enzyme production, and nucleotide uptake activity. These methods could be basically classified into different categories: (I) dye exclusion methods such as trypan blue dye exclusion assay, (II) methods based on metabolic activity, (III) ATP assay, (IV) sulforhodamine B assay, (V) protease viability marker assay, (VI) clonogenic cell survival assay, (VII) DNA synthesis cell proliferation assays and (V) raman micro-spectroscopy. In order to choose the optimal viability assay, the cell type, applied culture conditions, and the specific questions being asked should be considered in detail. This particular review aims to provide an overview of common cell proliferation and cytotoxicity assays together with their own advantages and disadvantages, their methodologies, comparisons and intended purposes.

  20. BWR stability analysis: methodology of the stability analysis and results of PSI for the NEA/NCR benchmark task; SWR Stabilitaetsanalyse: Methodik der Stabilitaetsanalyse und PSI-Ergebnisse zur NEA/NCR Benchmarkaufgabe

    Energy Technology Data Exchange (ETDEWEB)

    Hennig, D.; Nechvatal, L. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1996-09-01

    The report describes the PSI stability analysis methodology and the validation of this methodology based on the international OECD/NEA BWR stability benchmark task. In the frame of this work, the stability properties of some operation points of the NPP Ringhals 1 have been analysed and compared with the experimental results. (author) figs., tabs., 45 refs.

  1. RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - user`s manual

    Energy Technology Data Exchange (ETDEWEB)

    Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.; Neymotin, L.Y.

    1998-03-01

    This document is the User`s Manual for the Boiling Water Reactor (BWR), and Simplified Boiling Water Reactor (SBWR) systems transient code RAMONA-4B. The code uses a three-dimensional neutron-kinetics model coupled with a multichannel, nonequilibrium, drift-flux, phase-flow model of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients. Chapter 1 gives an overview of the code`s capabilities and limitations; Chapter 2 describes the code`s structure, lists major subroutines, and discusses the computer requirements. Chapter 3 is on code, auxillary codes, and instructions for running RAMONA-4B on Sun SPARC and IBM Workstations. Chapter 4 contains component descriptions and detailed card-by-card input instructions. Chapter 5 provides samples of the tabulated output for the steady-state and transient calculations and discusses the plotting procedures for the steady-state and transient calculations. Three appendices contain important user and programmer information: lists of plot variables (Appendix A) listings of input deck for sample problem (Appendix B), and a description of the plotting program PAD (Appendix C). 24 refs., 18 figs., 11 tabs.

  2. Strain-induced corrosion cracking in ferritic components of BWR primary circuits; Risskorrosion in druckfuehrenden ferritischen Komponenten des Primaerkreislaufes von Siedewasserreaktoren

    Energy Technology Data Exchange (ETDEWEB)

    Seifert, H.-P.; Ritter, S.; Ineichen, U.; Tschanz, U.; Gerodetti, B

    2003-04-01

    The present final report of the RIKORR project is a summary of a literature survey and of the experimental work performed by PSI on the environmentally-assisted cracking (EAC) and dynamic strain ageing (DSA) susceptibility of low-alloy steels (LAS) in high-temperature (HT) water. Within this project, the EAC crack growth behaviour of different low-alloy RPV steels, weld filler and weld heat-affected zone materials has been investigated under simulated transient and steady-state BWR/NWC power operation conditions. The strain-induced corrosion cracking (SICC) / low-frequency corrosion fatigue (CF) and stress corrosion cracking (SCC) crack growth behaviour of different low-alloy RPV steels under simulated transient and stationary BWR/NWC conditions was characterized by slow rising load / low-frequency corrosion fatigue and constant load / periodical partial unloading / ripple load tests with pre-cracked fracture mechanics specimens in oxygenated HT water at temperatures of either 288, 250, 200 or 150 {sup o}C. Modern high-temperature water loops, on-line crack growth monitoring and fractographic analysis by scanning electron microscopy (SEM) were used to quantify the cracking response. (author)

  3. Concept of a self-sustaining cooling system for after-heat removal in BWR-type reactors; Konzept eines autarken Kuehlsystems zur Nachwaermeabfuhr in Siedewasserreaktoren

    Energy Technology Data Exchange (ETDEWEB)

    Venker, J. [RWE Technology GmbH, Essen (Germany). Nukleartechnologie; Stuttgart Univ. (Germany). Inst. fuer Kernenergetik und Energiesysteme (IKE); Lavante, D. von [TUEV Rheinland, Koeln (Germany); Buck, M.; Starflinger, J. [Stuttgart Univ. (Germany). Inst. fuer Kernenergetik und Energiesysteme (IKE); Gitzel, D. [RWE Technology GmbH, Essen (Germany). Nukleartechnologie

    2013-07-01

    The concept, technical feasibility and potential capability of a new self-sustaining after-heat removal system based on supercritical carbon dioxide is described. The effect of the system on the plant behavior of appropriately retrofitted BWR-type reactors is discussed. Based on calculations using the thermal hydraulic code ATHLET it is shown that the safe after-heat removal time of existing BWR-type reactors in case of station blackout can be increased for several hours. The calculations have also shown that a enduring control of the station blackout situation cannot be reached by the retrofitting of the pressure relief system. The question is raised whether the pressure relief is reasonable independent of the accident scenario. Without the possibility of further coolant supply in case of station blackout the pressure relief will enhance the dry-out of the reactor core. The high-pressure path for the primary circuit increases the time for possible external measures to activate ECCS or active after-heat removal.

  4. Fuel design with low peak of local power for BWR reactors with increased nominal power; Diseno de un combustible con bajo pico de potencia local para reactores BWR con potencia nominal aumentada

    Energy Technology Data Exchange (ETDEWEB)

    Perusquia C, R.; Montes, J.L.; Hernandez, J.L.; Ortiz, J.J.; Castillo, A. [ININ, 52750 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: mrpc@nuclear.inin.mx

    2006-07-01

    The Federal Commission of Electricity recently announcement the beginning of the works related with the increase of the power to 120% of the original nominal one in the Boiling Water Reactors (BWR) of the Laguna Verde Central (CLV): In the National Institute of Nuclear Research (ININ) are carried out studies of the impact on the design of the recharge of derived fuel of this increase. One of the main effects of the power increase type that it is promoting, is the increment of the flow of generated vapor, what takes, to a bigger fraction of vacuum in the core presenting increased values of the maximum fraction to the limit, so much of the ratio of lineal heat generation (XFLPD) as of the ratio of critic power (MFLCPR). In the made studies, it is found that these fractions rise lineally with the increase of the nominal power. Considering that the reactors of the CLV at the moment operate to 105% of the original nominal power, it would imply an increment of the order of 13.35% in the XFLPD and in the MFLCPR operating to a nominal power of 120% of the original one. This would propitiate bigger problems to design appropriately the fuel cycle and the necessity, almost unavoidable, of to resort to a fuel assembly type more advanced for the recharges of the cores. As option, in the ININ the feasibility of continuing using the same type of it fuel assembles that one has come using recently in the CLV, the type GE12 is analyzed. To achieve it was outlined to diminish the peak factor of local power (LPPF) of the power cells that compose the fuel recharge in 13.35%. It was started of a fuel design previously used in the recharge of the unit 1 cycle 12 and it was re-design to use it in the recharge design of the cycle 13 of the unit 1, considering an increase to 120% of the original power and the same requirements of cycle extension. For the re-design of the fuel assembly cell it was used the PreDiCeldas computer program developed in the ININ. It was able to diminish the LPPF

  5. Cause-Effect relationship of the Laguna Verde BWR power instability by empirical mode decomposition; Relacion efecto-causa de la inestabilidad de potencia del BWR de Laguna Verde por descomposicion modal empirica

    Energy Technology Data Exchange (ETDEWEB)

    Blazquez, J.; Ruiz, J.; Castillo, R. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)

    2008-07-01

    The signals coming from natural phenomena are in essence non lineal and not stationary. A recent development, well-known as Empirical Mode Decomposition (EMD) it presents a novel focus that allows to represent in adaptative form non stationary signals as a sum of components of half zero. These components denominated Intrinsic Mode Functions (IMF) they help to the analysis of the frequency composition of unidimensional signals. The use of the EMD followed by the Hilbert transform of the IMFs it allows to carry out an analysis in time-frequency of the non lineal and not stationary data. This technique is known as the Hilbert Huang Transform (HHT). In this work a power instability event occurred in January 24, 1995 in the unit I of the nuclear power station of Laguna Verde (Mexico), corresponding to a BWR/5 is analyzed. When a Nuclear Plant suffers a power instability event, it is required obligatorily to explain to the Regulator Organism the effects and the causes of the event. The effects are described simply; not in vain there is a registration of signals in the Process Computer of where the required information is extracted. But the causes are not always immediate and easy for to identify. The power instability can happen during the start, when the refrigeration flow is relatively low in front to the power. By reason of that the reactivity coefficient by holes is negative, the power oscillates with a very defined frequency, generally of the order of 0.5 Hz. If the oscillations increase progressively of amplitude, we are in an instability event. It is interesting to include in the report the instant in that the began instability and the actions of the operator before and after the same one. As the actions are registered, the investigation is focused toward the instant of the beginning to be able to identify them. In this work the power signal in five empiric ways of Hilbert-Huang and a residual breaks down. The instability is only reflected in the way of smaller

  6. Decay profiles of {beta} and {gamma} for a radionuclide inventory in equilibrium cycle of a BWR type reactor; Perfiles de decaimiento de radiacion {beta} y {gamma} para un inventario de radionuclidos en ciclo de equilibrio de un reactor tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Salaices, M.; Sandoval, S.; Ovando, R. [Instituto de Investigaciones Electricas. Gerencia de Energia Nuclear, Av. Reforma 113 Col. Palmira. 62490 Cuernavaca, Morelos (Mexico)]. e-mail: sal@iie.org.mx

    2007-07-01

    Presently work the {beta} and {gamma} radiation decay profiles for a radionuclides inventory in equilibrium cycle of a BWR type reactor is presented. The profiles are presented in terms of decay in the activity of the total inventory as well as of the chemical groups that conform the inventory. In the obtaining of the radionuclides inventory in equilibrium cycle the ORIGEN2 code, version 1 was used, which simulates fuel burnup cycles and it calculates the evolution of the isotopic composition as a result of the burnt one, irradiation and decay of the nuclear fuel. It can be observed starting from the results that the decrease in the activity for the initial inventory and the different chemical groups that conform it is approximately proportional to the base 10 logarithm of the time for the first 24 hours of having concluded the burnt one. It can also be observed that the chemical groups that contribute in more proportion to the total activity of the inventory are the lanthanides-actinides and the transition metals, with 39% and 28%, respectively. The groups of alkaline earth metals, halogens, metalloids, noble gases and alkaline metals, contribute with percentages that go from the 8 to 5%. The groups that less they contribute to the total activity of the inventory they are the non metals and semi-metals with smaller proportions that 1%. The chemical groups that more contribute to the energy of {beta} and {gamma} radiation its are the transition metals and the lanthanides-actinides with a change in the order of importance at the end of the 24 hours period. The case of the halogens is of relevance for the case of the {gamma} radiation energy due that occupying the very near third site to the dimensions of the two previous groups. Additionally, the decay in the activity for the total inventory and the groups that conform it can be simulated by means of order 6 polynomials or smaller than describe its behavior appropriately. The results presented in this work, coupled

  7. Performance of the primary containment of a BWR during a severe accident whit the code RELAP/SCDAPSIM; Comportamiento del contenedor primario de un reactor BWR durante un accidente severo con el codigo RELAP/SCDAPSIM

    Energy Technology Data Exchange (ETDEWEB)

    Castillo G, F.

    2015-07-01

    In this thesis work, it was developed a model of the vacuum breaker valves and down comers for a BWR Mark II primary containment for the code RELAP/SCDAPSIM Mod. 3.4. This code was used to simulate a Station Blackout (Sbo) that evolves to a severe accident scenario. To accomplish this task, the vacuum breaker valves and down comers were included in a simplified model of the primary containment that includes both wet well and dry well, which was coupled with a model of the Nuclear Steam Supply System (NSSS), in order to study the behavior of the primary containment during the evolution of the accident scenario. In the analysis of the results of the simulation, the behavior of the wet well and dry well during the event was particularly monitored, by analyzing the evolution of temperature and pressure profiles in such volumes, this to determine the impact of the inclusion of the breaker vacuum valves and down comers. The results show that the effect of this extension of the model is that more conservative results are obtained, i.e., higher pressures are reached in both wet well and dry well than when it is used a containment model that does not include neither the vacuum valves nor the down comers. The most relevant results obtained show that the Rcic alone is able to keep the core fully covered, but even in such a case, it evaporates about 15% of the initial inventory of liquid water in the Pressure Suppression Pool (Psp). When the Rcic operation is lost, 20% more of the liquid water inventory in the Psp is further reduced within four to twelve hours (approximately), time at which the simulation crashed. Besides, there is a significant increase of pressure in the containment. As the accident evolves, the pressure in the containment continues increasing, but there is still considerable margin to reach the design pressure of the containment. At the end of the simulation, the results show a gauge pressure value of 224,550 Pa in the Psp and 187,482 Pa in the wet well

  8. Carbon 14 distribution in irradiated BWR fuel cladding and released carbon 14 after aqueous immersion of 6.5 years

    Energy Technology Data Exchange (ETDEWEB)

    Sakuragi, T. [Radioactive Waste Management Funding and Research Center, Tsukishima 1-15-7, Chuo City, Tokyo, 104-0052 (Japan); Yamashita, Y.; Akagi, M.; Takahashi, R. [TOSHIBA Corporation, Ukishima Cho 4-1, Kawasaki Ward, Kawasaki, 210-0862 (Japan)

    2016-07-01

    Spent fuel cladding which is highly activated and strongly contaminated is expected to be disposed of in an underground repository. A typical activation product in the activated metal waste is carbon 14 ({sup 14}C), which is mainly generated by the {sup 14}N(n,p){sup 14}C reaction and produces a significant exposure dose due to the large inventory, long half-life (5730 years), rapid release rate, and the speciation and consequent migration parameters. In the preliminary Japanese safety case, the release of radionuclides from the metal matrix is regarded as the corrosion-related congruent release, and the cladding oxide layer is regarded as a source of instant release fraction (IRF). In the present work, specific activity of {sup 14}C was measured using an irradiated BWR fuel cladding (Zircaloy-2, average rod burnup of 41.6 GWd/tU) which has an external oxide film having a thickness of 25.3 μm. The {sup 14}C specific activity of the base metal was 1.49*10{sup 4} Bq/g, which in the corresponding burnup is comparable to values in the existing literature, which were obtained from various irradiated claddings. Although the specific activity in oxide was 2.8 times the base metal activity due to the additive generation by the {sup 17}O(n,α){sup 14}C reaction, the {sup 14}C abundance in oxide was less than 10% of total inventory. A static leaching test using the cladding tube was carried out in an air-tight vessel filled with a deoxygenated dilute NaOH solution (pH of 12.5) at room temperature. After 6.5 years, {sup 14}C was found in each leachate fraction of gas phase and dissolved organics and inorganics, the total of which was less than 0.01% of the {sup 14}C inventory of the immersed cladding tube. A simple calculation based on the congruent release with Zircaloy corrosion has suggested that the 96.7% of released {sup 14}C was from the external oxide layer and 3.3% was from the base Zircaloy metal. However, both the {sup 14}C abundance and the low leaching rate

  9. Implementation of a methodology to perform the uncertainty and sensitivity analysis of the control rod drop in a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Reyes F, M. del C.

    2015-07-01

    A methodology to perform uncertainty and sensitivity analysis for the cross sections used in a Trace/PARCS coupled model for a control rod drop transient of a BWR-5 reactor was implemented with the neutronics code PARCS. A model of the nuclear reactor detailing all assemblies located in the core was developed. However, the thermohydraulic model designed in Trace was a simple model, where one channel representing all the types of assemblies located in the core, it was located inside a simple vessel model and boundary conditions were established. The thermohydraulic model was coupled with the neutronics model, first for the steady state and then a Control Rod Drop (CRD) transient was performed, in order to carry out the uncertainty and sensitivity analysis. To perform the analysis of the cross sections used in the Trace/PARCS coupled model during the transient, Probability Density Functions (PDFs) were generated for the 22 parameters cross sections selected from the neutronics parameters that PARCS requires, thus obtaining 100 different cases for the Trace/PARCS coupled model, each with a database of different cross sections. All these cases were executed with the coupled model, therefore obtaining 100 different outputs for the CRD transient with special emphasis on 4 responses per output: 1) The reactivity, 2) the percentage of rated power, 3) the average fuel temperature and 4) the average coolant density. For each response during the transient an uncertainty analysis was performed in which the corresponding uncertainty bands were generated. With this analysis it is possible to observe the results ranges of the responses chose by varying the uncertainty parameters selected. This is very useful and important for maintaining the safety in the nuclear power plants, also to verify if the uncertainty band is within of safety margins. The sensitivity analysis complements the uncertainty analysis identifying the parameter or parameters with the most influence on the

  10. Solution of the transport equation in stationary state, in one and two dimensions, for BWR assemblies using nodal methods; Solucion de la ecuacion de transporte en estado estacionario, en 1 y 2 dimensiones, para ensambles tipo BWR usando metodos nodales

    Energy Technology Data Exchange (ETDEWEB)

    Xolocostli M, J.V

    2002-07-01

    The main objective of this work is to solve the neutron transport equation in one and two dimensions (slab geometry and X Y geometry, respectively), with no time dependence, for BWR assemblies using nodal methods. In slab geometry, the nodal methods here used are the polynomial continuous (CMPk) and discontinuous (DMPk) families but only the Linear Continuous (also known as Diamond Difference), the Quadratic Continuous (QC), the Cubic Continuous (CC), the Step Discontinuous (also known as Backward Euler), the Linear Discontinuous (LD) and the Quadratic Discontinuous (QD) were considered. In all these schemes the unknown function, the angular neutron flux, is approximated as a sum of basis functions in terms of Legendre polynomials, associated to the values of the neutron flux in the edges (left, right, or both) and the Legendre moments in the cell, depending on the nodal scheme used. All these schemes were implemented in a computer program developed in previous thesis works and known with the name TNX. This program was modified for the purposes of this work. The program discreetizes the domain of concern in one dimension and determines numerically the angular neutron flux for each point of the discretization when the number of energy groups and regions are known starting from an initial approximation for the angular neutron flux being consistent with the boundary condition imposed for a given problem. Although only problems with two-energy groups were studied the computer program does not have limitations regarding the number of energy groups and the number of regions. The two problems analyzed with the program TNX have practically the same characteristics (fuel and water), with the difference that one of them has a control rod. In the part corresponding to two-dimensional problems, the implemented nodal methods were those designated as hybrids that consider not only the edge and cell Legendre moments, but also the values of the neutron flux in the corner points

  11. Domestic Politics and Nuclear Proliferation

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chul Min; Yim, Man Sung [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    The external security threat is known as the most important factor of nuclear weapons program, the domestic politics situation can also affect the nuclear proliferation decision of a country. For example, when a leader wants nuclear weapons as an ultimate weapon, the domestic politics situation can determine the effectiveness of the weapons program of a country. This study analyzes the current knowledge of the relationship between domestic politics and nuclear proliferation and suggests the main challenges of the quantitative models trying to calculate nuclear proliferation risk of countries. The domestic politics status is one of the most important indicators of nuclear program. However, some variables have never been used in quantitative analyses; for example, number of veto players and the public opinion on nuclear weapons; despite they are considered to be important in various qualitative studies. Future studies should focus on how should they be coded and how can they be linked with existing domestic politics variables.

  12. Calcium signaling and cell proliferation.

    Science.gov (United States)

    Pinto, Mauro Cunha Xavier; Kihara, Alexandre Hiroaki; Goulart, Vânia A M; Tonelli, Fernanda M P; Gomes, Katia N; Ulrich, Henning; Resende, Rodrigo R

    2015-11-01

    Cell proliferation is orchestrated through diverse proteins related to calcium (Ca(2+)) signaling inside the cell. Cellular Ca(2+) influx that occurs first by various mechanisms at the plasma membrane, is then followed by absorption of Ca(2+) ions by mitochondria and endoplasmic reticulum, and, finally, there is a connection of calcium stores to the nucleus. Experimental evidence indicates that the fluctuation of Ca(2+) from the endoplasmic reticulum provides a pivotal and physiological role for cell proliferation. Ca(2+) depletion in the endoplasmatic reticulum triggers Ca(2+) influx across the plasma membrane in an phenomenon called store-operated calcium entries (SOCEs). SOCE is activated through a complex interplay between a Ca(2+) sensor, denominated STIM, localized in the endoplasmic reticulum and a Ca(2+) channel at the cell membrane, denominated Orai. The interplay between STIM and Orai proteins with cell membrane receptors and their role in cell proliferation is discussed in this review. Copyright © 2015 Elsevier Inc. All rights reserved.

  13. Study of instabilities in phase by using the tool {sup D}ynamics{sup :} analysis of the evolution space temporary of the waves of density in channels of reactors BWR; Estudio de las Inestabilidades en Fase Mediante la Herramienta Dinamics: analisis de la Evolucion Espacio Temporal de las Ondas de Densidad en Canales de Reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Munoz-Cobo, J. L.; Escriva, R.; Merino, R.; Melara, J.

    2013-07-01

    This paper presents the basics of Dynamics V2 to code It allows calculations of stability for oscillations in phase in BWR reactors in the time domain. The equations of the model are exposed and is the integration of the equations. The model can be used in a large number of nodes thrust for the calculations to an acceptable computational cost, it has simplified dynamics of recirculation loop and the code has been incorporated the Oscillation in phase boundary conditions. The code incorporates the equations of boiling sub-cooled which allows to make more realistic calculations as well as subroutines to calculate the subroutines-based properties of the MATPRO and ASME.

  14. Performance of iron–chromium–aluminum alloy surface coatings on Zircaloy 2 under high-temperature steam and normal BWR operating conditions

    Energy Technology Data Exchange (ETDEWEB)

    Zhong, Weicheng; Mouche, Peter A.; Han, Xiaochun [University of Illinois, Department of Nuclear, Radiological, and Plasma Engineering, Urbana, IL 61801 (United States); Heuser, Brent J., E-mail: bheuser@illinois.edu [University of Illinois, Department of Nuclear, Radiological, and Plasma Engineering, Urbana, IL 61801 (United States); Mandapaka, Kiran K.; Was, Gary S. [University of Michigan, Department of Nuclear Engineering and Radiological Sciences, Ann Arbor, MI 48109 (United States)

    2016-03-15

    Iron-chromium-aluminum (FeCrAl) coatings deposited on Zircaloy 2 (Zy2) and yttria-stabilized zirconia (YSZ) by magnetron sputtering have been tested with respect to oxidation weight gain in high-temperature steam. In addition, autoclave testing of FeCrAl-coated Zy2 coupons under pressure-temperature-dissolved oxygen coolant conditions representative of a boiling water reactor (BWR) environment has been performed. Four different FeCrAl compositions have been tested in 700 °C steam; compositions that promote alumina formation inhibited oxidation of the underlying Zy2. Parabolic growth kinetics of alumina on FeCrAl-coated Zy2 is quantified via elemental depth profiling. Autoclave testing under normal BWR operating conditions (288 °C, 9.5 MPa with normal water chemistry) up to 20 days demonstrates observable weight gain over uncoated Zy2 simultaneously exposed to the same environment. However, no FeCrAl film degradation was observed. The 900 °C eutectic in binary Fe–Zr is addressed with the FeCrAl-YSZ system. - Graphical abstract: Weight gain normalized to total sample surface area versus time during 700 °C steam exposure for FeCrAl samples with different composition (A) and Fe/Cr/Al:62/4/34 (B). In both cases, the responses of uncoated Zry2 (Zry2-13A and Zry2-19A) are shown for comparison. This uncoated Zry2 response shows the expected pre-transition quasi-cubic kinetic behavior and eventual breakaway (linear) kinetics. Highlights: • FeCrAl coatings deposited on Zy2 have been tested with respect to oxidation in high-temperature steam. • FeCrAl compositions promoting alumina formation inhibited oxidation of Zy2 and delay weight gain. • Autoclave testing to 20 days of coated Zy2 in a simulated BWR environment demonstrates minimal weight gain and no film degradation. • The 900 °C eutectic in binary Fe-Zr is addressed with the FeCrAl-YSZ system.

  15. Dietary bovine lactoferrin increases intestinal cell proliferation in neonatal piglets.

    Science.gov (United States)

    Reznikov, Elizabeth A; Comstock, Sarah S; Yi, Cuiyi; Contractor, Nikhat; Donovan, Sharon M

    2014-09-01

    Lactoferrin is a bioactive milk protein that stimulates cell proliferation in vitro; however, limited in vivo evidence exists to allow lactoferrin to be incorporated into infant formula. Herein, the effect of dietary bovine lactoferrin (bLF) on neonatal intestinal growth and maturation was investigated guided by the hypothesis that bLF would increase cellular proliferation leading to functional differences in neonatal piglets. Colostrum-deprived piglets were fed formula containing 0.4 [control (Ctrl)], 1.0 (LF1), or 3.6 (LF3) g bLF/L for the first 7 or 14 d of life. To provide passive immunity, sow serum was provided orally during the first 36 h of life. Intestinal cell proliferation, histomorphology, mucosal DNA concentration, enzyme activity, gene expression, and fecal bLF content were measured. Intestinal enzyme activity, DNA concentration, and villus length were unaffected by bLF. However, crypt proliferation was 60% greater in LF1- and LF3-fed piglets than in Ctrl piglets, and crypt depth and area were 20% greater in LF3-fed piglets than in Ctrl piglets. Crypt cells from LF3-fed piglets had 3-fold higher β-catenin mRNA expression than did crypt cells from Ctrl piglets. Last, feces of piglets fed bLF contained intact bLF, suggesting that some bLF was resistant to digestion and could potentially affect intestinal proliferation through direct interaction with intestinal epithelial cells. This study is the first to our knowledge to show that dietary bLF stimulates crypt cell proliferation in vivo. The increased β-catenin expression indicates that Wnt signaling may in part mediate the stimulatory effect of bLF on intestinal cell proliferation. © 2014 American Society for Nutrition.

  16. Nuclear Proliferation: A Historical Overview

    Science.gov (United States)

    2008-03-01

    unsuccessful, however, and in 1981 Egypt ratified the Nuclear Non-Proliferation Treaty. In 1982 Egypt’s Hydrometallurgy Pilot Plant for reprocessing...Country Profile: Egypt,” http://www.iaea.org/DataCenter/index.html (accessed ɠ/8/2007>). 1982: Hydrometallurgy Pilot Plant for reprocessing

  17. Nuclear Proliferation Technology Trends Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Zentner, Michael D.; Coles, Garill A.; Talbert, Robert J.

    2005-10-04

    A process is underway to develop mature, integrated methodologies to address nonproliferation issues. A variety of methodologies (both qualitative and quantitative) are being considered. All have one thing in common, a need for a consistent set of proliferation related data that can be used as a basis for application. One approach to providing a basis for predicting and evaluating future proliferation events is to understand past proliferation events, that is, the different paths that have actually been taken to acquire or attempt to acquire special nuclear material. In order to provide this information, this report describing previous material acquisition activities (obtained from open source material) has been prepared. This report describes how, based on an evaluation of historical trends in nuclear technology development, conclusions can be reached concerning: (1) The length of time it takes to acquire a technology; (2) The length of time it takes for production of special nuclear material to begin; and (3) The type of approaches taken for acquiring the technology. In addition to examining time constants, the report is intended to provide information that could be used to support the use of the different non-proliferation analysis methodologies. Accordingly, each section includes: (1) Technology description; (2) Technology origin; (3) Basic theory; (4) Important components/materials; (5) Technology development; (6) Technological difficulties involved in use; (7) Changes/improvements in technology; (8) Countries that have used/attempted to use the technology; (9) Technology Information; (10) Acquisition approaches; (11) Time constants for technology development; and (12) Required Concurrent Technologies.

  18. anterior hyaloidal fibrovascular proliferation (ahfvp)

    African Journals Online (AJOL)

    Okonkwo

    fibrovascular proliferation in ischaemic diabetic eyes is known to occur predominantly in the region posterior to the equator, ie, the pre-equatorial fundus. There is a. 5 preponderance of posterior neovascularization occurring in proliferative diabetic retinopathy. This posterior. 7,8,9 proliferative disease in ischaemic diabetic ...

  19. Preliminary Assessment of the Possible BWR Core/Vessel Damage States for Fukushima Daiichi Station Blackout Scenarios Using RELAP/SCDAPSIM

    Directory of Open Access Journals (Sweden)

    C. M. Allison

    2012-01-01

    Full Text Available Immediately after the accident at Fukushima Daiichi, Innovative Systems Software and other members of the international SCDAP Development and Training Program started an assessment of the possible core/vessel damage states of the Fukushima Daiichi Units 1–3. The assessment included a brief review of relevant severe accident experiments and a series of detailed calculations using RELAP/SCDAPSIM. The calculations used a detailed RELAP/SCDAPSIM model of the Laguna Verde BWR vessel and related reactor cooling systems. The Laguna Verde models were provided by the Comision Nacional de Seguridad Nuclear y Salvaguardias, the Mexican nuclear regulatory authority. The initial assessment was originally presented to the International Atomic Energy Agency on March 21 to support their emergency response team and later to our Japanese members to support their Fukushima Daiichi specific analysis and model development.

  20. Calculations for a BWR Lattice with Adjacent Gadolinium Pins Using the Monte Carlo Cell Code Serpent v.1.1.7

    Directory of Open Access Journals (Sweden)

    Diego Ferraro

    2011-01-01

    Full Text Available Monte Carlo neutron transport codes are usually used to perform criticality calculations and to solve shielding problems due to their capability to model complex systems without major approximations. However, these codes demand high computational resources. The improvement in computer capabilities leads to several new applications of Monte Carlo neutron transport codes. An interesting one is to use this method to perform cell-level fuel assembly calculations in order to obtain few group constants to be used on core calculations. In the present work the VTT recently developed Serpent v.1.1.7 cell-oriented neutronic calculation code is used to perform cell calculations of a theoretical BWR lattice benchmark with burnable poisons, and the main results are compared to reported ones and with calculations performed with Condor v.2.61, the INVAP's neutronic collision probability cell code.

  1. Comprehensive assessment of the Ispra BWR and PWR subchannel experiments and code analysis with different two-phase models and solution schemes

    Energy Technology Data Exchange (ETDEWEB)

    Wolf, L.; Fischer, K.; Herkenrath, H.; Hufschmidt, W.

    1987-02-01

    This paper presents subchannel code results for a selection of the subchannel experiments performed at the Joint Research Centre of the CEC at Ispra with 16-rod BWR and PWR bundles at prototypical conditions. Three different well-known subchannel codes (COBRA-IIIC, CANAL, THERMIT) have been applied, each of which is representative of widely used two-phase flow methodologies (homogeneous, drift-flux, 2-fluid modelling) and numerical solution schemes. Their degrees of agreement with the measured subchannel exit mass flow and enthalpy distributions are reported as well as their limitations due to the lack of details in modelling two-phase flow rod bundle transport phenomena and/or numerical solution schemes both in transverse and axial directions. It is conclusively demonstrated that the extra expenditures of using advanced codes such as THERMIT effectively pay off when stringent requirements for optimizing LWR fuel rod bundles should be met.

  2. Cerenkov Characteristics of BWR Assemblies using a Prototype DCVD with a Back-Illuminated CCD. Prepared for the Canadian Safeguards Support Program and the Swedish Support Program

    Energy Technology Data Exchange (ETDEWEB)

    Chen, J.D.; Gerwing, A.F. [Channel Systems Inc., Pinawa MA (Canada); Maxwell, R. [Canadian Nuclear Safety Commission, Ottawa, ON (Canada); Larsson, M.; Axell, K.; Hildingsson, L. [Swedish Nuclear Power Inspectorate, Stockholm (Sweden); Lindberg, B. [LENS-TECH AB, Skellefteaa (Sweden); Sundkvist, E. [Teleca Design and Development, Stockholm (Sweden)

    2003-11-01

    The Canadian and Swedish Safeguards Support Programs have developed a prototype Digital Cerenkov Viewing Device (DCVD) to verify spent fuel. Field measurements in Swedish nuclear power reactor fuel bays on BWR fuel and non-fuel assemblies resulted in new Cerenkov information that offers the possibility of computer-assisted verification of spent-fuel assemblies. A number of fuel assemblies with missing fuel rods were examined. The missing fuel rods are easily detected when not hidden under the lifting handle of the fuel assembly. Initial studies of off-angle viewing of these assemblies show promise for the detection of the missing fuel rods under the lifting handle. The quantitative nature of the charge-coupled device was examined. A number of procedures were developed to quantify parameters such as image intensity and alignment (collimation) of fuel and no fuel assemblies. The quantitative studies on fuel assembly intensity as a function of cooling time showed excellent agreement with the theoretical calculations.

  3. Boiling water reactors with uranium-plutonium mixed oxide fuel. Report 5: Analysis of the reactivity coefficients and the stability of a BWR loaded with MOx fuel

    Energy Technology Data Exchange (ETDEWEB)

    Demaziere, C. [CEA Centre d' Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Reacteurs Nucleaires

    2000-01-01

    This report is a part of the project titled 'Boiling Water Reactors With Uranium-Plutonium Mixed Oxide (MOx) Fuel'. The aim of this study is to model the impact of a core loading pattern containing MOx bundles upon the main characteristics of a BWR (reactivity coefficients, stability, etc.). For this purpose, the Core Management System (CMS) codes of Studsvik Scandpower are used. This package is constituted by CASMO-4/TABLES-3/SIMULATE-3. It has been shown in previous reports that these codes are able to accurately represent and model MOx bundles. This report is thus devoted to the study of BWR cores loaded (partially or totally) with MOx bundles. The plutonium quality used is the Pu type 2016 (mostly Pu-239, 56 %, and Pu-240, 26 %), but a variation of the plutonium isotopic vector was also investigated, in case of a partial MOx loading. One notices that the reactivity coefficients do not present significant changes in comparison with a full UOx loading. Nevertheless, two main problems arise: the shutdown margin at BOC is lower than 1 % and the stability to in-phase oscillations is slightly decreased. (The SIMULATE-3 version used for this study does not contain the latest MOx enhancements described in literature, since these code developments have not been provided to the department. Nevertheless, as the nominal average enrichment of the MOx bundles is 5.41 % (total amount of plutonium), which can still be considered as a relatively low enrichment, the accuracy of the CMS codes is acceptable without the use of the MOx improvements for this level of Pu enrichment.

  4. Behavior to the fracture of an AISI 304 stainless steel sensitized in BWR reactor conditions (288 degrees Centigrade and 80 Kg/cm{sup 2}); Comportamiento a la fractura de un acero inoxidable AISI 304 sensibilizado en condiciones de reactor BWR (288 grados Centigrados y 80 Kg/cm{sup 2})

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez C, R.; Diaz S, A.; Garcia R, R.; Aguilar T, A.; Gachuz M, M.; Arganis J, C.; Merino C, J. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    1999-07-01

    It is a knew fact that ductility of a lot of structural alloys can be deteriorated by the environment effect which are exposed, and that their consequent embrittlement can put in doubt the safety of their functioning; such is the case of austenitic stainless steels used in internal components of the BWR type reactors which not only is subjected to the effect combined of the aggressive environment which surround it (pressure, temperature, corrosion potential, conductivity medium, local state of efforts, etc.), but also to the action of present neutron radiation, manifesting microstructural changes which are reflected in the augmentation of its susceptibility to the intergranular cracking, phenomena generally known as IASCC ''Irradiation Assisted Stress Corrosion Cracking''. Once appeared the cracking in the material, the useful life of a component is limited by the rapidity to growth of these cracking, making necessary evaluations which can to predict its behavior, therefore the present work shows the preliminary results for determining the behavior to the fracture of an AISI 304 stainless steel sensitized, in a dynamic recirculation circuit which allows to simulate the operation conditions of a BWR reactor (288 Centigrade and 80 kg/cm{sup 2}). (Author)

  5. Organizational Inertia and Excessive Product Proliferation

    OpenAIRE

    Sakuraki, Rie

    2015-01-01

    This study investigates the internal factors of excessive product proliferation. Since empirical literature on product over-proliferation focused on how to optimize existing product portfolio, the causes of excessive product proliferation have so far attracted little attention. This study employs a case study of Shiseido, a famous Japanese cosmetics company, with particular attention to product proliferation in the Shiseido chain store channel, because external factors are mostly absent from ...

  6. 15 CFR 12.2 - Undue proliferation.

    Science.gov (United States)

    2010-01-01

    ... 15 Commerce and Foreign Trade 1 2010-01-01 2010-01-01 false Undue proliferation. 12.2 Section 12.2 Commerce and Foreign Trade Office of the Secretary of Commerce FAIR PACKAGING AND LABELING § 12.2 Undue proliferation. (a) Information as to possible undue proliferation. Any person or group, including a State or...

  7. Climate Change, Nuclear Power and Nuclear Proliferation: Magnitude Matters

    Energy Technology Data Exchange (ETDEWEB)

    Robert J. Goldston

    2010-03-03

    Integrated energy, environment and economics modeling suggests electrical energy use will increase from 2.4 TWe today to 12 TWe in 2100. It will be challenging to provide 40% of this electrical power from combustion with carbon sequestration, as it will be challenging to provide 30% from renewable energy sources. Thus nuclear power may be needed to provide ~30% by 2100. Calculations of the associated stocks and flows of uranium, plutonium and minor actinides indicate that the proliferation risks at mid-century, using current light-water reactor technology, are daunting. There are institutional arrangements that may be able to provide an acceptable level of risk mitigation, but they will be difficult to implement. If a transition is begun to fast-spectrum reactors at mid-century, without a dramatic change in the proliferation risks of such systems, at the end of the century proliferation risks are much greater, and more resistant to mitigation. The risks of nuclear power should be compared with the risks of the estimated 0.64oC long-term global surface-average temperature rise predicted if nuclear power were replaced with coal-fired power plants without carbon sequestration. Fusion energy, if developed, would provide a source of nuclear power with much lower proliferation risks than fission.

  8. Climate Change, Nuclear Power and Nuclear Proliferation: Magnitude Matters

    Energy Technology Data Exchange (ETDEWEB)

    Robert J. Goldston

    2011-04-28

    Integrated energy, environment and economics modeling suggests that worldwide electrical energy use will increase from 2.4 TWe today to ~12 TWe in 2100. It will be challenging to provide 40% of this electrical power from combustion with carbon sequestration, as it will be challenging to provide 30% from renewable energy sources derived from natural energy flows. Thus nuclear power may be needed to provide ~30%, 3600 GWe, by 2100. Calculations of the associated stocks and flows of uranium, plutonium and minor actinides indicate that the proliferation risks at mid-century, using current light-water reactor technology, are daunting. There are institutional arrangements that may be able to provide an acceptable level of risk mitigation, but they will be difficult to implement. If a transition is begun to fast-spectrum reactors at mid-century, without a dramatic change in the proliferation risks of such systems, at the end of the century global nuclear proliferation risks are much greater, and more resistant to mitigation. Fusion energy, if successfully demonstrated to be economically competitive, would provide a source of nuclear power with much lower proliferation risks than fission.

  9. Proliferation of life from Enceladus

    Science.gov (United States)

    Czechowski, L.

    2017-09-01

    Enceladus is a medium-sized icy satellite (MIS) of Saturn. MIS are built of mixtures of rocks and ices. In 2014 [4] indicates that conditions in the core of this satellite allow for the life. In fact for hundreds of Myr the conditions in the interior of Enceladus were more favourable for origin of life than on the Earth [5, 6]. Presently we continue the research on the possible mechanism of life proliferation including additionally gravity assist as mechanism for deceleration of the body.

  10. Seismic resistance design of nuclear power plant building structures in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kitano, Takehito [Kansai Electric Power Co., Inc., Osaka (Japan)

    1997-03-01

    Japan is one of the countries where earthquakes occur most frequently in the world and has incurred a lot of disasters in the past. Therefore, the seismic resistance design of a nuclear power plant plays a very important role in Japan. This report describes the general method of seismic resistance design of a nuclear power plant giving examples of PWR and BWR type reactor buildings in Japan. Nuclear facilities are classified into three seismic classes and is designed according to the corresponding seismic class in Japan. Concerning reactor buildings, the short-term allowable stress design is applied for the S1 seismic load and it is confirmed that the structures have a safety margin against the S2 seismic load. (J.P.N.)

  11. Chemopreventive effects of metformin on obesity-associated endometrial proliferation.

    Science.gov (United States)

    Zhang, Qian; Celestino, Joseph; Schmandt, Rosemarie; McCampbell, Adrienne S; Urbauer, Diana L; Meyer, Larissa A; Burzawa, Jennifer K; Huang, Marilyn; Yates, Melinda S; Iglesias, David; Broaddus, Russell R; Lu, Karen H

    2013-07-01

    Obesity is a significant contributing factor to endometrial cancer risk. We previously demonstrated that estrogen-induced endometrial proliferation is enhanced in the context of hyperinsulinemia and insulin resistance. In this study, we investigate whether pharmacologic agents that modulate insulin sensitivity or normalize insulin levels will diminish the proliferative response to estrogen. Zucker fa/fa obese rats and lean controls were used as models of hyperinsulinemia and insulin resistance. Insulin levels were depleted in ovariectomized rats following treatment with streptozotocin, or modulated by metformin treatment. The number of BrdU-incorporated cells, estrogen-dependent proliferative and antiproliferative gene expression, and activation of mTOR and ERK1/2 MAPK signaling were studied. A rat normal endometrial cell line RENE1 was used to evaluate the direct effects of metformin on endometrial cell proliferation and gene expression in vitro. Streptozotocin lowered circulating insulin levels in obese rats and decreased the number of BrdU-labeled endometrial cells even in the presence of exogenous estrogen. Treatment with the insulin-sensitizing drug metformin attenuated estrogen-dependent proliferative expression of c-myc and c-fos in the obese rat endometrium compared to untreated controls and was accompanied by inhibition of phosphorylation of the insulin and IGF1 receptors (IRβ/IGF1R) and ERK1/2. In vitro studies indicated metformin inhibited RENE1 proliferation in a dose-dependent manner. These findings suggest that drugs that modulate insulin sensitivity, such as metformin, hinder estrogen-mediated endometrial proliferation. Therefore, these drugs may be clinically useful for the prevention of endometrial cancer in obese women. Copyright © 2013 Mosby, Inc. All rights reserved.

  12. Simulation of the BWR experiments CORA-17 and CORA-28 using ATHLET-CD and assessment of BWR modelling. 1{sup st} Technical report. Validation and interpretation of the ATHLET-CD model basis; Simulation der SWR-Versuche CORA-17 und CORA-28 mit dem Programmsystem ATHLET-CD und Bewertung der SWR-Modellbasis. 1. Technischer Fachbericht. Validierung und Interpretation der ATHLET-CD Modellbasis

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmann, M.; Gremme, F.; Koch, M.K.

    2013-08-15

    The 1st Technical Report was prepared for the research project ''Validation and Interpretation of the ATHLET-CD Model Basis'' funded by the Federal Ministry of Economics and Technology (BMWi1501385) and carried out at the Chair of Energy Systems and Energy Economics at Ruhr-Universitaet Bochum (RUB). This report provides results of the simulation of the Boiling Water Reactor (BWR) experiments CORA-17 and -28 with ATHLET-CD Mod. 2.2A. The system code ATHLET-CD (Analysis of Thermal-hydraulics of Leaks and Transients - Core Degradation) is developed by the German Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH. Code results are compared to measurements in order to assess and to analyze the capabilities of the current code version with regard to the modeling of BWR components. The CORA test series was carried out between the years 1987 and 1993 at the former Kernforschungszentrum Karlsruhe (KfK), now Karlsruhe Institute of Technology (KIT). The investigations provided experimental data regarding the material behavior during the early phase of core degradation in Light Water Reactors (LWR). The tests CORA17 and -28 represented a typical BWR arrangement of the fuel rod bundle and provided insights about the bundle behavior during the quenching process (CORA-17) and regarding the influence of a preoxidized bundle (CORA-28), respectively. The simulation results are analyzed and discussed in terms of the thermal bundle behavior, the zirconium oxidation in steam and the resulting hydrogen generation as well as the material relocation. In particular, the recently extended modeling capabilities of the code in terms of the relocation of BWR components like the absorber blade and the canister wall are assessed. The analysis shows that the code captures the thermal behavior in good agreement in both experiments. An even enhanced reproduction of the test CORA-28 is obtained in comparison to a calculation using the previous code version ATHLET-CD Mod

  13. Pattern recognition model to estimate intergranular stress corrosion cracking (IGSCC) at crevices and pit sites of 304 SS in BWR environments

    Energy Technology Data Exchange (ETDEWEB)

    Urquidi-Macdonald, Mirna [Penn State University, 212 Earth-Engineering Science Building, University Park, PA 16801 (United States)

    2004-07-01

    Many publications have shown that crack growth rates (CGR) due to intergranular stress corrosion cracking (IGSCC) of metals is dependent on many parameters related to the manufacturing process of the steel and the environment to which the steel is exposed. Those parameters include, but are not restricted to, the concentration of chloride, fluoride, nitrates, and sulfates, pH, fluid velocity, electrochemical potential (ECP), electrolyte conductivity, stress and sensitization applied to the steel during its production and use. It is not well established how combinations of each of these parameters impact the CGR. Many different models and beliefs have been published, resulting in predictions that sometimes disagree with experimental observations. To some extent, the models are the closest to the nature of IGSCC, however, there is not a model that fully describes the entire range of observations, due to the difficulty of the problem. Among the models, the Fracture Environment Model, developed by Macdonald et al., is the most physico-chemical model, accounting for experimental observations in a wide range of environments or ECPs. In this work, we collected experimental data on BWR environments and designed a data mining pattern recognition model to learn from that data. The model was used to generate CGR estimations as a function of ECP on a BWR environment. The results of the predictive model were compared to the Fracture Environment Model predictions. The results from those two models are very close to the experimental observations of the area corresponding to creep and IGSCC controlled by diffusion. At more negative ECPs than the potential corresponding to creep, the pattern recognition predicts an increase of CGR with decreasing ECP, while the Fracture Environment Model predicts the opposite. The results of this comparison confirm that the pattern recognition model covers 3 phenomena: hydrogen embrittlement at very negative ECP, creep at intermediate ECP, and IGSCC

  14. Application of the FFTBM method and the power relative contribution to the discharge transitory of the recirculation pumps of a BWR; Aplicacion del metodo FFTBM y de la contribucion relativa de potencia al transitorio de disparo de las bombas de recirculacion de un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Castillo D, R.; Ortiz V, J.; Fuentes M, L., E-mail: rogelio.castillo@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    In this work was realized the simulation of the discharge transitory of both recirculation pumps of a BWR with the Simulate-3K code. This type of transitory is used in the stability analyses for the licensing of the fuel reload. An analysis of the precision of the simulation is also presented, using the FFTBM method jointly with the power relative contribution. This way, instead of determining the total precision of the calculation, a weighed precision is obtained by the contribution of each relevant parameter of the transitory. The results show that the precision of the simulation is acceptable due to the small magnitude of the merit figure of the width total average. The error in the merit figure comes mainly from the parameters total flow in the core and temperature of the fuel in the core. (Author)

  15. Digital implementation, simulation and tests in MATLAB of the models of Steam line, the turbines, the pressure regulator of a BWR type nucleo electric power plant; Implementacion digital, simulacion y pruebas en MATLAB de los modelos de la linea de vapor, las turbinas y el regulador de presion de una central Nucleoelectrica tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Lopez R, A. [UNAM, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: andyskamx@yahoo.com.mx

    2004-07-01

    In this phase of the project they were carried out exhaustive tests to the models of the steam lines, turbines and pressure regulator of a BWR type nucleo electric central for to verify that their tendencies and behaviors are it more real possible. For it, it was necessary to also analyze the transfer functions of the different components along the steam line until the power generator. Such models define alone the dominant poles of the system, what is not limitation to reproduce a wide range of anticipated transitoriness of a power station operation. In the same manner, it was integrated and proved the integrated model form with the models of feeding water of the SUN-RAH, simulating the nuclear reactor starting from predetermined entrances of the prospective values of the vessel. Also it was coupled with the graphic interface developed with the libraries DirectX implementing a specific monitoring panel for this system. (Author)

  16. Strengthening the nuclear-reactor fuel cycle against proliferation

    Energy Technology Data Exchange (ETDEWEB)

    Travelli, A.; Snelgrove, J.; Persiani, P. [Argonne National Lab., IL (United States). Arms Control and Nonproliferation Program

    1992-12-31

    Argonne National Laboratory (ANL) conducts several research programs that serve to reduce the risks of fissile-material diversion from the nuclear-reactor fuel cycle. The objectives are to provide economical and efficient neutron or power generation with the minimum of inherent risks, and to further minimize risks by utilizing sophisticated techniques to detect attempts at material diversion. This paper will discuss the Reduced Enrichment Research and Test Reactor (RERTR) Program, the Isotope Correlation Technique (ICT), and Proliferation-Resistant Closed-Cycle Reactors. The first two are sponsored by the DOE Office of Arms Control and Nonproliferation.

  17. Study of intermediate configurations during the fuel reload in BWRs; Estudio de configuraciones intermedias durante la recarga de combustible en BWR's

    Energy Technology Data Exchange (ETDEWEB)

    Fuentes M, L.; Castillo M, J. A.; Ortiz S, J. J.; Perusquia del C, R. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Jacinto C, S., E-mail: luis.fuentes@inin.gob.mx [Universidad Autonoma del Estado de Yucatan, Calle 60 No. 491-A por 57, 97000 Merida, Yucatan (Mexico)

    2012-10-15

    The criticality state of the core of a boiling water reactor (BWR) was evaluated, during the reload process for the intermediate states between the load pattern of cycle end and the beginning of the next, using the information of the load pattern of the operation cycles 13 and 14 of Unit 1 of the nuclear power plant of Laguna Verde. For this evaluation the codes CASMO-4 and Simulate-3 for conditions of the core in cold were used. The strategy consisted on moving assemblies with 4 burned cycles of the reactor core. Later on were re situated the remaining assemblies, placing them in the positions to occupy in the next operation cycle. Finally, was carried out the assemblies load of fresh fuel. In each realized change, it was observing the behavior of the k-effective value that is the parameter used to evaluate the criticality state of each state of the core change. In a second stage, was designed a program that builds in automatic way each one of the intermediate cores and also analyzes the criticality state of the reactor core after each withdrawal, re situated and load of fuel assemblies. (Author)

  18. Possibilities with OHWC. Development and application of ECP-simulation in Swedish BWRs; Moejligheter med OHWC. Utveckling och tillaempning av ECP-simulering i svenska BWR

    Energy Technology Data Exchange (ETDEWEB)

    Lundgren, K. [ALARA Engineering, Skultuna (Sweden); Wikmark, G. [Advanced Nuclear Technology, Uppsala (Sweden)

    2000-02-01

    Hydrogen injection (HWC) to boiling water reactors has been used for two decades in Sweden, in order to reduce the impact of pipe cracking. The effect of HWC is to establish a sufficiently reducing environment in the systems to protect and hence mitigate the growth of existing stress corrosion cracks. Some disadvantages of HWC have been identified. One is the transitional increase of the dose rate of the main steam lines by up to seven times, another the corrosion release of systems with carbon steel components as a result of the reducing chemistry. In some cases, especially in the USA, an elevated activity build-up has been observed in a few plants in connection to the application of HWC. There is also a fear for increased hydrogen pick-up in fuel cladding and fuel channels by HWC operation. The hydrogen pick-up is already today in many cases limiting for fuel life. The objective of the current work has been to investigate the conditions by application of so called Optimised HWC. This implies a HWC operation with lower hydrogen addition rates than normally used. For this purpose, a computer model in order to simulate the radiolysis chemistry and the ECP (electrochemical corrosion potentials) in BWR systems has been developed. A previously developed radiolysis code, BwrChem, as well as a hydrogen peroxide decomposition code for piping, PEROX, have hence been equipped with ECP calculation modules. The ECP calculation algorithms have been based on fundamental electrochemical theory. The new model has been applied to simulate the radiolysis conditions in a large number of locations in typical BWRs. For the simulation, the external mechanical pump plant Barsebaeck-1 and the internal pump plant Forsmark-1 have been used. A wide range of hydrogen injection rates, down to 0. 1 ppm in the feed water, have been studied. The electrochemical model based on fundamental theory required adequate fundamental parameters. Significant effort has been used to scrutinise and evaluate

  19. Correlating activity incorporation with properties of oxide films formed on material samples exposed to BWR and PWR coolants in Finnish nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Bojinov, M.; Kinnunen, P.; Laitinen, T.; Maekelae, K.; Saario, T.; Sirkiae, P. [VTT Industrial Systems, Espoo (Finland); Buddas, T.; Halin, M.; Kvarnstroem, R.; Tompuri, K. [Fortum Power and Heat Oy, Loviisa Power Plant, Loviisa (Finland); Helin, M.; Muttilainen, E.; Reinvall, A. [Teollisuuden Voima Oy, Olkiluoto (Finland)

    2002-07-01

    The extent of activity incorporation on primary circuit surfaces in nuclear power plants is connected to the chemical composition of the coolant, to the corrosion behaviour of the material surfaces and to the structure and properties of oxide films formed on circuit surfaces due to corrosion. Possible changes in operational conditions may induce changes in the structure of the oxide films and thus in the rate of activity incorporation. To predict these changes, experimental correlations between water chemistry, oxide films and activity incorporation, as well as mechanistic understanding of the related phenomena need to be established. In order to do this, flow-through cells with material samples and facilities for high-temperature water chemistry monitoring have been installed at Olkiluoto unit 1 (BWR) and Loviisa unit 1 (PWR) in spring 2000. The cells are being used for two major purposes: To observe the changes in the structure and activity levels of oxide films formed on material samples exposed to the primary coolant. Correlating these observations with the abundant chemical and radiochemical data on coolant composition, dose rates etc. collected routinely by the plant, as well as with high-temperature water chemistry monitoring data such as the corrosion potentials of relevant material samples, the redox potential and the high-temperature conductivity of the primary coolant. We describe in this paper the scope of the work, give examples of the observations made and summarize the results on oxide films that have been obtained during one full fuel cycle at both plants. (authors)

  20. A novel auto-correlation function method and FORTRAN codes for the determination of the decay ratio in BWR stability analysis

    Energy Technology Data Exchange (ETDEWEB)

    Behringer, K

    2001-08-01

    A novel auto-correlation function (ACF) method has been investigated for determining the oscillation frequency and the decay ratio in BWR stability analyses. The report describes not only the method but also documents comprehensively the used and developed FORTRAN codes. The neutron signals are band-pass filtered to separate the oscillation peak in the power spectral density (PSD) from background. Two linear second-order oscillation models are considered. The ACF of each model, corrected for signal filtering and with the inclusion of a background term under the peak in the PSD, is then least-squares fitted to the ACF estimated on the previously filtered neutron signals, in order to determine the oscillation frequency and the decay ratio. The procedures of filtering and ACF estimation use fast Fourier transform techniques with signal segmentation. Gliding 'short-time' ACF estimates along a signal record allow the evaluation of uncertainties. Some numerical results are given which have been obtained from neutron signal data offered by the recent Forsmark I and Forsmark II NEA benchmark project. They are compared with those from other benchmark participants using different other analysis methods. (author)

  1. Validation of the Monte Carlo model developed to estimate doses around the irradiated fuel pool produced by activated control rods discharged from a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Abarca, Agustin; Gallardo, Sergio; Rodenas, Jose [Universidad Politecnica de Valencia (Spain). Dept. de Ingenieria Quimica y Nuclear], e-mail: ergalbe@iqn.upv.es

    2009-07-01

    BWR control rods are irradiated into the reactor by the neutron flux and consequently materials composing the rod become activated. When the control rods are withdrawn from the reactor, they must be stored into the spent fuel storage pool of the plant at a certain depth under water. Doses potentially received by plant workers in the area surrounding the pool edges as well as in a platform moving over the water surface should be calculated to assure the adequate protection. Irradiated fuel elements are stored at the bottom of the pool while controls rods are nearer the surface. Therefore, doses out of the pool are mainly produced by activated control rods. The MCNP5 code based on the Monte Carlo method has been applied to model the pool containing hanger devices with irradiated control rods and to estimate dose rates at points of interest. To obtain dose rates on the pool and around it an F4MESH tally has been used. Furthermore, the SSW/SSR technique has been applied in order to strongly reduce the computer time. Results have been compared with measurements in plant in order to validate the model. An interesting application of the validated model is the assessment of doses when some variation is introduced in the distribution of activated material. (author)

  2. 1D and 3D analyses of the Zy2 SCIP BWR ramp tests with the fuel codes METEOR and ALCYONE

    Energy Technology Data Exchange (ETDEWEB)

    Sercombe, J.; Agard, M.; Stuzik, C.; Michel, B.; Thouvenin, G.; Poussard, C.; Kallstrom, K. R. [CEA, Saint-Paul-lez Durance (France)

    2008-10-15

    In this paper, three power ramp tests performed on high burn-up Recrystallized Zircaloy2 - UO{sub 2} BWR fuel rods (56 to 63 MWd/kgU) within the SCIP project have been simulated with METEOR and ALCYONE 3D. Two of the ramp tests are of staircase type up to Linear Heat Rates of 420 and 520 W/cm and with long holding periods. Failure of the 420 W/cm fuel rod was observed after 40 minutes. The third ramp test consists in a more standard ramp test with a constant power rate of 80 W/cm/min till 410 W/cm and a short holding time. The tests were first simulated with the METEOR 1D fuel rod code leading accurate results in terms of profilometry and fission gas releases. The behaviour of a fuel pellet fragment and of the cladding piece on top of it was then investigated with ALCYONE 3D. The size and the main characteristics of the ridges after base irradiation and power ramp testing were recovered. Finally, the failure criteria validated for PWR conditions and fuel rods with low to medium burn ups were used to analyze the failure probability of the KKL rodlets during ramp testing.

  3. BWR spent fuel storage cask performance test. Volume 2. Pre- and post-test decay heat, heat transfer, and shielding analyses

    Energy Technology Data Exchange (ETDEWEB)

    Wiles, L.E.; Lombardo, N.J.; Heeb, C.M.; Jenquin, U.P.; Michener, T.E.; Wheeler, C.L.; Creer, J.M.; McCann, R.A.

    1986-06-01

    This report describes the decay heat, heat transfer, and shielding analyses conducted in support of performance testing of a Ridhihalgh, Eggers and Associates REA 2033 boiling water reactor (BWR) spent fuel storage cask. The cask testing program was conducted for the US Department of Energy (DOE) Commercial Spent Fuel Management Program by the Pacific Northwest Laboratory (PNL) and by General Electric at the latters' Morris Operation (GE-MO) as reported in Volume I. The analyses effort consisted of performing pretest calculations to (1) select spent fuel for the test; (2) symmetrically load the spent fuel assemblies in the cask to ensure lateral symmetry of decay heat generation rates; (3) optimally locate temperature and dose rate instrumentation in the cask and spent fuel assemblies; and (4) evaluate the ORIGEN2 (decay heat), HYDRA and COBRA-SFS (heat transfer), and QAD and DOT (shielding) computer codes. The emphasis of this second volume is on the comparison of code predictions to experimental test data in support of the code evaluation process. Code evaluations were accomplished by comparing pretest (actually pre-look, since some predictions were not completed until testing was in progress) predictions with experimental cask testing data reported in Volume I. No attempt was made in this study to compare the two heat transfer codes because results of other evaluations have not been completed, and a comparison based on one data set may lead to erroneous conclusions.

  4. Myostatin inhibits proliferation and insulin-stimulated glucose uptake in mouse liver cells.

    Science.gov (United States)

    Watts, Rani; Ghozlan, Mostafa; Hughey, Curtis C; Johnsen, Virginia L; Shearer, Jane; Hittel, Dustin S

    2014-06-01

    Although myostatin functions primarily as a negative regulator of skeletal muscle growth and development, accumulating biological and epidemiological evidence indicates an important contributing role in liver disease. In this study, we demonstrate that myostatin suppresses the proliferation of mouse Hepa-1c1c7 murine-derived liver cells (50%; p myostatin-responsive transcript in skeletal muscle, revealed a significant downregulation (25% and 50%, respectively; p myostatin-treated mice and liver cells. The importance of Malat1 in liver cell proliferation was confirmed via arrested liver cell proliferation (p Myostatin also significantly blunted insulin-stimulated glucose uptake and Akt phosphorylation in liver cells while increasing the phosphorylation of myristoylated alanine-rich C-kinase substrate (MARCKS), a protein that is essential for cancer cell proliferation and insulin-stimulated glucose transport. Together, these findings reveal a plausible mechanism by which circulating myostatin contributes to the diminished regenerative capacity of the liver and diseases characterized by liver insulin resistance.

  5. Identification of transcriptional networks involved in peroxisome proliferator chemical-induced hepatocyte proliferation

    Science.gov (United States)

    Peroxisome proliferator chemical (PPC) exposure leads to increases in rodent liver tumors through a non-genotoxic mode of action (MOA). The PPC MOA includes increased oxidative stress, hepatocyte proliferation and decreased apoptosis. We investigated the putative genetic regulato...

  6. Proliferation after the Iraq war; La proliferation apres la guerre d'Irak

    Energy Technology Data Exchange (ETDEWEB)

    Daguzan, J.F

    2004-09-15

    This article uses the Iraq war major event to analyze the approach used by the US to fight against proliferation. It questions the decision and analysis process which has led to the US-British intervention and analyzes the consequences of the war on the proliferation of other countries and on the expected perspectives. Finally, the future of proliferation itself is questioned: do we have to fear more threat or is the virtuous circle of non-proliferation well started? (J.S.)

  7. Control of cell proliferation by Myc

    DEFF Research Database (Denmark)

    Bouchard, C; Staller, P; Eilers, M

    1998-01-01

    Myc proteins are key regulators of mammalian cell proliferation. They are transcription factors that activate genes as part of a heterodimeric complex with the protein Max. This review summarizes recent progress in understanding how Myc stimulates cell proliferation and how this might contribute ...

  8. Teaching Activities on Horizontal Nuclear Proliferation.

    Science.gov (United States)

    Zola, John

    1990-01-01

    Provides learning activities concerning the horizontal proliferation of nuclear weapons. Includes step-by-step directions for four activities: (1) the life cycle of nuclear weapons; (2) nuclear nonproliferation: pros and cons; (3) the nuclear power/nuclear weapons connection; and (4) managing nuclear proliferation. (NL)

  9. Director`s series on proliferation

    Energy Technology Data Exchange (ETDEWEB)

    Bailey, K.C.; Price, M.E. [eds.

    1995-11-17

    This is an occasional publication of essays on the topics of nuclear, chemical, biological, and missile proliferation. The views represented are those of the author`s. Essay topics include: Nuclear Proliferation: Myth and Reality; Problems of Enforcing Compliance with Arms Control Agreements; The Unreliability of the Russian Officer Corps: Reluctant Domestic Warriors; and Russia`s Nuclear Legacy.

  10. SerpinB1 Promotes Pancreatic β Cell Proliferation

    Energy Technology Data Exchange (ETDEWEB)

    El Ouaamari, Abdelfattah; Dirice, Ercument; Gedeon, Nicholas; Hu, Jiang; Zhou, Jian-Ying; Shirakawa, Jun; Hou, Lifei; Goodman, Jessica; Karampelias, Christos; Qiang, Guifeng; Boucher, Jeremie; Martinez, Rachael; Gritsenko, Marina A.; De Jesus, Dario F.; Kahraman, Sevim; Bhatt, Shweta; Smith, Richard D.; Beer, Hans-Dietmar; Jungtrakoon, Prapaporn; Gong, Yanping; Goldfine, Allison B.; Liew, Chong Wee; Doria, Alessandro; Andersson, Olov; Qian, Wei-Jun; Remold-O’Donnell, Eileen; Kulkarni, Rohit N.

    2016-01-01

    Compensatory β-cell growth in response to insulin resistance is a common feature in diabetes. We recently reported that liver-derived factors participate in this compensatory response in the liver insulin receptor knockout (LIRKO) mouse, a model of significant islet hyperplasia. Here we show that serpinB1 is a liver-derived secretory protein that controls β-cell proliferation. SerpinB1 is abundant in the hepatocyte secretome and sera derived from LIRKO mice. SerpinB1 and small molecule compounds that partially mimic serpinB1 activity enhanced proliferation of zebrafish, mouse and human β-cells. We report that serpinB1-induced β-cell replication requires protease inhibition activity and mice lacking serpinB1 exhibit attenuated β-cell replication in response to insulin resistance. Finally, SerpinB1-treatment of islets modulated signaling proteins in growth and survival pathways such as MAPK, PKA and GSK3. Together, these data implicate SerpinB1 as a protein that can potentially be harnessed to enhance functional β-cell mass in patients with diabetes.

  11. Resistance-resistant antibiotics.

    Science.gov (United States)

    Oldfield, Eric; Feng, Xinxin

    2014-12-01

    New antibiotics are needed because drug resistance is increasing while the introduction of new antibiotics is decreasing. We discuss here six possible approaches to develop 'resistance-resistant' antibiotics. First, multitarget inhibitors in which a single compound inhibits more than one target may be easier to develop than conventional combination therapies with two new drugs. Second, inhibiting multiple targets in the same metabolic pathway is expected to be an effective strategy owing to synergy. Third, discovering multiple-target inhibitors should be possible by using sequential virtual screening. Fourth, repurposing existing drugs can lead to combinations of multitarget therapeutics. Fifth, targets need not be proteins. Sixth, inhibiting virulence factor formation and boosting innate immunity may also lead to decreased susceptibility to resistance. Although it is not possible to eliminate resistance, the approaches reviewed here offer several possibilities for reducing the effects of mutations and, in some cases, suggest that sensitivity to existing antibiotics may be restored in otherwise drug-resistant organisms. Copyright © 2014 Elsevier Ltd. All rights reserved.

  12. RELAP4/MOD5: a computer program for transient thermal-hydraulic analysis of nuclear reactors and related systems. User's manual. Volume I. RELAP4/MOD5 description. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    1976-09-01

    RELAP4 is a computer program written in FORTRAN IV for the digital computer analysis of nuclear reactors and related systems. It is primarily applied in the study of system transient response to postulated perturbations such as coolant loop rupture, circulation pump failure, power excursions, etc. The program was written to be used for water-cooled (PWR and BWR) reactors and can be used for scale models such as LOFT and SEMISCALE. Additional versatility extends its usefulness to related applications, such as ice condenser and containment subcompartment analysis. Specific options are available for reflood (FLOOD) analysis and for the NRC Evaluation Model.

  13. Use of Molecular Imaging Markers of Glycolysis, Hypoxia and Proliferation (18F-FDG, 64Cu-ATSM and 18F-FLT) in a Dog with Fibrosarcoma

    DEFF Research Database (Denmark)

    Zornhagen, Kamilla; Clausen, Malene; Hansen, Anders Elias

    2015-01-01

    Glycolysis, hypoxia, and proliferation are important factors in the tumor microenvironment contributing to treatment-resistant aggressiveness. Imaging these factors using combined functional positron emission tomography and computed tomography can potentially guide diagnosis and management...

  14. Harmine stimulates proliferation of human neural progenitors

    Directory of Open Access Journals (Sweden)

    Vanja Dakic

    2016-12-01

    Full Text Available Harmine is the β-carboline alkaloid with the highest concentration in the psychotropic plant decoction Ayahuasca. In rodents, classical antidepressants reverse the symptoms of depression by stimulating neuronal proliferation. It has been shown that Ayahuasca presents antidepressant effects in patients with depressive disorder. In the present study, we investigated the effects of harmine in cell cultures containing human neural progenitor cells (hNPCs, 97% nestin-positive derived from pluripotent stem cells. After 4 days of treatment, the pool of proliferating hNPCs increased by 71.5%. Harmine has been reported as a potent inhibitor of the dual specificity tyrosine-phosphorylation-regulated kinase (DYRK1A, which regulates cell proliferation and brain development. We tested the effect of analogs of harmine, an inhibitor of DYRK1A (INDY, and an irreversible selective inhibitor of monoamine oxidase (MAO but not DYRK1A (pargyline. INDY but not pargyline induced proliferation of hNPCs similarly to harmine, suggesting that inhibition of DYRK1A is a possible mechanism to explain harmine effects upon the proliferation of hNPCs. Our findings show that harmine enhances proliferation of hNPCs and suggest that inhibition of DYRK1A may explain its effects upon proliferation in vitro and antidepressant effects in vivo.

  15. Alternative cooling water flow path for RHR heat exchanger and its effect on containment response during extended station blackout for Chinshan BWR-4 plant

    Energy Technology Data Exchange (ETDEWEB)

    Yuann, Yng-Ruey, E-mail: ryyuann@iner.gov.tw

    2016-04-15

    Highlights: • Motivating alternative RHR heat exchanger tube-side flow path and determining required capacity. • Calculate NSSS and containment response during 24-h SBO for Chinshan BWR-4 plant. • RETRAN and GOTHIC models are developed for NSSS and containment, respectively. • Safety relief valve blowdown flow and energy to drywell are generated by RETRAN. • Analyses are performed with and without reactor depressurization, respectively. - Abstract: The extended Station Blackout (SBO) of 24 h has been analyzed with respect to the containment response, in particular the suppression pool temperature response, for the Chinshan BWR-4 plant of MARK-I containment. The Chinshan plant, owned by Taiwan Power Company, has twin units with rated core thermal power of 1840 MW each. The analysis is aimed at determining the required alternative cooling water flow capacity for the residual heat removal (RHR) heat exchanger when its tube-side sea water cooling flow path is blocked, due to some reason such as earthquake or tsunami, and is switched to the alternative raw water source. Energy will be dissipated to the suppression pool through safety relief valves (SRVs) of the main steam lines during SBO. The RETRAN model is used to calculate the Nuclear Steam Supply System (NSSS) response and generate the SRV blowdown conditions, including SRV pressure, enthalpy, and mass flow rate. These conditions are then used as the time-dependent boundary conditions for the GOTHIC code to calculate the containment pressure and temperature response. The shaft seals of the two recirculation pumps are conservatively assumed to fail due to loss of seal cooling and a total leakage flow rate of 36 gpm to the drywell is included in the GOTHIC model. Based on the given SRV blowdown conditions, the GOTHIC containment calculation is performed several times, through the adjustment of the heat transfer rate of the RHR heat exchanger, until the criterion that the maximum suppression pool temperature

  16. Uncertainty analysis of suppression pool heating during an ATWS in a BWR-5 plant. An application of the CSAU methodology using the BNL engineering plant analyzer

    Energy Technology Data Exchange (ETDEWEB)

    Wulff, W.; Cheng, H.S.; Mallen, A.N. [Brookhaven National Lab., Upton, NY (United States); Johnsen, G.W. [Idaho National Engineering Lab., Idaho Falls, ID (United States); Lellouche, G.S. [Technical Data Services, Chicago, IL (United States)

    1994-03-01

    The uncertainty has been estimated of predicting the peak temperature in the suppression pool of a BWR power plant, which undergoes an NRC-postulated Anticipated Transient Without Scram (ATWS). The ATWS is initiated by recirculation-pump trips, and then leads to power and flow oscillations as they had occurred at the LaSalle-2 Power Station in March of 1988. After limit-cycle oscillations have been established, the turbines are tripped, but without MSIV closure, allowing steam discharge through the turbine bypass into the condenser. Postulated operator actions, namely to lower the reactor vessel pressure and the level elevation in the downcomer, are simulated by a robot model which accounts for operator uncertainty. All balance of plant and control systems modeling uncertainties were part of the statistical uncertainty analysis that was patterned after the Code Scaling, Applicability and Uncertainty (CSAU) evaluation methodology. The analysis showed that the predicted suppression-pool peak temperature of 329.3 K (133{degrees}F) has a 95-percentile uncertainty of 14.4 K (26{degrees}F), and that the size of this uncertainty bracket is dominated by the experimental uncertainty of measuring Safety and Relief Valve mass flow rates under critical-flow conditions. The analysis showed also that the probability of exceeding the suppression-pool temperature limit of 352.6 K (175{degrees}F) is most likely zero (it is estimated as < 5-104). The square root of the sum of the squares of all the computed peak pool temperatures is 350.7 K (171.6{degrees}F).

  17. Effect of nonlinear void reactivity on bifurcation characteristics of a lumped-parameter model of a BWR: A study relevant to RBMK

    Energy Technology Data Exchange (ETDEWEB)

    Verma, Dinkar, E-mail: dinkar@iitk.ac.in [Nuclear Engineering and Technology Program, Indian Institute of Technology Kanpur, Kanpur 208 016 (India); Kalra, Manjeet Singh, E-mail: drmanjeet.singh@dituniversity.edu.in [DIT University, Dehradun 248 009 (India); Wahi, Pankaj, E-mail: wahi@iitk.ac.in [Department of Mechanical Engineering, Indian Institute of Technology Kanpur, Kanpur 208 016 (India)

    2017-04-15

    Highlights: • A simplified model with nonlinear void reactivity feedback is studied. • Method of multiple scales for nonlinear analysis and oscillation characteristics. • Second order void reactivity dominates in determining system dynamics. • Opposing signs of linear and quadratic void reactivity enhances global safety. - Abstract: In the present work, the effect of nonlinear void reactivity on the dynamics of a simplified lumped-parameter model for a boiling water reactor (BWR) is investigated. A mathematical model of five differential equations comprising of neutronics and thermal-hydraulics encompassing the nonlinearities associated with both the reactivity feedbacks and the heat transfer process has been used. To this end, we have considered parameters relevant to RBMK for which the void reactivity is known to be nonlinear. A nonlinear analysis of the model exploiting the method of multiple time scales (MMTS) predicts the occurrence of the two types of Hopf bifurcation, namely subcritical and supercritical, leading to the evolution of limit cycles for a range of parameters. Numerical simulations have been performed to verify the analytical results obtained by MMTS. The study shows that the nonlinear reactivity has a significant influence on the system dynamics. A parametric study with varying nominal reactor power and operating conditions in coolant channel has also been performed which shows the effect of change in concerned parameter on the boundary between regions of sub- and super-critical Hopf bifurcations in the space constituted by the two coefficients of reactivities viz. the void and the Doppler coefficient of reactivities. In particular, we find that introduction of a negative quadratic term in the void reactivity feedback significantly increases the supercritical region and dominates in determining the system dynamics.

  18. Thermodynamic consideration of hydrogen injection in BWR coolant. Estimation of potential for SCC control and oxidation-reduction condition of reactor coolant

    Energy Technology Data Exchange (ETDEWEB)

    Miyajima, Kaori; Hirano, Hideo; Domae, Masashi [Central Research Inst. of Electric Power Industry, Komae, Tokyo (Japan). Komae Research Lab; Kushida, H.

    2001-04-01

    Hydrogen injection into BWR coolant has been carried out in order to reduce stress corrosion cracking (SCC). It was clarified by in-plant test that SCC can be reduced under corrosion potential -0.23 V(v.s.SHE), but the theoretical basis has not been clarified. On the other hand, highly precise water quality analysis of re-circulatory-system water is generally performed. Especially, nitrogen compound changes chemical from to NO{sub 3}{sup -} -> NO{sub 2}{sup -} -> NH{sub 3}, and the NH{sub 3} becomes the cause of the increase of dose rate of the main steamy system in connection with the increase in the amount of hydrogen injection. However, the relation between this chemical form, oxidisation reduction potential, and temperature is not clear: Then, in this paper, these two points were considered by thermodynamics calculation at 25-300degC using the thermodynamics data in the high temperature accumulated in CRIEPI, and calculation results are summarized as follows; (1) the potential of the stainless steel to which the chemical form change to FeCr{sub 2}O{sub 4} from NiFe{sub 2}O{sub 4} is equilibrium is about -0.23 V at 288degC so this change is expected as one of factors for reduction of SCC, (2) the changes of chemical form of nitrogen compounds show oxidation-reduction of reactor coolant, so it can be useful as the index for control of dose rate. (author)

  19. Comparison study of the thermal mechanical performance of fuel rods during BWR fuel preconditioning operations using the computer codes FUELSIM and FEMAXI-V

    Energy Technology Data Exchange (ETDEWEB)

    Pantoja C, R. [IPN, Escuela Superior de Fisica y Matematicas, Departamento de Ingenieria Nuclear, Av. Instituto Politecnico Nacional s/n, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Ortiz V, J.; Castillo D, R., E-mail: rafael.pantoja10@yahoo.com.m [ININ, Departamento de Sistemas Nucleares, Carretera Mexico-Toluca s/n, Ocoyoacac 52750, Estado de Mexico (Mexico)

    2010-10-15

    The safety of nuclear power plants requires monitoring those parameters having some direct or indirect effect on safety. The thermal limits are values set for those parameters considered having most impact on the safe operation of a nuclear power reactor. Some thermal limits monitoring requires the thermal-mechanical analysis of the rods containing the nuclear fuel. The fuel rod thermal-mechanical behaviour under irradiation is a complex process in which there exists a great deal of interrelated physical and chemical phenomena, so that the fuel rod performance analysis in the core of a nuclear power reactor is generally accomplished by using computer codes, which integrate several of the phenomena that are expected to occur during the lifetime of the fuel rod in the core. In the operation of a nuclear power reactor, pre-conditioning simulations are necessary to determine in advance limit values for the power that can be generated in a fuel rod during any power ramp, and mainly during reactor startup, and thus avoiding any rod damage. In this work, a first analysis of the thermal-mechanical performance of typical fuel rods used in nuclear reactors of the type BWR is performed. This study includes two types of fuel rods: one from a fuel assembly design with array 8 x 8, and the other one from a 10 x 10 fuel assembly design, and a comparison of the thermal-mechanical performance between the two different rod designs is performed. The performance simulations were performed by the code FUELSIM, and compared against results previously obtained from similar simulation with the code FEMAXI-V. (Author)

  20. Inspection findings in austenitic RPV internals of German BWR plants and BWRs built in other countries and resulting measures for Isar 1 nuclear power station

    Energy Technology Data Exchange (ETDEWEB)

    Erve, M.; Hurlebaus, D.; Marschke, D.; Senski, G. [Siemens AG, Erlangen (Germany). Power Generation Group; Maier, V. [Bayernwerk Kernenergie GmbH, Muenchen (Germany); Baeumler, H.-J.; Winter, F. [TUeV Anlagen-und Umwetttechnik GmbH, Muenchen (Germany)

    1999-06-01

    As visual examinations carried out in autumn 1994 detected cracks in a German BWR plant due to intergranular stress corrosion cracking (IGSCC) in several core shroud components manufactured from 1.4550 steel, precautionary examinations and assessments were performed for all other plants. In accordance with these analyses, it can be stated for Isar 1 that the heat treatment to which the components in question were subjected in the course of manufacture cannot have caused sensitization of the material, and that crack formation due to the damage mechanism primarily identified in the reactor vessel internals at Wuergassen Nuclear Power Station need not be feared. Although the material and corrosion-chemical assessments performed to date did not give any indications for the other crack formation mechanisms that are theoretically relevant for reactor vessel internals (IGSCC due to weld sensitization, IASCC (irradiation assisted stress corrosion cracking)), visual examinations with a limited scope will be carried out with the independant expert`s agreement during the scheduled inservice inspections. The fluid-dynamic and structure-mechanical analyses showed that the individual components are subjected only to low loadings, even in the event of accidents, and that the safety objectives shutdown and residual heat removal can be fulfilled even in the case of large postulated cracks. The fracture-mechanics analyses indicated critical through-wall crack lengths which, however, can be promptly and reliably detected during random inservice inspections even when assuming stress corrosion cracking and irradiation-induced low-toughness material conditions. In addition, both the VGB and the Isar 1 plant are pursuing further prophylactic measures such as alternative water chemistry modes and an appropriate repair and replacement concept. (orig.) 18 refs.

  1. VAV3 mediates resistance to breast cancer endocrine therapy

    NARCIS (Netherlands)

    H. Aguilar (Helena); A. Urruticoechea (Ander); P. Halonen (Pasi); K. Kiyotani (Kazuma); T. Mushiroda (Taisei); X. Barril (Xavier); J. Serra-Musach (Jordi); A.B.M.M.K. Islam (Abul); L. Caizzi (Livia); L. Di Croce (Luciano); E. Nevedomskaya (Ekaterina); W. Zwart (Wilbert); J. Bostner (Josefine); E. Karlsson (Elin); G. Pérez Tenorio (Gizeh); T. Fornander (Tommy); D.C. Sgroi (Dennis); R. Garcia-Mata (Rafael); M.P.H.M. Jansen (Maurice); N. García (Nadia); N. Bonifaci (Núria); F. Climent (Fina); E. Soler (Eric); A. Rodríguez-Vida (Alejo); M. Gil (Miguel); J. Brunet (Joan); G. Martrat (Griselda); L. Gómez-Baldó (Laia); A.I. Extremera (Ana); J. Figueras; J. Balart (Josep); R. Clarke (Robert); K.L. Burnstein (Kerry); K.E. Carlson (Kathryn); J.A. Katzenellenbogen (John); M. Vizoso (Miguel); M. Esteller (Manel); A. Villanueva (Alberto); A.B. Rodríguez-Peña (Ana); X.R. Bustelo (Xosé); Y. Nakamura (Yusuke); H. Zembutsu (Hitoshi); O. Stål (Olle); R.L. Beijersbergen (Roderick); M.A. Pujana (Miguel)

    2014-01-01

    textabstractIntroduction: Endocrine therapies targeting cell proliferation and survival mediated by estrogen receptor α (ERα) are among the most effective systemic treatments for ERα-positive breast cancer. However, most tumors initially responsive to these therapies acquire resistance through

  2. Neutron activation analysis and activity in the vessel steel of a BWR reactor for their study without radiological risks in microscopy and spectrometry; Analisis de activacion neutronica y actividad en el acero de la vasija de un reactor nuclear tipo BWR para su estudio sin riesgos radiologicos en microscopia y espectrometria

    Energy Technology Data Exchange (ETDEWEB)

    Moranchel, M.; Garcia B, A. [IPN, Escuela Superior de Fisica y Matematicas, Departamento de Fisica, Unidad Profesional Adolfo Lopez Mateos, Zacatenco, 07738 Mexico D. F. (Mexico); Longoria G, L. C., E-mail: mmoranchel@ipn.mx [IAEA, Department of Technical Cooperation, Division for Latin America, Room B1109 Wagramerstrasse 5, PO Box 100, A-1400, Vienna (Austria)

    2012-07-01

    The vessel material of nuclear reactors is subject to irradiation damage induced by the bombardment of neutrons coming from the reactor core. Neutrons are classified as fast and thermal, which produce different effects. Fast neutrons cause damage to the material by dislocation or displacement of atoms in the crystal structure, while the effect of thermal neutrons is a nuclear transmutation that can significantly change the properties of the material. The type and intensity of damage is based on the characteristics of the material, the flow of neutrons and the modes of neutrons interaction with the atomic structures of the material, among others. This work, alluding to nuclear transmutation, makes an analysis of neutron activation of all isotopes in a steel boiling water nuclear reactor (BWR) vessel. An analytical expression is obtained in order to model activity of steel, on the basis of the weight percentage of its atomic components. Its activity is theoretically estimated in a witness sample of the same material as that of the vessel, placed within the nuclear reactor since the beginning of its commercial operation in April 1995, up to August 2010. It was theoretically determined that the witness sample, with a 0.56 g mass (1 x 1 x 0.07 cm{sup 3} dimensions or equivalent) does not present a radiological risks during the stage of preparation, observation and analysis of it in electron microscopy and X-ray diffraction equipment s. The theoretical results were checked experimentally by measuring the activity of the sample by means of gamma spectrometry, measurement of the exposure levels around the sample, as well as the induced level to whole body and limbs, using thermo-luminescent dosimetry (TLD). As a result of the theoretical analysis, new chemical elements are predicted, as a result of the activation phenomena and radioactive decay, whose presence can be a fundamental factor of change in the properties of the vessel. This work is a preamble to the

  3. Evaluation of the thermal-mechanical performance of fuel rods of a BWR during a power ramp using the FUELSIM code; Evaluacion del desempeno termomecanico de barras de combustible de un reactor BWR durante una rampa de potencia utilizando el codigo FUELSIM

    Energy Technology Data Exchange (ETDEWEB)

    Pantoja C, R.

    2010-07-01

    To avoid the risk to environment due to release of radioactive material, because of occurrence of an accident, it is the priority of the design and performance of the diverse systems of safety of a commercial nuclear power plant. The safety of nuclear power plants requires, therefore, monitoring those parameters having some direct or indirect effect on safety. The thermal limits are values set for those parameters considered having most impact on the safe operation of a nuclear power reactor. Some thermal limits monitoring requires the thermal-mechanical analysis of the rods containing the nuclear fuel. The fuel rod thermal-mechanical behavior under irradiation is a complex process in which there exists a great deal of interrelated physical and chemical phenomena, so that the fuel rod performance analysis in the core of a nuclear power reactor is generally accomplished by using computer codes, which integrate several of the phenomena that are expected to occur during the lifetime of the fuel rod in the core. The main application of the thermal-mechanical analysis codes is the prediction of occurrence of conditions and/or phenomena that could lead to the deterioration or even mechanical failure of the fuel rod cladding, as, for example, the pellet-cladding interaction. In the operation of a nuclear power reactor, fuel preconditioning operations refer to the operational procedures employed to reduce the fuel rod failure probability due to fuel-cladding interaction, specially during reactor startup. Preconditioning simulations are therefore necessary to determine in advance limit values for the power that can be generated in a fuel rod, and thus avoiding any rod damage. In this work, a first analysis of the thermal-mechanical performance of typical fuel rods used in nuclear reactors of the type BWR 5/6, as those two nuclear reactors in Laguna Verde, Veracruz, is performed. This study includes two types of fuel rods: one from a fuel assembly design with an array 8 x 8

  4. NATO's Response to the Proliferation Challenge

    National Research Council Canada - National Science Library

    Joseph, Robert

    1996-01-01

    ... operations in an NBC environment. The success of the NATO initiative to counter the proliferation threat, however, will only be assured when allies make national and collective commitments to field the necessary military capabilities...

  5. Handbook for nuclear non-proliferation

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byung Wook; Oh, Keun Bae; Lee, Kwang Seok; Lee, Dong Jin; Ko, Han Seok

    1997-05-01

    This book analyzed international non-proliferation regime preventing from spread of nuclear weapon. This book took review from the historical background of non-proliferation regime to the recent changes and status. The regime, here, is divided into multilateral and bilateral regime. First of all, this book reports four multilateral treaties concluded for non-proliferation such as NPT, NWFZ, CTBT and others. Secondly, international organization and regimes concerned with non-proliferation are analyzed with emphasis of UN, IAEA, ZC and NSG, Regional Safeguards System and international conference. Finally, this book report the current circumstances of nuclear cooperation agreement related with Korea which is an important means for bilateral regime. (author). 13 tabs., 2 figs.

  6. Institutional arrangements for the reduction of proliferation risks formulation, evaluation, and implementation of institutional concepts

    Energy Technology Data Exchange (ETDEWEB)

    Kratzer, M.B.; Wonder, E.F.; Gray, J.E.; Shantzis, S.B.; Sievering, N.F.; Paige, H.W.; Jones, B.M.

    1979-12-01

    The purpose of this study was to: (1) identify alternative institutional arrangements applicable to the sensitive steps in the back-end of the fuel cycle that might reduce their associated proliferation risks; and (2) assess their advantages and disadvantages from the standpoint of nonproliferation effectiveness and political, economic, and operational acceptability. The concept of international or multinational custody of sensitive materials and facilities was found to offer a high degree of proliferation resistance and to likely be more acceptable to prospective participants than other institutional arrangements that intrude upon proprietary areas, such as facility ownership and management.

  7. Determination of the neutron fluence in the welding of the 'Core shroud' of the BWR reactor core; Determinacion de la fluencia neutronica en las soldaduras del 'core shroud' del nucleo de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Lucatero, M.A.; Xolocostli M, J.V.; Gomez T, A.M.; Palacios H, J.C. [ININ, 52750 Ocoyoacac, Estado de mexico (Mexico)]. e-mail: mal@nuclear.inin.mx

    2006-07-01

    With the purpose of defining the inspection frequency, in function of the embrittlement of the materials that compose the welding of the 'Core Shroud' or encircling of the core of a BWR type reactor, is necessary to know the neutron fluence received for this welding. In the work the calculated values of neutron fluence accumulated maxim (E > 1 MeV) during the first 8 operation cycles of the reactor are presented. The calculations were carried out according to the NRC Regulatory Guide 1.190, making use of the DORT code, which solves the transport equation in discreet ordinate in two dimensions (xy, r{theta}, and rz). The results in 3D were obtained applying the Synthesis method according to the guide before mentioned. Results are presented for the horizontal welding H3, H4, and H5, showing the corresponding curves to the fluence accumulated to the cycle 8 and a projection for the cycle 14 is presented. (Author)

  8. Calculation of the linear heat generation rates which violate the thermomechanical limit of plastic deformation of the fuel cladding in function of the burn up of a BWR fuel rod type; Calculo de las razones de generacion de calor lineal que violen el limite termomecanico de deformacion plastica de la camisa en funcion del quemado de una barra combustible tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Lucatero, M.A.; Hernandez L, H. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: mal@nuclear.inin.mx

    2003-07-01

    The linear heat generation rates (LHGR) for a BWR type generic fuel rod, as function of the burnup that violate the thermomechanical limit of circumferential plastic deformation of the can (canning) in nominal operation in stationary state of the fuel rod are calculated. The evaluation of the LHGR in function of the burnt of the fuel, is carried out under the condition that the deformation values of the circumferential plastic deformation of the can exceeds in 0.1 the thermomechanical value operation limit of 1%. The results of the calculations are compared with the generation rates of linear operation heat in function of the burnt for this fuel rod type. The calculations are carried out with the FEMAXI-V and RODBURN codes. The results show that for exhibitions or burnt between 0 and 16,000 M Wd/tU a minimum margin of 160.8 W/cm exists among LHGR (439.6 W/cm) operation peak for the given fuel and maximum LHGR of the fuel (calculated) to reach 1.1% of circumferential plastic deformation of the can, for the peak factor of power of 1.40. For burnt of 20,000 MWd/tU and 60,000 MWd/tU exist a margin of 150.3 and 298.6 W/cm, respectively. (Author)

  9. Growth and stability of stress corrosion cracks in large-diameter BWR piping. Volume 1: summary. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Hale, D A; Heald, J D; Horn, R M; Jewett, C W; Kass, J N; Mehta, H S; Ranganath, S; Sharma, S R

    1982-07-01

    This report presents the results of a research program conducted to evaluate the behavior of hypothetical stress corrosion cracks in large diameter austenitic piping. The program included major tasks, a design margin assessment, an evaluation of crack growth and crack arrest, and development of a predictive model. As part of the margin assessment, the program developed diagrams which predicted net section collapse as a function of crack size. In addition, plasticity and dynamic load effects were also considered in evaluating collapse. Analytical methods for evaluating these effects were developed and were benchmarked by dynamic tests of 4-in.-diameter piping. The task of evaluating the growth behavior of stress corrosion cracks focused on developing constant load and cyclic growth rate data that could be used with the predictive model. Secondly, laboratory tests were performed to evaluate the conditions under which growing stress corrosion cracks would arrest when they intersected stress corrosion resistant weld metal. The third task successfully developed a model to predict the behavior of cracks in austenitic piping. This model relies on crack growth data and the critical crack size predicted by the net section collapse approach.

  10. Peroxisome Proliferator-Activated Receptors in Diabetic Nephropathy

    Directory of Open Access Journals (Sweden)

    Shinji Kume

    2008-01-01

    Full Text Available Diabetic nephropathy is a leading cause of end-stage renal disease, which is increasing in incidence worldwide, despite intensive treatment approaches such as glycemic and blood pressure control in patients with diabetes mellitus. New therapeutic strategies are needed to prevent the onset of diabetic nephropathy. Peroxisome proliferator-activated receptors (PPARs are ligand-activated nuclear transcription factors that play important roles in lipid and glucose homeostases. These agents might prevent the progression of diabetic nephropathy, since PPAR agonists improve dyslipidemia and insulin resistance. Furthermore, data from murine models suggest that PPAR agonists also have independent renoprotective effects by suppressing inflammation, oxidative stress, lipotoxicity, and activation of the renin-angiotensin system. This review summarizes data from clinical and experimental studies regarding the relationship between PPARs and diabetic nephropathy. The therapeutic potential of PPAR agonists in the treatment of diabetic nephropathy is also discussed.

  11. Director`s series on proliferation

    Energy Technology Data Exchange (ETDEWEB)

    Bailey, K.C. [ed.

    1993-09-07

    Two essays are included in this booklet. Their titles are ``The Dynamics of the NPT Extension Decision`` and ``North Korea`s Nuclear Gambit.`` The first paper discusses the conference to be held in 1995 to review the Nuclear Non-Proliferation Treaty (NPT) which will decide whether the treaty shall continue in force indefinitely, or shall be extended for an additional fixed period or periods. Topics relevant to this discussion are: Arms control issues, the nuclear test ban, the limited test ban treaty, the French nuclear testing moratorium, former Soviet nuclear weapons, Iraq, North Korea, nuclear-weapon-free zones, security, controls on nuclear weapon materials, peaceful uses of nuclear energy, safeguards, politics, and organizational and procedural issues. The second paper examines short, medium, and long term issues entailed in Korea`s nuclear proliferation. Topics considered include: Korean unification, North Korean politics, the nuclear issue as leverage, and the Nuclear Non- Proliferation Treaty.

  12. Leukocytic promotion of prostate cellular proliferation.

    Science.gov (United States)

    McDowell, Kristy L; Begley, Lesa A; Mor-Vaknin, Nirit; Markovitz, David M; Macoska, Jill A

    2010-03-01

    Histological evidence of pervasive inflammatory infiltrate has been noted in both benign prostatic hyperplasia/hypertrophy (BPH) and prostate cancer (PCa). Cytokines known to attract particular leukocyte subsets are secreted from prostatic stroma consequent to aging and also from malignant prostate epithelium. Therefore, we hypothesized that leukocytes associated with either acute or chronic inflammation attracted to the prostate consequent to aging or tumorigenesis may promote the abnormal cellular proliferation associated with BPH and PCa. An in vitro system designed to mimic the human prostatic microenvironment incorporating prostatic stroma (primary and immortalized prostate stromal fibroblasts), epithelium (N15C6, BPH-1, LNCaP, and PC3 cells), and inflammatory infiltrate (HL-60 cells, HH, and Molt-3 T-lymphocytes) was developed. Modified Boyden chamber assays were used to test the ability of prostate stromal and epithelial cells to attract leukocytes and to test the effect of leukocytes on prostate cellular proliferation. Antibody arrays were used to identify leukocyte-secreted cytokines mediating prostate cellular proliferation. Leukocytic cells migrated towards both prostate stromal and epithelial cells. CD4+ T-lymphocytes promoted the proliferation of both transformed and non-transformed prostate epithelial cell lines tested, whereas CD8+ T-lymphocytes as well as dHL-60M macrophagic and dHL-60N neutrophilic cells selectively promoted the proliferation of PCa cells. The results of these studies show that inflammatory cells can be attracted to the prostate tissue microenvironment and can selectively promote the proliferation of non-transformed or transformed prostate epithelial cells, and are consistent with differential role(s) for inflammatory infiltrate in the etiologies of benign and malignant proliferative disease in the prostate. Prostate 70: 377-389, 2010. (c) 2009 Wiley-Liss, Inc.

  13. Dedifferentiation and proliferation of mammalian cardiomyocytes.

    Directory of Open Access Journals (Sweden)

    Yiqiang Zhang

    2010-09-01

    Full Text Available It has long been thought that mammalian cardiomyocytes are terminally-differentiated and unable to proliferate. However, myocytes in more primitive animals such as zebrafish are able to dedifferentiate and proliferate to regenerate amputated cardiac muscle.Here we test the hypothesis that mature mammalian cardiomyocytes retain substantial cellular plasticity, including the ability to dedifferentiate, proliferate, and acquire progenitor cell phenotypes. Two complementary methods were used: 1 cardiomyocyte purification from rat hearts, and 2 genetic fate mapping in cardiac explants from bi-transgenic mice. Cardiomyocytes isolated from rodent hearts were purified by multiple centrifugation and Percoll gradient separation steps, and the purity verified by immunostaining and RT-PCR. Within days in culture, purified cardiomyocytes lost their characteristic electrophysiological properties and striations, flattened and began to divide, as confirmed by proliferation markers and BrdU incorporation. Many dedifferentiated cardiomyocytes went on to express the stem cell antigen c-kit, and the early cardiac transcription factors GATA4 and Nkx2.5. Underlying these changes, inhibitory cell cycle molecules were suppressed in myocyte-derived cells (MDCs, while microRNAs known to orchestrate proliferation and pluripotency increased dramatically. Some, but not all, MDCs self-organized into spheres and re-differentiated into myocytes and endothelial cells in vitro. Cell fate tracking of cardiomyocytes from 4-OH-Tamoxifen-treated double-transgenic MerCreMer/ZEG mouse hearts revealed that green fluorescent protein (GFP continues to be expressed in dedifferentiated cardiomyocytes, two-thirds of which were also c-kit(+.Contradicting the prevailing view that they are terminally-differentiated, postnatal mammalian cardiomyocytes are instead capable of substantial plasticity. Dedifferentiation of myocytes facilitates proliferation and confers a degree of stemness

  14. Director`s series on proliferation

    Energy Technology Data Exchange (ETDEWEB)

    Bailey, K.C.; Price, M.E. [eds.

    1994-12-27

    The Director`s Series on Proliferation is an occasional publication of essays on the topics of nuclear, chemical, biological, and missile proliferation. The seven papers presented in this issue cover the following topics: Should the Treaty on the Nonproliferation of Nuclear Weapons (NPT) be amended?; NPT extension - Legal and procedural issues; An Indonesian view of NPT review conference issues; The treaty of Tlatelolco and the NPT - Tools for peace and development; Perspectives on cut-off, weapons dismantlement, and security assurances; Belarus and NPT challenges; A perspective on the chemical weapons convention - Lessons learned from the preparatory commission.

  15. Fatty acids and breast cancer cell proliferation.

    Science.gov (United States)

    Hardy, R W; Wickramasinghe, N S; Ke, S C; Wells, A

    1997-01-01

    We and others have shown that fatty acids are important regulators of breast cancer cell proliferation. In particular individual fatty acids specifically alter EGF-induced cell proliferation in very different ways. This regulation is mediated by an EGFR/G-protein signaling pathway. Understanding the molecular mechanisms of how this signaling pathway functions and how fatty acids regulate it will provide important information on the cellular and molecular basis for the association of dietary fat and cancer. Furthermore these in vitro studies may explain data previously obtained from in vivo animal studies and identify "good" as well as "bad" fatty acids with respect to the development of cancer.

  16. Proliferation risks; Proliferatierisico's

    Energy Technology Data Exchange (ETDEWEB)

    Carchon, R

    1998-09-01

    The report gives an overview of different aspects related to safeguards of fissile materials. Existing treaties including the Non-Proliferation Treaty, and the Tlatelolco and the Rarotonga Treaties are discussed. An overview of safeguards systems for the control of fissile materials as well as the role of various authorities is given. An overall overview of proliferation risks, the physical protection of fissile materials and the trade in fissile materials is given. Finally, the status in problem countries and de facto nuclear weapon states is discussed.

  17. Supervised Semantic Classification for Nuclear Proliferation Monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Vatsavai, Raju [ORNL; Cheriyadat, Anil M [ORNL; Gleason, Shaun Scott [ORNL

    2010-01-01

    Existing feature extraction and classification approaches are not suitable for monitoring proliferation activity using high-resolution multi-temporal remote sensing imagery. In this paper we present a supervised semantic labeling framework based on the Latent Dirichlet Allocation method. This framework is used to analyze over 120 images collected under different spatial and temporal settings over the globe representing three major semantic categories: airports, nuclear, and coal power plants. Initial experimental results show a reasonable discrimination of these three categories even though coal and nuclear images share highly common and overlapping objects. This research also identified several research challenges associated with nuclear proliferation monitoring using high resolution remote sensing images.

  18. Estimate of radiation-induced steel embrittlement in the BWR core shroud and vessel wall from reactor-grade MOX/UOX fuel for the nuclear power plant at Laguna Verde, Veracruz, Mexico

    Science.gov (United States)

    Vickers, Lisa Rene

    The government of Mexico has expressed interest to utilize the Laguna Verde boiling water reactor (BWR) nuclear power plant for the disposition of reprocessed spent uranium oxide (UOX) fuel in the form of reactor-grade mixed-oxide (MOX) fuel. MOX fuel would replace spent UOX fuel as a fraction in the core from 18--30% depending on the fuel loading cycle. MOX fuel is expected to increase the neutron fluence, flux, fuel centerline temperature, reactor core pressure, and yield higher energy neutrons. There is concern that a core with a fraction of MOX fuel (i.e., increased 239Pu wt%) would increase the radiation-induced steel embrittlement within the core shroud and vessel wall as compared to only conventional, enriched UOX fuel in the core. The evaluation of radiation-induced steel embrittlement within the core shroud and vessel wall is a concern because of the potentially adverse affect to plant and public safety, environment, and operating life of the reactor. This dissertation provides computational results of the neutron fluence, flux, energy spectrum, and radiation damage displacements per atom per second (dpa-s-1) in steel within the core shroud and vessel wall of the Laguna Verde Unit 1 BWR. The results were computed using the nuclear data processing code NJOY99 and the continuous energy Monte Carlo Neutral Particle transport code MCNP4B. The MCNP4B model of the reactor core was for maximum core loading fractions of ⅓ MOX and ⅔ UOX reactor-grade fuel in an equilibrium core. The primary conclusion of this dissertation was that the addition of the maximum fraction of ⅓ MOX fuel to the LV1 BWR core did significantly accelerate the radiation-induced steel embrittlement such that without mitigation of steel embrittlement by periodic thermal annealing or reduction in operating parameters such as, neutron fluence, core temperature and pressure, it posed a potentially adverse affect to the plant and public safety, environment, and operating life of the reactor.

  19. Adrenaline promotes cell proliferation and increases chemoresistance in colon cancer HT29 cells through induction of miR-155

    Energy Technology Data Exchange (ETDEWEB)

    Pu, Jun [Department of General Surgery, Tangdu Hospital of the Fourth Military Medical University, Xi' an 710038 (China); Bai, Danna [Department of Cardiology, 323 Hospital of PLA, Xi' an 710054 (China); Yang, Xia [Department of Teaching and Medical Administration, Tangdu Hospital of the Fourth Military Medical University, Xi' an 710038 (China); Lu, Xiaozhao [Department of Nephrology, The 323 Hospital of PLA, Xi' an 710054 (China); Xu, Lijuan, E-mail: 13609296272@163.com [Department of Nephrology, The 323 Hospital of PLA, Xi' an 710054 (China); Lu, Jianguo, E-mail: lujianguo029@yahoo.com.cn [Department of General Surgery, Tangdu Hospital of the Fourth Military Medical University, Xi' an 710038 (China)

    2012-11-16

    Highlights: Black-Right-Pointing-Pointer Adrenaline increases colon cancer cell proliferation and its resistance to cisplatin. Black-Right-Pointing-Pointer Adrenaline activates NF{kappa}B in a dose dependent manner. Black-Right-Pointing-Pointer NF{kappa}B-miR-155 pathway contributes to cell proliferation and resistance to cisplatin. -- Abstract: Recently, catecholamines have been described as being involved in the regulation of cancer genesis and progression. Here, we reported that adrenaline increased the cell proliferation and decreased the cisplatin induced apoptosis in HT29 cells. Further study found that adrenaline increased miR-155 expression in an NF{kappa}B dependent manner. HT29 cells overexpressing miR-155 had a higher cell growth rate and more resistance to cisplatin induced apoptosis. In contrast, HT29 cells overexpressing miR-155 inhibitor displayed decreased cell proliferation and sensitivity to cisplatin induced cell death. In summary, our study here revealed that adrenaline-NF{kappa}B-miR-155 pathway at least partially contributes to the psychological stress induced proliferation and chemoresistance in HT29 cells, shedding light on increasing the therapeutic strategies of cancer chemotherapy.

  20. Peroxisome Proliferator-Activated Receptor Ligands and Their Role in Chronic Myeloid Leukemia: Therapeutic Strategies.

    Science.gov (United States)

    Yousefi, Bahman; Samadi, Nasser; Baradaran, Behzad; Shafiei-Irannejad, Vahid; Zarghami, Nosratollah

    2016-07-01

    Imatinib therapy remains the gold standard for treatment of chronic myeloid leukemia; however, the acquired resistance to this therapeutic agent in patients has urged the scientists to devise modalities for overcoming this chemoresistance. For this purpose, initially therapeutic agents with higher tyrosine kinase activity were introduced, which had the potential for inhibiting even mutant forms of Bcr-Abl. Furthermore, coupling imatinib with peroxisome proliferator-activated receptor ligands also showed beneficial effects in chronic myeloid leukemia cell proliferation. These combination protocols inhibited cell growth and induced apoptosis as well as differentiation in chronic myeloid leukemia cell lines. In addition, peroxisome proliferator-activated receptors ligands increased imatinib uptake by upregulating the expression of human organic cation transporter 1. Taken together, peroxisome proliferator-activated receptors ligands are currently being considered as novel promising therapeutic candidates for chronic myeloid leukemia treatment, because they can synergistically enhance the efficacy of imatinib. In this article, we reviewed the potential of peroxisome proliferator-activated receptors ligands for use in chronic myeloid leukemia treatment. The mechanism of action of these therapeutics modalities are also presented in detail. © 2016 John Wiley & Sons A/S.

  1. The depletion of nuclear glutathione impairs cell proliferation in 3t3 fibroblasts.

    Directory of Open Access Journals (Sweden)

    Jelena Markovic

    2009-07-01

    Full Text Available Glutathione is considered essential for survival in mammalian cells and yeast but not in prokaryotic cells. The presence of a nuclear pool of glutathione has been demonstrated but its role in cellular proliferation and differentiation is still a matter of debate.We have studied proliferation of 3T3 fibroblasts for a period of 5 days. Cells were treated with two well known depleting agents, diethyl maleate (DEM and buthionine sulfoximine (BSO, and the cellular and nuclear glutathione levels were assessed by analytical and confocal microscopic techniques, respectively. Both agents decreased total cellular glutathione although depletion by BSO was more sustained. However, the nuclear glutathione pool resisted depletion by BSO but not with DEM. Interestingly, cell proliferation was impaired by DEM, but not by BSO. Treating the cells simultaneously with DEM and with glutathione ethyl ester to restore intracellular GSH levels completely prevented the effects of DEM on cell proliferation.Our results demonstrate the importance of nuclear glutathione in the control of cell proliferation in 3T3 fibroblasts and suggest that a reduced nuclear environment is necessary for cells to progress in the cell cycle.

  2. Antibiotic Resistance

    Science.gov (United States)

    ... But there is a growing problem of antibiotic resistance. It happens when bacteria change and become able ... of an antibiotic. Using antibiotics can lead to resistance. Each time you take antibiotics, sensitive bacteria are ...

  3. Effects of high glucose on mesenchymal stem cell proliferation and differentiation

    DEFF Research Database (Denmark)

    Li, Yu-Ming; Schilling, Tatjana; Benisch, Peggy

    2007-01-01

    High glucose (HG) concentrations impair cellular functions and induce apoptosis. Exposition of mesenchymal stem cells (MSC) to HG was reported to reduce colony forming activity and induce premature senescence. We characterized the effects of HG on human MSC in vitro using telomerase...... was not influenced by HG in both cell types. MSC treatment with HG favored osteogenic differentiation. MSC are resistant to HG toxicity, depending on the stemness of MSC. Proliferation and osteogenic differentiation are stimulated by HG. Effects of HG on the transient amplifying compartment of MSC may differ from...... those in mature cells. Further research is needed to unravel the molecular mechanisms of HG resistance of MSC...

  4. Analysis of the microstructural evolution of the damage by neutron irradiation in the pressure vessel of a nuclear power reactor BWR; Analisis de la evolucion microestructural del dano por irradiacion neutronica en la vasija de presion de un reactor nuclear de potencia BWR

    Energy Technology Data Exchange (ETDEWEB)

    Moranchel y R, M.

    2012-07-01

    Nuclear reactor pressure vessel type BWR, installed in Mexico and in many other countries, are made of an alloy of low carbon steel. The American Society for Testing and Materials (Astm) classifies this alloy as A533-B, class 1. Both the vessel and other internal structures are continuously exposed to the neutron flux from the reactions of fission in nuclear fuel. A large number of neutrons reach the vessel and penetrate certain depth depending on their energy. Its penetration in the neutron collides with the nuclei of the atoms out of their positions in the crystal lattice of steel, producing vacancies, interstitial, segregations, among other defects, capable of affecting its mechanical properties. Analyze the micro-structural damage to the vessel due to neutron irradiation, is essential for reasons of integrity of this enclosure and safety of any nuclear power plant. The objective of this thesis work is theoretical and experimentally determine the microstructural damage of a type nuclear reactor vessel steel BWR, due to neutron radiation from the reactor core, using microscopic and spectroscopic techniques as well as Monte Carlo simulation. Microscopy Optical, Scanning Electron Microscopy, Transmission Electron Microscopy, Energy Dispersion of X-rays Spectrometry and X-rays Diffractometry were the techniques used in this research. These techniques helped in the characterization of both the basis of design of pressure vessel steel and steel irradiated, after eight years of neutron irradiation on the vessel, allowing know the surface morphology and crystal structures of the previous steel and post-irradiation, analyze the change in the microstructure of the steel vessel, morphological damage to surface level in an irradiated sample, among which are cavities in the order of microns produced by Atomic displacements due to the impact of neutronic, above all in the first layers of thickness of the vessel, the effect of swelling, regions of greater damage and Atomic

  5. Nuclear Proliferation and Authority in World Politics

    Directory of Open Access Journals (Sweden)

    Brian Frederking

    2009-11-01

    Full Text Available We apply the “security-hierarchy paradox” to nuclear proliferation. Global security requires a certain amount of hierarchy. A world in which no nuclear proliferation rules exist to constrain states, for example, would not be secure. Global security requires legitimate and authoritative rules, which we define as rules that are mutually negotiated, binding to all andwhich provide a stable social order. Too much hierarchy, however, amounts to coercion andundermines global security. Rules that are not mutually negotiated, binding to all or do not provide a stable social order are not authoritative. We argue that North Korea and Iran haveattempted to build nuclear weapons because they interpret the proliferation rules to lack authority. The coercive U.S. approaches to enforcing proliferation rules – including diplomatic isolation, preemption, and regime change – have undermined the legitimacy of those rules. When the U.S. pursues less hierarchical policies, as it has recently toward North Korea, the ensuing negotiations have facilitated progress toward an agreement. When theU.S. pursues a consistently hierarchical approach, as it has toward Iran, no progress is made. Our analysis suggests that it is worth attempting a less hierarchical approach toward Iran and encourage it to accept a deal similar to the one negotiated with North Korea.

  6. Does programmed CTL proliferation optimize virus control?

    DEFF Research Database (Denmark)

    Wodarz, Dominik; Thomsen, Allan Randrup

    2005-01-01

    CD8 T-cell or cytotoxic T-lymphocyte responses develop through an antigen-independent proliferation and differentiation program. This is in contrast to the previous thinking, which was that continuous antigenic stimulation was required. This Opinion discusses why nature has chosen the proliferati...

  7. The Global Chilling Effects of Antidumping Proliferation

    NARCIS (Netherlands)

    Vandenbussche, H.; Zanardi, M.

    2006-01-01

    Advocates of antidumping (AD) laws downplay their effects by arguing that the trade flows that are subject to AD are small and their distortions negligible.This paper is the first to counter that notion by quantifying the worldwide effect of AD laws on aggregate trade flows.The recent proliferation

  8. Luteoloside Inhibits Proliferation of Human Chronic Myeloid ...

    African Journals Online (AJOL)

    Purpose: To investigate the effects of luteoloside on the proliferation of human chronic myeloid leukemia K562 cells and whether luteoloside induces cell cycle arrest and apoptosis in K562 cells. Methods: Luteoloside's cytotoxicity was assessed using a cell counting kit. Cell cycle distribution was analysed by flow cytometry ...

  9. Nuclear war, nuclear proliferation, and their consequences

    Energy Technology Data Exchange (ETDEWEB)

    Sanruddin, A.K.

    1986-01-01

    The proceedings of a colloquium convened by the Groupe de Bellerive offers the contributions of Carl Sagan, Gabriel Garcia Marquez, Kenneth Galbraith, Pierre Trudeau, Edward Kennedy, and other eminent scientists, politicians, and strategists on the subject of the proliferation of nuclear weaponry and its potential ramifications.

  10. [Humoral regulation of stem cell proliferation].

    Science.gov (United States)

    Musashi, M; Ogawa, M

    1991-05-01

    The central feature of hematopoiesis is life-long, stable cell renewal. This process is supported by hemopoietic stem cells which, in the steady state, appear to be dormant in cell cycling. The recruitment of the dormant stem cells into cell cycle may be promoted by such factors as interleukin (IL)-1, IL-6, granulocyte-colony stimulating factor (G-CSF), and newly discovered IL-11. The effects of IL-1 on stem cells may be indirect. Once the stem cells leave Go and begin proliferation, the subsequent process is characterized by continued proliferation and differentiation. Though several models of stem cell differentiation have been proposed, micromanipulation studies of individual progenitors suggest that the commitment of multipotential progenitors to single lineages is a stochastic process. The proliferation of early hemopoietic progenitors requires the presence of IL-3 and/or IL-4, and the intermediate process appears to be supported by granulocyte/macrophage-CSF (GM-CSF). Once the progenitors are committed to individual lineages, the subsequent maturation process appears to be supported by late-acting, lineage-specific factors such as erythropoietin (erythropoiesis), G-CSF (neutrophil production), and IL-5 (eosinophilopoiesis). Thus, hemopoietic proliferation appears to be regulated by a cascade of factors directed at different developmental stages.

  11. Fibrosarcoma in bizarre parosteal osteochondromatous proliferation

    Energy Technology Data Exchange (ETDEWEB)

    Choi, J.H.; Gu, M.J.; Kim, M.J.; Choi, W.H. [Dept. of Pathology, College of Medicine, Yeungnam Univ. (Korea); Shin, D.S. [Dept. of Orthopedic Surgery, College of Medicine, Yeungnam Univ. (Korea); Cho, K.H. [Dept. of Diagnostic Radiology, College of Medicine, Yeungnam Univ. (Korea)

    2001-01-01

    Bizarre parosteal osteochondromatous proliferation (BPOP) is a rare benign lesion predominantly involving the small bones of the hands and feet. Malignant transformation in BPOP has not been documented in the English literature. This report presents the coexistence of fibrosarcoma with BPOP in the right distal fibula of an 18-year-old woman. (orig.)

  12. Immunohistochemical Assessment of Proliferating Cell Nuclear ...

    African Journals Online (AJOL)

    Background: Proliferating cell nuclear antigen (PCNA) is a nuclear protein synthesized in the late G1 and S‑phase of the cell cycle. ... Two sections were taken from each one for H and E. Other sections were stained according to super sensitive polymer horseradish peroxidase method for identifying PCNA expression.

  13. Effect of chloroquine on human lymphocyte proliferation

    DEFF Research Database (Denmark)

    Bygbjerg, Ib Christian; Flachs, H

    1986-01-01

    The effect of chloroquine on human blood mononuclear cells was studied. High concentrations of chloroquine in vitro profoundly suppressed the proliferation of mitogen- and antigen-stimulated cells, as indicated by decreased 14C-thymidine incorporation. Lower concentrations of chloroquine increase...

  14. Transcriptional peroxisome proliferator-activated receptor γ ...

    African Journals Online (AJOL)

    user

    Peroxisome proliferator-activated receptor γ coactivator (PGC)-1ɑ, a well-known member of PGC-1 transcriptional coactivator's family, plays a key role in various metabolic pathways. Here, we investigated the role of PGC-1ɑ in the transformation of muscle fiber type in Schizothorax prenanti. The expression of PGC-1ɑ was ...

  15. Proliferation, angiogenesis and differentiation related markers in ...

    African Journals Online (AJOL)

    The upregulation of these proliferation- and angiogenesis-related factors in endothelial cells and/or fibroblasts and not in follicular cells of compact carcinoma compared to healthy glands supports the relevance of stromal cells in cancer progression. Keywords: Canine, Histology, Immunohistochemistry, Thyroid carcinoma.

  16. Evaluation of the reduction of boron-10 in the control rods in the BWR of the Laguna Verde Central, through steady state calculations; Evaluacion de la reduccion del Boro-10 en las barras de control en los BWR de la CLV, mediante calculos en estado estacionario

    Energy Technology Data Exchange (ETDEWEB)

    Montes T, J.L.; Perusquia, R.; Hernandez, J.L.; Ramirez S, J.R. [Departamento de Sistemas Nucleares, ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2003-07-01

    One of the more important aspects related with the safety and economy in the operation of a nuclear power reactor, it is without a doubt the control of the reactivity. During the normal operation of a reactor of boiling water (BWR-Boiling Water Reactor), the control of the reactivity in the nucleus it is strongly determined by the efficiency of the control rods. In the case of the Laguna Verde Nuclear power station (CNLV) the nucleus of the reactors has 109 control rods grouped in 4 sets. The CNLV at the moment uses the CCC method (Control Cell Core) in the design of the cycle. With this method only the A2 group is used for the control of the reactivity at full power. With the purpose of quantifying the effect of the decrease of the burnable poison (B{sub 4}C) of the control rods and in particular to the effect due to the postulated lost of 10% of Boron 10, it was carried out a series of calculations of the nucleus in stationary state by means of the system of HELIOS/CM-PRESTO codes. In this work the main derived results of these 3D simulations(three dimensions) of the reactors of the CNLV are presented. It was analyzed the one behavior of the infinite neutron multiplication factor (K{sub infinite}), at fuel assemble cell level used in an equilibrium cycle for the CNLV. It was also analyzed the effect in the shutdown margin (ShutDown Margin- SDM) in cold condition CZP (Cold Zero Power). Its are also included those results of the ARI cases (All Rods In) and SRO (Strong Rod Out). From the cases in condition HFP (Hot Full Power) the behavior of the effective multiplication factor (K{sub eff}) is presented. (Author)

  17. Simulation of a large break loss of coolant (LBLOCA), without actuation of the emergency injection systems (ECCS) for a BWR-5; Simulacion de un escenario de perdida de refrigerante grande (LBLOCA), sin actuacion de los sistemas de inyeccion de emergencia (ECCS) para un reactor BWR-5

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas V, J.; Mugica R, C. A.; Lopez M, R., E-mail: jaime.cardenas@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779, Col. Narvarte, 03020 Ciudad de Mexico (Mexico)

    2015-09-15

    In this paper the analysis of scenario for the loss of coolant case was realized with break at the bottom of a recirculation loop of a BWR-5 with containment type Mark II and a thermal power of 2317 MWt considering that not have coolant injection. This in order to observe the speed of progression of the accident, the phenomenology of the scenario, the time to reach the limit pressure of containment venting and the amount of radionuclides released into the environment. This simulation was performed using the MELCOR code version 2.1. The scenario posits a break in one of the shear recirculation loops. The emergency core cooling system (ECCS) and the reactor core isolation cooling (Rcic) have not credit throughout the event, which allowed achieve greater severity on scenario. The venting of the primary containment was conducted via valve of 30 inches instead of the line of 24 inches of wet well, this in order to have a larger area of exhaust of fission products directly to the reactor building. The venting took place when the pressure in the primary containment reached the 4.5 kg/cm{sup 2} and remained open for the rest of the scenario to maximize the amount released of radionuclides to the atmosphere. The safety relief valves were considered functional they do not present mechanical failure or limit their ability to release pressure due to the large number of performances in safety mode. The results of the analysis covers about 48 hours, time at which the accident evolution was observed; behavior of level, pressure in the vessel and the fuel temperature profile was analyzed. For progression of the scenario outside the vessel, the pressure and temperature of the primary containment, level and temperature of the suppression pool, the hydrogen accumulation in the container and the radionuclides mass released into the atmosphere were analyzed. (Author)

  18. Simulation of a scenario of total loss of external and internal power (Sbo) for different vent pressures of the containment of a BWR-5; Simulacion de un escenario de perdida total de potencia externa e interna (SBO) para distintas presiones de venteo de la contencion de un reactor BWR-5

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas V, J.; Mugica R, C. A.; Godinez S, V., E-mail: Jaime.cardenas@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

    2014-10-15

    The simulation of a Station Black Out (Sbo) was realized with intervention of the vent containment by means of a rigid vent coming from the dry-well and that discharges directly to the atmosphere, with the MELCOR code version 2.1. This scenario was carried out for a BWR-5 and containment type Mark II, with a thermal power of 2317 MWt similar to the reactor of nuclear power plant of Laguna Verde. For this scenario was considered as only available system for coolant injection to the reactor to the Reactor Core Isolation Cooling (Rcic), which remained operating 4 hours with batteries bank. The Security and Relief Valves (SR V) were considered functional (by simplicity) and that they mechanically do not exceed their capacity to liberate pressure due to the performances in their safety way. The operator maneuver to perform the SR V and to de pressurize the vessel until the pressure (13 kg/cm{sup 2}) to operate the low pressure systems was modeled. The results cover approximately 48 hours (172000 seconds), time in which was observed the behavior of the level and pressure in the vessel. Also the scenario evolution was analyzed to different vent pressures of the primary containment (2.0, 3.0, 4.5, 6.0, and 10.0 kg/cm{sup 2}), the temperature profiles of the dry-well, the hydrogen accumulation in the containment, the radio-nuclides liberation through rigid vent to the atmosphere and the inventory of these. In this work an analysis of the pressure behavior in the primary containment is presented, with the purpose of minimizing liberated fission products to the environment. (Author)

  19. Modelling of the dynamics of the vessel and circuits of recirculation of a BWR type nucleo electric as part of the SUN-RAH university simulator; Modelado de la dinamica de la vasija y circuitos de recirculacion de una nucleoelectrica tipo BWR como parte del simulador universitario SUN-RAH

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez S, R.A. [DEPFI, Campus Morelos, en IMTA, Jiutepec, Morelos (Mexico)]. e-mail: rsanchez_15@yahoo.com.mx

    2003-07-01

    In the present project, the development of a model for the dynamics of the process of energy transport generated in the nuclear fuel until the main steam lines of a nucleo electric central with BWR type nuclear reactor, using mathematical models of reduced order is presented. These models present the main characteristics of the reactor vessel and of the recirculation system, defined by the main phenomena that intervene in those physical processes. Likewise, the objective of the general project of the one University student nucleo electric simulator with Boiling Water Reactor (SUN-RAH) for later on to establish the modeling equations for each part of the nuclear reactor as well as of the load pursuit system. Also, its were described the graphic interfaces implemented in an three layers architecture in which the different measuring variables are presented in the monitor. It fits signalize that the advantage presented by the University student nucleo electric simulator is the possibility to carry out changes in the magnitudes of those different variables that intervene in the physical processes made in the one reactor and in the recirculation system in execution time of the same one. Of same way, the creation of a graphic intuitive interface, friendly, and designed with the same technology with the one that the video games are programmed in the present time. Besides all the above mentioned, the pending goals inside of the project are exposed, as well as the developments in construction process or conceptualized to be included in future versions of the simulator. Finally its are thinking about possible scenarios of applications of SUN-RAH, as well as their reaches. (Author)

  20. Post-processor for simulations of the ORIGEN program and calculation of the composition of the activity of a burnt fuel core by a BWR type reactor; Post-procesador para simulaciones del programa ORIGEN y calculo de la composicion de la actividad de un nucleo de combustible quemado por un reactor tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Sandoval V, S. [IIE, Av. Reforma 113, Col. Palmira, 62490 Cuernavaca, Morelos (Mexico)]. e-mail: sandoval@iie.org.mx

    2006-07-01

    The composition calculation and the activity of nuclear materials subject to processes of burnt, irradiation and decay periods are of utility for diverse activities inside the nuclear industry, as they are it: the processes design and operations that manage radioactive material, the calculation of the inventory and activity of a core of burnt nuclear fuel, for studies of type Probabilistic Safety Analysis (APS), as well as for regulation processes and licensing of nuclear facilities. ORIGEN is a program for computer that calculates the composition and the activity of nuclear materials subject to periods of burnt, irradiation and decay. ORIGEN generates a great quantity of information whose processing and analysis are laborious, and it requires thoroughness to avoid errors. The automation of the extraction, conditioning and classification of that information is of great utility for the analyst. By means of the use of the post-processor presented in this work it is facilitated, it speeds up and wide the capacity of analysis of results, since diverse consultations with several classification options and filtrate of results can be made. As illustration of the utility of the post-processor, and as an analysis of interest for itself, it is also presented in this work the composition of the activity of a burned core in a BWR type reactor according to the following classification criteria: by type of radioisotope (fission products, activation products and actinides), by specie type (gassy, volatile, semi-volatile and not volatile), by element and by chemical group. The results show that the total activity of the studied core is dominated by the fission products and for the actinides, in proportion four to one, and that the gassy and volatile species conform a fifth part of the total activity of the core. (Author)

  1. Nuclear non proliferation and disarmament; Non-proliferation nucleaire et desarmement

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-07-01

    In the framework of the publication of a document on the ''weapons mastership, disarmament and non proliferation: the french action'', by the ministry of Foreign Affairs and the ministry of Defense, the French Documentation organization presents a whole document. This document describes and details the following topics: the conference on the treaty of non proliferation of nuclear weapons, the France, Usa and Non Governmental Organizations position, the threats of the proliferation, the french actions towards the disarmament, the disarmament in the world, a chronology and some bibliographic resources. (A.L.B.)

  2. Blue light inhibits proliferation of melanoma cells

    Science.gov (United States)

    Becker, Anja; Distler, Elisabeth; Klapczynski, Anna; Arpino, Fabiola; Kuch, Natalia; Simon-Keller, Katja; Sticht, Carsten; van Abeelen, Frank A.; Gretz, Norbert; Oversluizen, Gerrit

    2016-03-01

    Photobiomodulation with blue light is used for several treatment paradigms such as neonatal jaundice, psoriasis and back pain. However, little is known about possible side effects concerning melanoma cells in the skin. The aim of this study was to assess the safety of blue LED irradiation with respect to proliferation of melanoma cells. For that purpose we used the human malignant melanoma cell line SK-MEL28. Cell proliferation was decreased in blue light irradiated cells where the effect size depended on light irradiation dosage. Furthermore, with a repeated irradiation of the melanoma cells on two consecutive days the effect could be intensified. Fluorescence-activated cell sorting with Annexin V and Propidium iodide labeling did not show a higher number of dead cells after blue light irradiation compared to non-irradiated cells. Gene expression analysis revealed down-regulated genes in pathways connected to anti-inflammatory response, like B cell signaling and phagosome. Most prominent pathways with up-regulation of genes were cytochrome P450, steroid hormone biosynthesis. Furthermore, even though cells showed a decrease in proliferation, genes connected to the cell cycle were up-regulated after 24h. This result is concordant with XTT test 48h after irradiation, where irradiated cells showed the same proliferation as the no light negative control. In summary, proliferation of melanoma cells can be decreased using blue light irradiation. Nevertheless, the gene expression analysis has to be further evaluated and more studies, such as in-vivo experiments, are warranted to further assess the safety of blue light treatment.

  3. Peroxisome proliferator-activated receptor-gamma (PPARgamma) Pro12Ala polymorphism and risk for pediatric obesity

    NARCIS (Netherlands)

    Dedoussis, George V; Vidra, Nikoleta; Butler, Johannah; Papoutsakis, Constantina; Yannakoulia, Mary; Hirschhorn, Joel N; Lyon, Helen N; Vidra, Nikoletta

    BACKGROUND: Variation in the peroxisome-proliferator-activated receptor gamma (PPARgamma) gene has been reported to alter the risk for adiposity in adults. METHODS: We investigated the gender related association between the Pro12Ala variant (rs1801282) in obesity and insulin resistance traits in 794

  4. Updating of the costs of the nuclear fuels of the equilibrium reloading of the A BWR and EPR reactors; Actualizacion de los costos de combustible nuclear de la recarga de equilibrio de los reactores ABWR y EPR

    Energy Technology Data Exchange (ETDEWEB)

    Ortega C, R.F. [FI-UNAM, 04510 Mexico D.F. (Mexico)]. e-mail: rortega@fi-b.unam.mx

    2008-07-01

    In the last two and a half years, the price of the uranium in the market spot has ascended of US$20.00 dollars by lb U{sub 3O}8 in January, 2005 to a maximum of US$137.00 dollars by Ib U{sub 3}O{sub 8} by the middle of 2007. At the moment this price has been stabilized in US$90.00 dollars by Ib U{sub 3}O{sub 8} such for the market spot, like for the long term contracts. In this work the reasons of this increment are analyzed, as well as their impact in the fuel prices of the balance recharge of the advanced reactors of boiling water (A BWR) and of the advanced water at pressure reactors (EPR). (Author)

  5. CFD analysis for the hydrogen transport in the primary contention of a BWR using the codes OpenFOAM and Gas-Flow; Analisis CFD para el transporte de hidrogeno en la contencion primaria de un reactor BWR usando los codigos OpenFOAM y GasFlow

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez P, D. A.

    2014-07-01

    using a limited number of semi-empirical data, and instead, mathematical relationships are used taking into account the various physical phenomena as well the interactions that occur among them, such as heat transfer between the fluid and the solid walls condensation of water vapor on the walls, the turbulent effects in areas of restricted passage, etc. Taking into account these advantages, this study presents a qualitative and quantitative comparison between the CFD codes OpenFOAM and Gas-Flow related to the transport phenomena of Hydrogen and other gases in the primary containment of a BWR reactor. Gas-Flow is a code of commercial license that is well validated, developed in Germany to analyze the transport of gases in nuclear reactor containments. On the other hand, OpenFOAM is an open source CFD code offering several solvers for different phenomena assessments, in this work, the reacting Foam solver is used because it has a strong similarity to the intended application of Hydrogen transport. In this thesis the results obtained using the reacting Foam solver of OpenFOAM for the calculation of transport of Hydrogen are compared with the results of the Gas-Flow code in order to assess if it is feasible to use the open source code OpenFOAM in the case of Hydrogen transport in primary containment of a BWR reactor. Some differences in the qualitative and quantitative results from both codes were found, the differences (with a maximum error rate of 4%) in the quantitative results were found are small and are considered more than acceptable for this type of analysis, moreover, these differences are mainly attributed to the transport models used, mainly because OpenFOAM uses a homogeneous mixture model and Gas-Flow a heterogeneous one. Implementing appropriate solvers in codes like OpenFOAM has the goal to develop own tools that are applicable to the transport of Hydrogen in the primary containment of a BWR reactor and thus, to gain some independence while not relying on

  6. Thorium-based fuel cycles: Reassessment of fuel economics and proliferation risk

    Energy Technology Data Exchange (ETDEWEB)

    Serfontein, Dawid E., E-mail: Dawid.Serfontein@nwu.ac.za [Senior Lecturer at the School of Mechanical and Nuclear Engineering, North West University (PUK-Campus), PRIVATE BAG X6001, Internal Post Box 360, Potchefstroom 2520 (South Africa); Mulder, Eben J. [Professor at the School of Mechanical and Nuclear Engineering, North West University (South Africa)

    2014-05-01

    At current consumption and current prices, the proven reserves for natural uranium will last only about 100 years. However, the more abundant thorium, burned in breeder reactors, such as large High Temperature Gas-Cooled Reactors, and followed by chemical reprocessing of the spent fuel, could stretch the 100 years for uranium supply to 15,000 years. Thorium-based fuel cycles are also viewed as more proliferation resistant compared to uranium. However, several barriers to entry caused all countries, except India and Russia, to abandon their short term plans for thorium reactor projects, in favour of uranium/plutonium fuel cycles. In this article, based on the theory of resonance integrals and original analysis of fast fission cross sections, the breeding potential of {sup 232}Th is compared to that of {sup 238}U. From a review of the literature, the fuel economy of thorium-based fuel cycles is compared to that of natural uranium-based cycles. This is combined with a technical assessment of the proliferation resistance of thorium-based fuel cycles, based on a review of the literature. Natural uranium is currently so cheap that it contributes only about 10% of the cost of nuclear electricity. Chemical reprocessing is also very expensive. Therefore conservation of natural uranium by means of the introduction of thorium into the fuel is not yet cost effective and will only break even once the price of natural uranium were to increase from the current level of about $70/pound yellow cake to above about $200/pound. However, since fuel costs constitutes only a small fraction of the total cost of nuclear electricity, employing reprocessing in a thorium cycle, for the sake of its strategic benefits, may still be a financially viable option. The most important source of the proliferation resistance of {sup 232}Th/{sup 233}U fuel cycles is denaturisation of the {sup 233}U in the spent fuel by {sup 232}U, for which the highly radioactive decay chain potentially poses a large

  7. Antibiotic Resistance

    DEFF Research Database (Denmark)

    Munck, Christian

    of antimicrobial resistance: (1) adaptive mutations and (2) horizontal acquisition of resistance genes from antibiotic gene reservoirs. By studying the geno- and phenotypic changes of E. coli in response to single and drug-pair exposures, I uncover the evolutionary trajectories leading to adaptive resistance. I......Bacteria can avoid extinction during antimicrobial exposure by becoming resistant. They achieve this either via adaptive mutations or horizontally acquired resistance genes. If resistance emerges in clinical relevant species, it can lead to treatment failure and ultimately result in increasing...... morbidity and mortality as well as an increase in the cost of treatment. Understanding how bacteria respond to antibiotic exposure gives the foundations for a rational approach to counteract antimicrobial resistance. In the work presented in this thesis, I explore the two fundamental sources...

  8. Cyclic AMP inhibition of proliferation of hepatocellular carcinoma cells is mediated by Akt.

    Science.gov (United States)

    Liu, Lunhua; Xie, Yili; Lou, Liguang

    2005-11-01

    Cyclic AMP (cAMP), one of the most important intracellular second messengers, has been reported to inhibit proliferation of human hepatocellular carcinoma (HCC) cells via negatively regulating p42/44 mitogen-activated protein kinase. Here, we reported that cAMP inhibited the proliferation of HCC BEL-7402 cells via a novel mechanism. Forskolin, an activator of adenylate cyclase, inhibited fetal bovine serum (FBS)-stimulated BEL-7402 cell proliferation in a dose- and time-dependent manner, along with the inhibition of FBS-stimulated serine/threoine protein kinase Akt (also known as PKB) phosphorylation which is required for Akt activation and this effect was mimicked by 8-Br cAMP. Forskolin also inhibited Akt phosphorylation stimulated by other growth factors such as IGF-1, epidermal growth factor, and insulin. These inhibitions were found not only in BEL-7402 cells, but also in another HCC cell line SMMC-7721 cells. Myr-Akt (myristolated-Akt), a constitutively active Akt which was relatively resistant to cAMP inhibition, conferred BEL-7402 cells resistance to cAMP treatment. However, overexpression of Myr-Akt alone was not sufficient to stimulate BEL-7402 cell proliferation. cAMP inhibited FBS-stimulated Akt phosphorylation in a cAMP-dependent protein kinase-dependent manner. Further studies demonstrated that cAMP inhibited FBS-induced membrane localization of 3-phosphoinositide-dependent kinase 1 (PDK-1) which is a required process for PDK-1 to phosphorylate Akt, but had no significant effect on phosphoinositide 3-kinase activity. These results indicate that cAMP inhibition of proliferation of HCC cells is mediated by Akt and cAMP inhibits Akt activation via blocking membrane localization of PDK-1.

  9. Creating Genetic Resistance to HIV

    Science.gov (United States)

    Burnett, John C.; Zaia, John A.; Rossi, John J.

    2012-01-01

    HIV/AIDS remains a chronic and incurable disease, in spite of the notable successes of highly active antiretroviral therapy. Gene therapy offers the prospect of creating genetic resistance to HIV that supplants the need for antiviral drugs. In sight of this goal, a variety of anti-HIV genes have reached clinical testing, including gene-editing enzymes, protein-based inhibitors, and RNA-based therapeutics. Combinations of therapeutic genes against viral and host targets are designed to improve the overall antiviral potency and reduce the likelihood of viral resistance. In cell-based therapies, therapeutic genes are expressed in gene modified T lymphocytes or in hematopoietic stem cells that generate an HIV-resistant immune system. Such strategies must promote the selective proliferation of the transplanted cells and the prolonged expression of therapeutic genes. This review focuses on the current advances and limitations in genetic therapies against HIV, including the status of several recent and ongoing clinical studies. PMID:22985479

  10. New approaches to nuclear proliferation policy.

    Science.gov (United States)

    Nye, J S

    1992-05-29

    Nuclear proliferation is not one but a complex of problems. One relates to the collapse of the Soviet Union and its effect on the spread of nuclear weapons and knowledge. Second, Iraq's violation of its Non-Proliferation Treaty obligation has exposed certain weaknesses in the traditional regime of multilateral nonproliferation institutions and treaties. Third, Pakistan's achievement of a nuclear weapons capability in the late 1980s brings the postproliferation question to the forefront in South Asia. There is no single solution to this complex set of problems, but the beginning of wisdom is to build upon the successes of the past, add new policy procedures, and, above all, increase the priority given to the issue. Otherwise, we may be faced with the ironic outcome that the widely welcomed end of the Cold War may increase the prospect of nuclear use.

  11. XIAP antagonist embelin inhibited proliferation of cholangiocarcinoma cells.

    Directory of Open Access Journals (Sweden)

    Cody J Wehrkamp

    Full Text Available Cholangiocarcinoma cells are dependent on antiapoptotic signaling for survival and resistance to death stimuli. Recent mechanistic studies have revealed that increased cellular expression of the E3 ubiquitin-protein ligase X-linked inhibitor of apoptosis (XIAP impairs TRAIL- and chemotherapy-induced cytotoxicity, promoting survival of cholangiocarcinoma cells. This study was undertaken to determine if pharmacologic antagonism of XIAP protein was sufficient to sensitize cholangiocarcinoma cells to cell death. We employed malignant cholangiocarcinoma cell lines and used embelin to antagonize XIAP protein. Embelin treatment resulted in decreased XIAP protein levels by 8 hours of treatment with maximal effect at 16 hours in KMCH and Mz-ChA-1 cells. Assessment of nuclear morphology demonstrated a concentration-dependent increase in nuclear staining. Interestingly, embelin induced nuclear morphology changes as a single agent, independent of the addition of TNF-related apoptosis inducing ligand (TRAIL. However, caspase activity assays revealed that increasing embelin concentrations resulted in slight inhibition of caspase activity, not activation. In addition, the use of a pan-caspase inhibitor did not prevent nuclear morphology changes. Finally, embelin treatment of cholangiocarcinoma cells did not induce DNA fragmentation or PARP cleavage. Apoptosis does not appear to contribute to the effects of embelin on cholangiocarcinoma cells. Instead, embelin caused inhibition of cell proliferation and cell cycle analysis indicated that embelin increased the number of cells in S and G2/M phase. Our results demonstrate that embelin decreased proliferation in cholangiocarcinoma cell lines. Embelin treatment resulted in decreased XIAP protein expression, but did not induce or enhance apoptosis. Thus, in cholangiocarcinoma cells the mechanism of action of embelin may not be dependent on apoptosis.

  12. Director`s series on proliferation

    Energy Technology Data Exchange (ETDEWEB)

    Bailey, K.C.; Price, M.E. [eds.

    1994-10-17

    This series is an occasional publication of essays on the topics of nuclear, chemical, biological, and missile proliferation. Essays contained in this document include: Key issues on NPT renewal and extension, Africa and nuclear nonproliferation, Kenya`s views on the NPT, Prospects for establishing a zone free of weapons of mass destruction in the middle east, effects of a special nuclear weapon materials cut-off convention, and The UK view of NPT renewal.

  13. Proliferation Persuasion. Coercive Bargaining with Nuclear Technology

    Energy Technology Data Exchange (ETDEWEB)

    Volpe, Tristan A. [George Washington Univ., Washington, DC (United States)

    2015-08-31

    Why do states wait for prolonged periods of time with the technical capacity to produce nuclear weapons? Only a handful of countries have ever acquired the sensitive nuclear fuel cycle technology needed to produce fissile material for nuclear weapons. Yet the enduring trend over the last five decades is for these states to delay or forgo exercising the nuclear weapons option provided by uranium enrichment or plutonium reprocessing capabilities. I show that states pause at this threshold stage because they use nuclear technology to bargain for concessions from both allies and adversaries. But when does nuclear latency offer bargaining benefits? My central argument is that challengers must surmount a dilemma to make coercive diplomacy work: the more they threaten to proliferate, the harder it becomes to reassure others that compliance will be rewarded with nuclear restraint. I identify a range of mechanisms able to solve this credibility problem, from arms control over breakout capacity to third party mediation and confidence building measures. Since each step towards the bomb raises the costs of implementing these policies, a state hits a sweet spot when it first acquires enrichment and/or reprocessing (ENR) technology. Subsequent increases in proliferation capability generate diminishing returns at the bargaining table for two reasons: the state must go to greater lengths to make a credible nonproliferation promise, and nuclear programs exhibit considerable path dependency as they mature over time. Contrary to the conventional wisdom about power in world politics, less nuclear latency thereby yields more coercive threat advantages. I marshal new primary source evidence from archives and interviews to identify episodes in the historical record when states made clear decisions to use ENR technology as a bargaining chip, and employ this theory of proliferation persuasion to explain how Japan, North Korea, and Iran succeeded and failed to barter concessions from the

  14. Anticoagulant Resistance

    DEFF Research Database (Denmark)

    Heiberg, Ann-Charlotte

    to represent different sewer rat management strategies i) no anticoagulants for approx. 20 years ii) no anticoagulants for the last 5 years and iii) continuous control for many years. Animals were tested for resistance to bromadiolone by Blood-Clotting Response test, as bromadiolone is the most frequently used...... agent in the sewers. Low level of resistance was found in locations regardless of management strategy. Mutations in the VKORC1 gene have been proposed to confer anticoagulant resistance and an Y139C VKORC1 mutation has been identified in Danish resistant rats. All animals were tested with an Y139C...... specific PCR to verify this genetic form of resistance, but in contrast to animals tested from various surface populations, we could not confirm the Y139C mutation in any of the sewer rats. Our findings could indicate that resistance in surface and sewer population may be caused by different mechanism...

  15. Camptothecin resistance

    DEFF Research Database (Denmark)

    Brangi, M; Litman, Thomas; Ciotti, M

    1999-01-01

    The mitoxantrone resistance (MXR) gene encodes a recently characterized ATP-binding cassette half-transporter that confers multidrug resistance. We studied resistance to the camptothecins in two sublines expressing high levels of MXR: S1-M1-80 cells derived from parental S1 colon cancer cells...... and MCF-7 AdVp3,000 isolated from parental MCF-7 breast cancer cells. Both cell lines were 400- to 1,000-fold more resistant to topotecan, 9-amino-20(S)-camptothecin, and the active metabolite of irinotecan, 7-ethyl-10-hydroxycamptothecin (SN-38), than their parental cell lines. The cell lines...... demonstrated much less resistance to camptothecin and to several camptothecin analogues. Reduced accumulation and energy-dependent efflux of topotecan was demonstrated by confocal microscopy. A significant reduction in cleavable complexes in the resistant cells could be observed after SN-38 treatment...

  16. NSAIDs and Cell Proliferation in Colorectal Cancer

    Directory of Open Access Journals (Sweden)

    Raj Ettarh

    2010-06-01

    Full Text Available Colon cancer is common worldwide and accounts for significant morbidity and mortality in patients. Fortunately, epidemiological studies have demonstrated that continuous therapy with NSAIDs offers real promise of chemoprevention and adjunct therapy for colon cancer patients. Tumour growth is the result of complex regulation that determines the balance between cell proliferation and cell death. How NSAIDs affect this balance is important for understanding and improving treatment strategies and drug effectiveness. NSAIDs inhibit proliferation and impair the growth of colon cancer cell lines when tested in culture in vitro and many NSAIDs also prevent tumorigenesis and reduce tumour growth in animal models and in patients, but the relationship to inhibition of tumour cell proliferation is less convincing, principally due to gaps in the available data. High concentrations of NSAIDs are required in vitro to achieve cancer cell inhibition and growth retardation at varying time-points following treatment. However, the results from studies with colon cancer cell xenografts are promising and, together with better comparative data on anti-proliferative NSAID concentrations and doses (for in vitro and in vivo administration, could provide more information to improve our understanding of the relationships between these agents, dose and dosing regimen, and cellular environment.

  17. Lysophosphatidic Acid Up-Regulates Hexokinase II and Glycolysis to Promote Proliferation of Ovarian Cancer Cells.

    Science.gov (United States)

    Mukherjee, Abir; Ma, Yibao; Yuan, Fang; Gong, Yongling; Fang, Zhenyu; Mohamed, Esraa M; Berrios, Erika; Shao, Huanjie; Fang, Xianjun

    2015-09-01

    Lysophosphatidic acid (LPA), a blood-borne lipid mediator, is present in elevated concentrations in ascites of ovarian cancer patients and other malignant effusions. LPA is a potent mitogen in cancer cells. The mechanism linking LPA signal to cancer cell proliferation is not well understood. Little is known about whether LPA affects glucose metabolism to accommodate rapid proliferation of cancer cells. Here we describe that in ovarian cancer cells, LPA enhances glycolytic rate and lactate efflux. A real time PCR-based miniarray showed that hexokinase II (HK2) was the most dramatically induced glycolytic gene to promote glycolysis in LPA-treated cells. Analysis of the human HK2 gene promoter identified the sterol regulatory element-binding protein as the primary mediator of LPA-induced HK2 transcription. The effects of LPA on HK2 and glycolysis rely on LPA2, an LPA receptor subtype overexpressed in ovarian cancer and many other malignancies. We further examined the general role of growth factor-induced glycolysis in cell proliferation. Like LPA, epidermal growth factor (EGF) elicited robust glycolytic and proliferative responses in ovarian cancer cells. Insulin-like growth factor 1 (IGF-1) and insulin, however, potently stimulated cell proliferation but only modestly induced glycolysis. Consistent with their differential effects on glycolysis, LPA and EGF-dependent cell proliferation was highly sensitive to glycolytic inhibition while the growth-promoting effect of IGF-1 or insulin was more resistant. These results indicate that LPA- and EGF-induced cell proliferation selectively involves up-regulation of HK2 and glycolytic metabolism. The work is the first to implicate LPA signaling in promotion of glucose metabolism in cancer cells. Copyright © 2015 The Authors. Published by Elsevier Inc. All rights reserved.

  18. Nilotinib impairs skeletal myogenesis by increasing myoblast proliferation

    Directory of Open Access Journals (Sweden)

    Osvaldo Contreras

    2018-02-01

    Full Text Available Abstract Background Tyrosine kinase inhibitors (TKIs are effective therapies with demonstrated antineoplastic activity. Nilotinib is a second-generation FDA-approved TKI designed to overcome Imatinib resistance and intolerance in patients with chronic myelogenous leukemia (CML. Interestingly, TKIs have also been shown to be an efficient treatment for several non-malignant disorders such fibrotic diseases, including those affecting skeletal muscles. Methods We investigated the role of Nilotinib on skeletal myogenesis using the well-established C2C12 myoblast cell line. We evaluated the impact of Nilotinib during the time course of skeletal myogenesis. We compared the effect of Nilotinib with the well-known p38 MAPK inhibitor SB203580. MEK1/2 UO126 and PI3K/AKT LY294002 inhibitors were used to identify the signaling pathways involved in Nilotinib-related effects on myoblast. Adult primary myoblasts were also used to corroborate the inhibition of myoblasts fusion and myotube-nuclei positioning by Nilotinib. Results We found that Nilotinib inhibited myogenic differentiation, reducing the number of myogenin-positive myoblasts and decreasing myogenin and MyoD expression. Furthermore, Nilotinib-mediated anti-myogenic effects impair myotube formation, myosin heavy chain expression, and compromise myotube-nuclei positioning. In addition, we found that p38 MAPK is a new off-target protein of Nilotinib, which causes inhibition of p38 phosphorylation in a similar manner as the well-characterized p38 inhibitor SB203580. Nilotinib induces the activation of ERK1/2 and AKT on myoblasts but not in myotubes. We also found that Nilotinib stimulates myoblast proliferation, a process dependent on ERK1/2 and AKT activation. Conclusions Our findings suggest that Nilotinib may have important negative effects on muscle homeostasis, inhibiting myogenic differentiation but stimulating myoblasts proliferation. Additionally, we found that Nilotinib stimulates the activation

  19. Full MOX core in BWR

    Energy Technology Data Exchange (ETDEWEB)

    Aoyama, Motoo [Power and Industrial Systems R and D Laboratory, Hitachi Ltd., Hitachi, Ibaraki (Japan)

    1999-12-01

    Studies on the core design, the fuel rod thermal-mechanical design and the safety evaluation have been summarized for the Full MOX-ABWR, loaded with MOX fuels up to 100% of the core. Fuel bundle configuration for MOX fuels is identical to the STEP II fuel design and the discharge burnup is about 33 GWd/t. Core performance evaluations and fuel rod thermal-mechanical design analyses have been performed, and it has been confirmed that the design criteria are satisfied with enough margin like the UO{sub 2} fuel loaded core. Safety analyses on transients and accidents have also been performed by considering the MOX fuel and core characteristics adequately through selecting appropriate input data for each safety analysis. All safety criteria are satisfied like the UO{sub 2} core. (author)

  20. On the future of civilian plutonium: An assessment of technological impediments to nuclear terrorism and proliferation

    Science.gov (United States)

    Avedon, Roger Edmond

    This dissertation addresses the value of developing diversion- and theft-resistant nuclear power technology, given uncertain future demand for nuclear power, and uncertain risks of nuclear terrorism and of proliferation from the reprocessing of civilian plutonium. The methodology comprises four elements: Economics. An economic growth model coupled with market penetration effects for plutonium and for the hypothetical new technology provides a range of estimates for future nuclear demand. A flow model accounts for the longevity of capital assets (nuclear plants) over time. Terrorism. The commercial nuclear fuel cycle may provide a source of fissile material for terrorists seeking to construct a crude nuclear device. An option value model is used to estimate the effects of the hypothetical new technology on reducing the probability of theft. A game theoretic model is used to explore the deterrence value of physical security and then to draw conclusions about how learning on the part of terrorists or security forces might affect the theft estimate. The principal uncertainties in the theft model can be updated using Bayesian techniques as new data emerge. Proliferation. Access to fissile material is the principal technical impediment to a state's acquisition of nuclear weapons. A game theoretic model is used to determine the circumstances under which a state may proliferate via diversion. The model shows that the hypothetical new technology will have little value for counter-proliferation if diversion is not a preferred proliferation method. A technology policy analysis of the choice of proliferation method establishes that diversion is unlikely to be used because it has no constituency among the important parties to the decision, namely the political leadership, the scientific establishment, and the military. Value. The decision whether to develop a diversion- and theft-resistant fuel cycle depends on the perceived value of avoiding nuclear terrorism and proliferation

  1. Nuclear Energy, Nuclear Weapons Proliferation, and the Arms Race.

    Science.gov (United States)

    Hollander, Jack, Ed.

    A symposium was organized to reexamine the realities of vertical proliferation between the United States and the Soviet Union and to place into perspective the horizontal proliferation of nuclear weapons throughout the world, including the possible role of commercial nuclear power in facilitating proliferation. The four invited symposium…

  2. Oxidative stress induced pulmonary endothelial cell proliferation is ...

    African Journals Online (AJOL)

    Cellular hyper-proliferation, endothelial dysfunction and oxidative stress are hallmarks of the pathobiology of pulmonary hypertension. Indeed, pulmonary endothelial cells proliferation is susceptible to redox state modulation. Some studies suggest that superoxide stimulates endothelial cell proliferation while others have ...

  3. Combating the Proliferation of Weapons of Mass Destruction.

    Science.gov (United States)

    Jenkins, Bonnie

    1997-01-01

    Reveals the growing threat posed to all countries by the proliferation of weapons of mass destruction. Discusses the international effort combating this proliferation including the Nuclear Non-Proliferation Treaty, Strategic Arms Reduction Treaties, Biological Weapons Convention, and Chemical Weapons Convention. Also considers regional arms…

  4. Expression of a novel non-coding mitochondrial RNA in human proliferating cells

    Science.gov (United States)

    Villegas, Jaime; Burzio, Veronica; Villota, Claudio; Landerer, Eduardo; Martinez, Ronny; Santander, Marcela; Martinez, Rodrigo; Pinto, Rodrigo; Vera, María I.; Boccardo, Enrique; Villa, Luisa L.; Burzio, Luis O.

    2007-01-01

    Previously, we reported the presence in mouse cells of a mitochondrial RNA which contains an inverted repeat (IR) of 121 nucleotides (nt) covalently linked to the 5′ end of the mitochondrial 16S RNA (16S mtrRNA). Here, we report the structure of an equivalent transcript of 2374 nt which is over-expressed in human proliferating cells but not in resting cells. The transcript contains a hairpin structure comprising an IR of 815 nt linked to the 5′ end of the 16S mtrRNA and forming a long double-stranded structure or stem and a loop of 40 nt. The stem is resistant to RNase A and can be detected and isolated after digestion with the enzyme. This novel transcript is a non-coding RNA (ncRNA) and several evidences suggest that the transcript is synthesized in mitochondria. The expression of this transcript can be induced in resting lymphocytes stimulated with phytohaemagglutinin (PHA). Moreover, aphidicolin treatment of DU145 cells reversibly blocks proliferation and expression of the transcript. If the drug is removed, the cells re-assume proliferation and over-express the ncmtRNA. These results suggest that the expression of the ncmtRNA correlates with the replicative state of the cell and it may play a role in cell proliferation. PMID:17962305

  5. Effect of mechanical and electrical behavior of gelatin hydrogels on drug release and cell proliferation.

    Science.gov (United States)

    Biswal, Dibyajyoti; Anupriya, B; Uvanesh, K; Anis, Arfat; Banerjee, Indranil; Pal, Kunal

    2016-01-01

    The present study was aimed to explore the effect of the mech