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Sample records for bwr proliferation resistance

  1. Proliferation resistance fuel cycle technology

    International Nuclear Information System (INIS)

    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)

  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. To judge the degree of proliferation resistance

    International Nuclear Information System (INIS)

    This paper identified the Co-Chairman's proposals for assessing the degree of proliferation resistance. It identifies nine assessment factors which have been suggested in INFCE WG4 discussions and in the open literature as appropriate. These nine assessment factors are then considered in respect to their relative importance to the question of proliferation resistance. The paper is supported by an Appendix which gives selected extracts from the literature where the various assessment factors are discussed

  5. Proliferation resistance of small modular reactors fuels

    International Nuclear Information System (INIS)

    In this paper the proliferation resistance of different types of Small Modular Reactors (SMRs) has been examined and classified with criteria available in the literature. In the first part of the study, the level of proliferation attractiveness of traditional low-enriched UO2 and MOX fuels to be used in SMRs based on pressurized water technology has been analyzed. On the basis of numerical simulations both cores show significant proliferation risks. Although the MOX core is less proliferation prone in comparison to the UO2 core, it still can be highly attractive for diversion or undeclared production of nuclear material. In the second part of the paper, calculations to assess the proliferation attractiveness of fuel in typical small sodium cooled fast reactor show that proliferation risks from spent fuel cannot be neglected. The core contains a highly attractive plutonium composition during the whole life cycle. Despite some aspects of the design like the sealed core that enables easy detection of unauthorized withdrawal of fissile material and enhances proliferation resistance, in case of open Non-Proliferation Treaty break-out, weapon-grade plutonium in sufficient quantities could be extracted from the reactor core

  6. Proliferation resistance assessment of thermal recycle systems

    International Nuclear Information System (INIS)

    This paper examines the major proliferation aspects of thermal recycle systems and the extent to which technical or institutional measures could increase the difficulty or detectability of misuse of the system by would-be proliferators. It does this by examining the various activities necessary to acquire weapons-usable material using a series of assessment factors; resources required, time required, detectability. It is concluded that resistance to proliferation could be improved substantially by collecting reprocessing, conversion and fuel fabrication plants under multi national control and instituting new measures to protect fresh MOX fuel. Resistance to theft at sub-national level could be improved by co-location of sensitive facilities high levels of physical protection at plants and during transportation and possibly by adding a radiation barrier to MOX prior to shipment

  7. Proliferation resistance features in nuclear reactor designs

    International Nuclear Information System (INIS)

    Full text: The presentation gives an overview of the fundamental principles of non-proliferation of nuclear materials and technologies in the process of designing the nuclear reactors. The nuclear power engineering includes the activities involving the risk of proliferation of nuclear weapons (such as separation of uranium isotopes (enrichment), long-term storage of irradiated fuel, reprocessing of irradiated fuel by means of separation of plutonium and/or uranium wherefrom, storage of separated fissile materials). Proliferation resistance can be defined as the characteristic of a given nuclear power system which would prevent change-over or unauthorized production and use of the nuclear materials or technologies intended to possession of nuclear weapons or other nuclear explosives. The basic principles of non-proliferation as formulated in the frame of IAEA-sponsored international project INPRO have been analyzed for their relevance in designing the innovative nuclear power systems based on lead-cooled fast reactors. (author)

  8. Proliferation resistance features in nuclear reactor designs

    International Nuclear Information System (INIS)

    The paper presents a review of the main principles for technologies and materials protection from unauthorized proliferation and application to be considered in nuclear reactors designing. Nuclear power features certain operations sensitive to nuclear weapons proliferation (such as separation of uranium isotopes (enrichment), long storage of spent fuel, processing of spent fuel, plutonium and/or uranium recovery from spent fuel, storage of recovered fissile materials). Proliferation resistance is defined as a nuclear energy system characteristic that impedes the diversion or undeclared production of nuclear material, or misuse of technology with the purpose of acquiring nuclear weapons or other nuclear explosive devices. The basic principles of non-proliferation established in the INPRO international project sponsored by IAEA have been discussed as implemented for designing of the innovative nuclear energy systems based on fast lead-cooled nuclear reactors

  9. Intergranular corrosion resistance of 304 stainless steel welded junctions for BWR piping

    International Nuclear Information System (INIS)

    Intergranular corrosion of stainless steel pipings has been noticed in welded areas in almost every boiling water nuclear power plants so as to attract the attention of safety authorities. The research activity developed during more than one decade allowed to apply and qualify a series of provisions aiming at assuring the tensio-corrosion resistence and satisfying the strigent limitations imposed by control authorities. Results achieved by ENEL in cooperation with CISE concerning the study of origin materials, made on mockups of 4'' to 8'' welded tubings and weldings of BWR plants

  10. Proliferation resistance criteria for fissile material disposition

    International Nuclear Information System (INIS)

    The 1994 National Academy of Sciences study open-quotes Management and Disposition of Excess Weapons Plutoniumclose quotes defined options for reducing the national and international proliferation risks of materials declared excess to the nuclear weapons program. This report proposes criteria for assessing the proliferation resistance of these options. The criteria are general, encompassing all stages of the disposition process from storage through intermediate processing to final disposition including the facilities, processing technologies and materials, the level of safeguards for these materials, and the national/subnational threat to the materials

  11. Proliferation resistance assessment of nuclear systems

    International Nuclear Information System (INIS)

    The paper focuses on examining the degree to which nuclear systems could be used to acquire nuclear weapons material. It establishes a framework for proliferation resistance assessment and illustrates its applicability through an analysis of reference systems for once-through cycles, breeder cycles and thermal recycle. On a more tentative basis, the approach is applied to various alternative technical and institutional measures. This paper was also submitted to Working Groups 5 and 8

  12. 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.

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

    International Nuclear Information System (INIS)

    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

  14. Proliferation resistance assessment of nuclear systems

    International Nuclear Information System (INIS)

    The first part of the present paper describes the basic assessment procedure that is adopted in the analysis of the three generic nuclear systems. Once-through, fast breeder, and thermal recycle systems are then treated in Sections II, III, and IV, respectively. In each of these sections, a reference system is examined, possible technical and institutional improvements are considered, and alternative system types are indicated. Section V then discusses the relative proliferation resistance of the three generic systems. Although this paper emphasizes the analysis and comparison of individual fuel cycle alternatives, Section V indicates briefly how these analyses then have to be considered in a broader context where systems coexist

  15. Study on proliferation time and response time for proliferation resistance evaluation

    International Nuclear Information System (INIS)

    'Proliferation time' is one of the proliferation resistance measures adopted by the Generation IV Nuclear Energy Systems International Forum (GIF) Proliferation Resistance and Physical Protection (PR and PP) evaluation methodology. A longer proliferation time would provide the international society with more time to intervene politically in order to dissuade the State from completing its nuclear weapons program. A longer proliferation time would therefore contribute to the enhancement of proliferation resistance of a given nuclear energy system. Two methods are considered for judging whether the proliferation time is long enough: 1) comparison of the proliferation times between a reference nuclear energy system and the subject system, and 2) comparison between the proliferation time and the response time, which can be defined as the time available to the international society to make a political intervention. This paper focuses on the latter method and examines how the response time can be estimated by reviewing prior incidents. (author)

  16. Can light water reactors be proliferation resistant?

    International Nuclear Information System (INIS)

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

  17. Assessing the proliferation resistance of innovative nuclear fuel cycles

    International Nuclear Information System (INIS)

    Full text: The National Nuclear Security Administration (NNSA) is developing methods for nonproliferation assessments to support the development and implementation of U.S. nonproliferation policy. A Nonproliferation Assessment Methodology (NPAM) Working Group, comprised of representatives from the DOE laboratories and academia, was established to prepare guidelines for the selection of methods and for the performance of nonproliferation assessments. The guidelines address the full scope of proliferation issues that must be addressed by NNSA in support of the development of nonproliferation policy. The guidelines were completed in November 2002 and submitted for peer review. Proliferation is defined as the 'acquisition of one or more nuclear weapons by a nation or subnational group that currently does not have them'. Nonproliferation assessments generally attempt to measure the proliferation resistance of a particular alternative or the proliferation risk of a certain action or proposition. It is important to distinguish between the two types of assessment because they rely on different measures. Proliferation Resistance - Degree of difficulty that a nuclear material, facility, process, or activity poses to the acquisition of one or more nuclear weapons). Proliferation Risk - The likelihood of a nation or subnational group acquiring one or more nuclear weapons within a given time period. Proliferation resistance is an attribute of a nuclear system (a commercial fuel cycle, a facility, transportation of nuclear material, etc.). Proliferation risk, on the other hand, can apply to actions or activities not necessarily part of a physical nuclear system. Acquisition of specific technologies or skills, industrial capabilities, etc., can bear on the risk of proliferation. In comparing the proliferation characteristics of innovative nuclear fuel cycle systems, proliferation resistance is a more appropriate measure of performance than proliferation risk because of the focus

  18. Proliferation resistance of plutonium based on decay heat

    International Nuclear Information System (INIS)

    Proliferation resistance of plutonium can be enhanced by increasing the decay heat of plutonium. For example, it can be enhanced by increasing the isotopic fraction of 238Pu, which has the largest decay heat among plutonium isotopes, produced by transmutation of Minor Actinides (Protected Plutonium Production: P3). In the present paper, proliferation resistance of plutonium was evaluated based on decay heat with physical assessment model. As a summary of the evaluation, new criteria to evaluate proliferation resistance of plutonium based on its isotopic composition from the view point of decay heat were suggested. The present methodology and the criteria were applied to evaluate the impact of P3 by the transmutation of Minor Actinides in fast breeder reactor blanket on proliferation resistance of plutonium. (author)

  19. Model for nuclear proliferation resistance analysis using decision making tools

    International Nuclear Information System (INIS)

    The nuclear proliferation risks of nuclear fuel cycles is being considered as one of the most important factors in assessing advanced and innovative nuclear systems in GEN IV and INPRO program. They have been trying to find out an appropriate and reasonable method to evaluate quantitatively several nuclear energy system alternatives. Any reasonable methodology for integrated analysis of the proliferation resistance, however, has not yet been come out at this time. In this study, several decision making methods, which have been used in the situation of multiple objectives, are described in order to see if those can be appropriately used for proliferation resistance evaluation. Especially, the AHP model for quantitatively evaluating proliferation resistance is dealt with in more detail. The theoretical principle of the method and some examples for the proliferation resistance problem are described. For more efficient applications, a simple computer program for the AHP model is developed, and the usage of the program is introduced here in detail. We hope that the program developed in this study could be useful for quantitative analysis of the proliferation resistance involving multiple conflict criteria

  20. Proliferation resistance criteria for fissile material disposition issues

    International Nuclear Information System (INIS)

    The 1994 National Acdaemy of Sciences study ''Management and Disposition of Excess Weapons Plutonium'' defined options for reducing the national and international proliferation risks of materials declared excess to the nuclear weapons program. This paper proposes criteria for assessing the proliferation resistance of these options as well defining the ''Standards'' from the report. The criteria are general, encompassing all stages of the disposition process from storage through intermediate processing to final disposition including the facilities, processing technologies and materials, the level of safeguards for these materials, and the national/subnational threat to the materials

  1. Proliferation resistance and safeguardability of innovative nuclear fuel cycles

    International Nuclear Information System (INIS)

    Full text: The nuclear non-proliferation regime can be strengthened by the introduction of proliferation-resistant features at relevant stages of the fuel cycle that would serve as intrinsic technical barriers to proliferation. These technical barriers can be used to reinforce the existing institutional barriers to proliferation such as treaty regimes and associated verification arrangements. This has not been a priority to date, because institutional efforts at containing the spread of sensitive technology have been largely effective, and because civil nuclear .programs involve very little weapons-grade material. However, the increasing use of plutonium fuels is prompting renewed interest in technical approaches in support of non-proliferation objectives. While there is no such thing as a fully proliferation-resistant fuel cycle, numerous concepts have been put forward by experts of various States with the aim of developing a nuclear fuel cycle with enhanced resistance to proliferation. At the very least, technical barriers make breakout from the non- proliferation regime more difficult and time-consuming, thus providing enhanced deterrence and better opportunities for the international community to intervene. The manufacture of nuclear weapons requires either pure uranium metal at very high enrichment levels or pure plutonium metal preferably with a very high proportion of Pu-239. Historically such material has been produced in facilities designed and operated for this particular purpose. Weapons-usable materials are very different to those normally produced in civil programs. The strategic value of the materials involved in the civil fuel cycles can be reduced further by increases in the proliferation barriers associated with the isotopic composition and chemical form of the material and also with the radiation hazard associated with the material. These barriers affect the complexity of materials handling at each step in the fuel cycle, the attractiveness of the

  2. Enhancing VVER annular proliferation resistance fuel with minor actinides

    International Nuclear Information System (INIS)

    Key aspects of the Global Nuclear Energy Partnership (GNEP) are to significantly advance the science and technology of nuclear energy systems and the Advanced Fuel Cycle (AFC) program. It consists of both innovative nuclear reactors and innovative research in separation and transmutation. To accomplish these goals, international cooperation is very important and public acceptance is crucial. The merits of nuclear energy are high-density energy, with low environmental impacts (i.e. almost zero greenhouse gas emission). Planned efforts involve near term and intermediate-term improvements in fuel utilization and recycling in current light water reactors (LWRs) as well as the longer-term development of new nuclear energy systems that offer much improved fuel utilization and proliferation resistance, along with continued advances in operational safety. The challenges are solving the energy needs of the world, protection against nuclear proliferation, the problem of nuclear waste, and the global environmental problem. 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 LWR (Gen-III Plus), which not only can reduce the spent fuel for storage, but also increase the 238Pu and 240Pu 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, the minor actinides (MA) are viewed more as a resource to be recycled, or transmuted to less hazardous and possibly more useful forms, rather than simply as a waste stream to be disposed of in expensive repository facilities. 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

  3. 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)

  4. Analysis of nuclear proliferation resistance reprocessing and recycling technologies

    International Nuclear Information System (INIS)

    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 a decrease in proliferation risks. On the other hand, at this moment, advanced technologies with reduced proliferation risks are being developed. 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 COEXTM 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 U.S., 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 and D and robust flow-sheets. Finally, future generation recycling schemes will likely 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 have less proliferation risk than the commercial PUREX process. In this paper, we review the actual PUREX process along with the advanced recycling technologies that will

  5. SIGMA: The first intrinsic proliferation resistant uranium enrichment technology

    International Nuclear Information System (INIS)

    Full text: The problem of proliferation using originally peaceful nuclear systems is a challenge that the nuclear energy must solve, if it is wanted that the nuclear energy could play a major role in the energy basket for all the planet that want to peacefully use it. Technology restrictions that finally reflects resources transfers, or competitiveness advantages obtained by countries that will have both, peaceful and military use, could become a major obstacle if its really wanted that in the future, nuclear energy could be a massive option like modern gas turbines or even piston motors. To deal with this challenge, different international and national efforts (INPRO, GEN IV, etc.) proposed to open the discussion about new 'proliferation resistant' approaches, in order to help a sustainable large scale use of nuclear energy in the future. Now is clear that proliferation resistance is a feature that is useful only if its applied to all the nuclear system itself, and includes the nuclear power plant together with all fuel cycle facilities. At present there are many proliferation resistance concept proposed for future advanced nuclear systems, new reactors and back end fuel cycles facilities, but up to now there are no new proposals for present front end of nuclear fuel cycle. In general, nuclear material present in the front end of LWR fuel cycle has been taken as very low attractiveness material because it must be enriched significantly to be usable in a weapon. But there are strong concerns about that plants designed to enrich LEU could be easily modified to produce HEU. As was stated previously, it was a well-known issue that ultracentrifuge could be easily used to potentially produce, in a short time, significant quantities of direct usable weapon grade fissile material. All these issues could be seen as minor problem if the final proliferation risk could be addressed by institutional safeguards without significant cost. But leveled fuel cycle safeguard cost

  6. Modeling and evaluating proliferation resistance of nuclear energy systems for strategy switching proliferation

    International Nuclear Information System (INIS)

    Highlights: ► Sensitivity analysis is carried out for the model and physical input parameters. ► Interphase drag has minor effect on the dryout heat flux (DHF) in 1D configuration. ► Model calibration on pressure drop experiments fails to improve prediction of DHF. ► Calibrated classical model provides the best agreement with DHF data from 1D tests. ► Further validation of drag models requires data from 2D and 3D experiments on DHF. - Abstract: This paper reports a Markov model based approach to systematically evaluating the proliferation resistance (PR) of nuclear energy systems (NESs). The focus of the study is on the development of the Markov models for a class of complex PR scenarios, i.e., mixed covert/overt strategy switching proliferation, for NESs with two modes of material flow, batch and continuous. In particular, a set of diversion and/or breakout scenarios and covert/overt misuse scenarios are studied in detail for an Example Sodium Fast Reactor (ESFR) system. Both probabilistic and deterministic PR measures are calculated using a software tool that implements the proposed approach and can be used to quantitatively compare proliferation resistant characteristics of different scenarios for a given NES, according to the computed PR measures

  7. Assessment of proliferation resistance of thermal recycle systems

    International Nuclear Information System (INIS)

    An assessment is made of the proliferation resistance of thermal recycle systems. The safeguards aspects are not addressed. Three routes to the acquisition of materials for nuclear weapons are addressed namely; a deliberate political decision by a government involving the use of dedicated facilities, a deliberate political decision by government involving abuse of nuclear fuel cycle facilities and theft by a subnational group. The most sensitive parts of the reference fuel cycle and the alternative technical measures are examined to judge their relative sensitivity. This is done by examining the difference forms in which plutonium can exist in the fuel cycle. The role which different institutional arrangements can play is also evaluated. From this comparative assessment it is concluded that, taking into account the qualitative nature of the assessment, the different stages of development of the various fuel cycles, the various realizations possible in respect of the deployment of facilities within individual countries and the evolutionary nature of the technical and institutional improvements foreseeable no fuel cycle can be made completely free from abuse. Furthermore it appears that following progressive introduction of features that will improve proliferation resistance there will not be significant differences between the various fuel cycles when compared at the point in time when they are introduced into widespread use. Provided such features are developed and implemented there is no reason on proliferation grounds to prefer one cycle to another

  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. Teleconference highlights-NE-NA proliferation resistance review

    International Nuclear Information System (INIS)

    Herczeg gave a readout from the kickoff meeting with Paul Lisowski - namely develop a common definition of proliferation resistance (for use by S-1, other upper management, public affairs, etc.), and to evaluate possible framework where a metric could be assigned for fuel cycle comparisons (integral, easy to communicate). Sprinkle raised concern about 'trivializing' notion of proliferation resistance (PR), with idea of making sure we don't lose the concept that strong safeguards and security are required within a nonproliferation framework that support U.S. policy goals. Integrated Safeguards by Design notion was brought up in this context. Round table discussion of the term PR, its misuse (even unintentional), fact that Chu is using term and apparently in context of proliferation proof. It was noted that there has been much work already done in this area and we should not reinvent the wheel. One of the first tasks needs to be gathering up old reports (TOPS, Como, PRPP, etc) and distributing to group (action item for all). It was also noted that there are multiple definitions of PR, including the recent NPIA, supporting the need for this type of activity. Miller described the current work package under AFCI, with $50k of funding from the campaign management account. Herczeg asked about additional funds should it become clear that a larger effort is required (tension between current program and getting something out relatively soon). Goldner to look into potential additional funds. Miller notes that within current work package, easy to engage LANL participants and that Per Peterson can participate under UCB funding (a new center is being established with UC fee awards from LANL and LLNL - the Berkeley Nuclear Research Center). Consensus that Per would be a good external member of the group. Sprinkle notes that held like to coordinate the NE and NA work packages. Miller and Sprinkle to work offline. Wallace talked about the possibility of being more quantitative in

  10. Partitioning and transmutation - Technical feasibility, proliferation resistance and safeguardability

    International Nuclear Information System (INIS)

    combined with a transmutation cycle (second stratum). Here, the first separation of radiotoxic elements from the PUREX high-level liquid waste could be achieved by advanced aqueous partitioning. In the following transmutation cycle, pyroreprocessing should be used, because of a number of advantages; those are a higher compactness of equipment and the possibility to form an integrated system between irradiation and reprocessing facility (reduced transport of nuclear materials and process costs in general) higher radiation stability of the salt in the pyrochemical process compared to the organic solvent in the hydrochemical process offers an important advantage when dealing with highly active spent MA fuel compared with aqueous methods, dry reprocessing results in less pure and thus more proliferation resistant fractions of Pu, Np and Am. In particular the latter aspect is important in view of the attractiveness of products for proliferation. In the paper the different partitioning processes, aqueous and dry, will be briefly described and analyzed for their strengths and weaknesses in view of safeguards and proliferation. Furthermore, the advantages and drawbacks of homogeneous and heterogeneous cycles will be discussed in view of proliferation resistance and safeguardability. (author)

  11. Final report on LDRD project "proliferation-resistant fuel cycles"

    Energy Technology Data Exchange (ETDEWEB)

    Brown, N W; Hassberger, J A

    1999-02-25

    This report provides a summary of LDRD work completed during 1997 and 1998 to develop the ideas and concepts that lead to the Secure, Transportable, Autonomous Reactor (STAR) program proposals to the DOE Nuclear Energy Research Initiative (NERI). The STAR program consists of a team of three national laboratories (LLNL, ANL, and LANL), three universities, (UC Berkeley, TAMU, and MIT) and the Westinghouse Research Center. Based on the LLNL work and their own efforts on related work this team prepared and integrated a package of twelve proposals that will carry the LDRD work outlined here into the next phase of development. We are proposing to develop a new nuclear system that meets stringent requirements for a high degree of safety and proliferation resistance, and also deals directly with the related nuclear waste and spent fuel management issues.

  12. To judge the degree of proliferation resistance (revised)

    International Nuclear Information System (INIS)

    After an introduction which discusses the problems of undertaking an assessment of proliferation resistance and reproduces the advice provided by the INFCE Technical Co-Ordinating Committee, the remainder of the paper is divided into three sections. In the first section the same assessment factors as listed in INFCE/DEP./WG-4/104 are set out, with two additional factors, and regrouped along the lines of the discussions held in INFCE/WG4 30th November 1978. In the second section three factors are applied to the three main fuel cycles; the once-through fuel cycle, plutonium recycle in LWR reactors and the fast reactor fuel cycle. The third section suggests a first tentative judgement

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

    International Nuclear Information System (INIS)

    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

  14. Proliferation resistance assessments during the design phase of a recycling facility as a means of reducing proliferation risks

    Energy Technology Data Exchange (ETDEWEB)

    Lindell, M.A.; Grape, S.; Haekansson, A.; Jacobsson Svaerd, S. [Department of Physics and Astronomy, Uppsala University: Box 516, SE-75120 Uppsala (Sweden)

    2013-07-01

    The sustainability criterion for Gen IV nuclear energy systems inherently presumes the availability of efficient fuel recycling capabilities. One area for research on advanced fuel recycling concerns safeguards aspects of this type of facilities. Since a recycling facility may be considered as sensitive from a non-proliferation perspective, it is important to address these issues early in the design process, according to the principle of Safeguards By Design. Presented in this paper is a mode of procedure, where assessments of the proliferation resistance (PR) of a recycling facility for fast reactor fuel have been performed so as to identify the weakest barriers to proliferation of nuclear material. Two supplementing established methodologies have been applied; TOPS (Technological Opportunities to increase Proliferation resistance of nuclear power Systems) and PR-PP (Proliferation Resistance and Physical Protection evaluation methodology). The chosen fuel recycling facility belongs to a small Gen IV lead-cooled fast reactor system that is under study in Sweden. A schematic design of the recycling facility, where actinides are separated using solvent extraction, has been examined. The PR assessment methodologies make it possible to pinpoint areas in which the facility can be improved in order to reduce the risk of diversion. The initial facility design may then be slightly modified and/or safeguards measures may be introduced to reduce the total identified proliferation risk. After each modification of design and/or safeguards implementation, a new PR assessment of the revised system can then be carried out. This way, each modification can be evaluated and new ways to further enhance the proliferation resistance can be identified. This type of iterative procedure may support Safeguards By Design in the planning of new recycling plants and other nuclear facilities. (authors)

  15. Proliferation Resistant Fuel for Pebble Bed Modular Reactors

    International Nuclear Information System (INIS)

    Proliferation of nuclear weapons produced with power reactors plutonium has always been amajor problem of the nuclear energy industry. This includes the PebbleBed Modular Reactor(PBMR), which is a specific design of a GenIV High-Temperature Reactor (HTR), mainly due to its online refueling feature, which may be misused for the production of weapons gradeplutonium. A promising approach toward preventing the proliferation of power reactorplutoniumis to denaturate the plutonium by increasing the ratio of 238Pu to total Pu in the spentfuel(1). The 238Pu isotope is characterized by a high heat rate (approximately 567 W/kg) due to thealpha decay of the 238Pu with half-life of 87.74 yr, in addition to its high spontaneous fissionneutron emission, which is higher than that of 240Pu. Thus, the presence of 238Pu in Pu considerably complicates the design and construction of nuclear weapons based on Pu, owing tothese characteristics of 238Pu. Recent papers(2,3) show that a Pu mixture is proliferation resistant given that the weight ratio of 238Pu to Pu is larger than 6%. In this paper we have studied afeasible technique for ensuring that the 238Pu to Pu ratio, in the Pu produced in PBMR, is larger than 6% during the entire fuel cycle. Contamination of the spent fuel with 238Pu may be achieved by doping the nuclear fuel witheither 241Am or 237Np(4-13). The 238Pu isotope is obtained from both 241Am and 237Np by a neutron-capture reaction and the subsequent decay of the reaction products(13).The 237Np isotopeis by itself a potential weapons grade material. However, its large critical mass of 57±4 kg(14) andthe difficulty of extracting it from irradiated fuel elements make it impractical for weapons purposes. On the other hand, the critical mass of 241Am is smaller, i.e. 34 to 45 kg. However, withdecay heat production of 114W/kg, the critical mass becomes a heat source of 3.9 to 5.1 KW,which makes 241Am unsuitable for weapons applications(3). As a result, it is a non-proliferating

  16. Development of Proliferation Resistance Assessment Methodology based on International Standard

    International Nuclear Information System (INIS)

    Proliferation resistance is one of the requirement to be met in GEN IV and INPRO for next generation nuclear energy system. Internationally, the evaluation methodology on PR had been already initiated from 1980, but the systematic development was started at 2000s. In Korea, for the export of nuclear energy system and the increase of international credibility and transparence of domestic nuclear system and fuel cycle development, the independent development of PR evaluation methodology was started in 2007 as a nuclear long term R and D project and the development is being performed for the model of PR evaluation methodology. In 1st year, comparative study of GEN-IV/INPRO, PR indicator development, quantification of indicator and evaluation model development, analysis of technology system and international technology development trend had been performed. In 2nd and 3rd year, feasibility study of indicator, allowable limit of indicator, review of technical requirement of indicator, technical standard, design of evaluation model were done. The results of PR evaluation must be applied in the beginning of conceptual design of nuclear system. Through the technology development of PR evaluation methodology, the methodology will be applied in the regulatory requirement for authorization and permission to be developed

  17. 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.

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

    International Nuclear Information System (INIS)

    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

  19. A Study on the Regulatory Requirements of Nuclear Energy Systems in the Area of Proliferation Resistance

    International Nuclear Information System (INIS)

    The study indicates that reasonable guidelines can be developed based on the concepts, principles and fundamentals of proliferation resistance. The regulatory body is responsible for drafting and establishing regulatory requirements for the licensing process of nuclear energy systems, in line with State's commitments, obligations and policies regarding non-proliferation. The requirements would include enforcement ordinance, enforcement regulations, including technical codes and standards for design, operation, and maintenance, in the area of proliferation resistance of an NES. KAERI has been developing potential regulatory requirements of nuclear energy systems in the area of proliferation resistance based on the INPRO methodology. This paper presents general concepts and fundamentals, including relevant issues, of proliferation resistance that are to be considered in the licensing process of nuclear energy systems

  20. An Introduction to The Interdisciplinary Concept of Risk-Informed Proliferation Resistance

    International Nuclear Information System (INIS)

    The nonproliferation community is rich in diverse attempts aimed at predicting and preventing attempts at nuclear proliferation. Such efforts, however, are rarely incorporated into a holistic approach well-suited to solving both existing and emerging proliferation dilemmas. This division is particularly apparent with respect to the partition that separates the socio-political and technical approaches to solving such problems, approaches which are very often developed and utilized in isolation. Complicating matters further, are the diverse positions taken by various entities as they relate to the secure implementation of nuclear energy world-wide. Such positions range from the obdurate belief in the supposed inherent proliferation resistance of traditional spent fuel, to those regarding all nuclear energy systems as inherently proliferation prone, given their core reliance on the sensitive technologies of enrichment and reprocessing. Accordingly, moderates have argued for a risk-informed approach to combat nuclear proliferation, combining institutional factors, such as IAEA safeguards, with innovative reactor and fuel designs to bring about an acceptable notion of proliferation resistance. To this end, methodologies have been developed, which seek to assess and score attained proliferation resistance. Most notably, these include, the Proliferation Resistance and Physical Protection (PR and PP) methodology prepared by the GEN-IV International Forum and the IAEA's International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO). Such approaches have greatly advanced the concept of proliferation resistance; however, they remain limited by their exclusive concentration on the technological determinants of proliferation and non-consideration of other, equally important, socio-political determinants. This limitation is significant as numerous incidents have illustrated the ability of atypical states to find alternate paths to proliferation, for example

  1. Neutronic simulation calculations to assess the proliferation resistance of nuclear technologies

    International Nuclear Information System (INIS)

    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.)

  2. A collaboration on development of requirements and guidelines for proliferation resistance of future nuclear system in the IAEA INPRO

    International Nuclear Information System (INIS)

    This study surveyed and analyzed the existing activities and international status concerning proliferation resistance of nuclear energy systems, reviewed the features of proliferation resistance, and derived the requirements of future innovative nuclear energy systems. In IAEA INPRO, guidance for the evaluation of innovative nuclear reactors and fuel cycles on proliferation resistance was finalized through collaboration of member countries including Korea in reviewing technological status and developing the methodology for evaluation of proliferation resistance. This report, first, describes the progress of INPRO and the participation status of Korea in the project, and briefly summarizes the report of phase IA of INPRO. Next, features of proliferation resistance of nuclear systems, collaboration in the GIF and the INPRO for development of requirements and guidelines for proliferation resistance, and the final result of guidance for the evaluation of proliferation resistance were described. Finally, this study proposed measures for participation of further progress of the INPRO

  3. A collaboration on development of requirements and guidelines for proliferation resistance of future nuclear system in the IAEA INPRO

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Keun Bae; Lee, Kwang Seok; Kim, Hyun Jun; Jeong, Ik; Yang, Myung Seung; Ko, Won Il

    2003-10-01

    This study surveyed and analyzed the existing activities and international status concerning proliferation resistance of nuclear energy systems, reviewed the features of proliferation resistance, and derived the requirements of future innovative nuclear energy systems. In IAEA INPRO, guidance for the evaluation of innovative nuclear reactors and fuel cycles on proliferation resistance was finalized through collaboration of member countries including Korea in reviewing technological status and developing the methodology for evaluation of proliferation resistance. This report, first, describes the progress of INPRO and the participation status of Korea in the project, and briefly summarizes the report of phase IA of INPRO. Next, features of proliferation resistance of nuclear systems, collaboration in the GIF and the INPRO for development of requirements and guidelines for proliferation resistance, and the final result of guidance for the evaluation of proliferation resistance were described. Finally, this study proposed measures for participation of further progress of the INPRO.

  4. BWR 90: The ABB advanced BWR design

    International Nuclear Information System (INIS)

    ABB has two evolutionary advanced fight water reactors available today - the BWR 90 boiling water reactor and the System 80+ pressurised water reactor. The BWR 90 is based on the design, construction, commissioning and operation of the BWR 75 plants. The operation experience of the six plants of this advanced design has been very good. The average annual energy availability is above 90%, and the total power generation costs have been low. In the development of BWR 90 specific changes were introduced to the reference design, to adapt to technological progress, new safety requirements and to achieve cost savings. The thermal power rating of BWR 90 is 3800 MWth (providing a nominal 1374 MWe net), slightly higher dim that of the reference plant ABB Atom has taken advantage of margins gained using a new generation of its SVEA fuel to attain this power rating without major design modifications. The BWR 90 design was completed and offered to the TVO utility in Finland in 1991, as one of the contenders for the fifth Finnish nuclear power plant project. Thus, the design is available today for deployment in new plant projects. Utility views were incorporated through co-operation with the Finnish utility TVO, owner and operator of the two Olkiluoto plants of BWR 75 design. A review against the European Utility Requirement (EUR) set of requirements has been performed, since the design, in 1997, was selected by the EUR Steering Committee to be the first BWR to be evaluated against the EUR documents. The work is scheduled for completion in 1998. It will be the subject of an 'EUR Volume 3 Subset for BWR 90' document. ABB is continuing its BWR development work with the 'evolutionary' design BWR 90+. The primary design goal is to develop the BWR as a competitive option for the anticipated revival of the market for new nuclear plants beyond the turn of the century, as well as feeding ideas and inputs to the continuous modernisation efforts at operating plants. The development is

  5. A Study on the Improvement of the INPRO Proliferation Resistance Assessment Methodology

    International Nuclear Information System (INIS)

    Within the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO), a methodology for evaluating proliferation resistance (INPRO PR methodology) has been developed. However, User Requirement (UR) 4 regarding multiplicity and robustness of barriers against proliferation ('innovative nuclear energy systems should incorporate multiple proliferation resistance features and measures') remains to be developed. Because the development of a methodology for evaluating User Requirement 4 requires an acquisition/diversion pathway analysis, a systematic approach was developed for the identification and analysis of pathways for the acquisition of weapons-useable nuclear material. This approach was applied to the DUPIC fuel cycle which identified several proliferation target materials and plausible acquisition/diversion pathways. Based on these results, proliferation strategies that a proliferant State could adopt for undeclared removal of nuclear material from the DUPIC fuel cycle have been developed based on the objectives of the proliferation of the State, the quality and quantity of the target material, the time required to acquire the material for the proliferation, and the technical and financial capabilities of the potential proliferant State. The diversion pathway for fresh DUPIC fuel was analyzed using the INPRO User Requirements 1, 2 and 3, and based on these results an assessment procedure and metrics for evaluating the multiplicity and robustness of proliferation barriers has been developed. In conclusion, the multiplicity and robustness of proliferation barriers is not a function of the number of barriers, or of their individual characteristics but is an integrated function of the whole. The robustness of proliferation barriers is measured by determining whether the safeguards goals can be met. The harmonization of INPRO PR methodology with the GIF PR and PP methodology was also considered. It was suggested that, as also confirmed by IAEA

  6. Multi-attribute analysis of nuclear fuel cycle resistance to nuclear weapons proliferation

    International Nuclear Information System (INIS)

    Calculation study has been carried out to analyze the proliferation resistance of different scenarios of nuclear fuel cycle organization. Scenarios of stable and developing nuclear power were considered with involvement of thermal and fast reactors. The attention was paid mainly to the cycle with extended plutonium breeding on the basis of fast reactor technology, and to the schemes of fuel cycle organization allowing to minimize the proliferation risk

  7. 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?

  8. Proliferation resistance of advanced sustainable nuclear fuel cycles

    International Nuclear Information System (INIS)

    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)

  9. INPRO Collaborative Project: Proliferation Resistance: Acquisition/Diversion Pathway Analysis (PRADA)

    International Nuclear Information System (INIS)

    This publication contributes to strengthening the assessment area of proliferation resistance of the INPRO methodology. The basic principle for this area requires that multiple intrinsic features and extrinsic measures of proliferation resistance be implemented throughout the full life cycle of an innovative nuclear energy system to help ensure that the system will continue to be an unattractive means of acquiring fissile material for a nuclear weapons programme. A typical intrinsic feature is the dilution of plutonium with fission products as found in irradiated material, and a typical extrinsic measure is the placing of nuclear material under international safeguards.

  10. 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...

  11. Improving the proliferation resistance of research and test reactors

    International Nuclear Information System (INIS)

    Elimination, or substantial reduction, of the trade in highly enriched fuel elements for research and test reactors would significantly reduce the proliferation risk associated with the current potential for diversion of these materials. To this end, it is the long-term goal of U.S. policy to fuel all new and existing research and test reactors with uranium of less than 20% enrichment (but substantially greater than natural) excepting, perhaps, only a small number of high-power, high-performance, reactors. The U.S. development program for enrichment reduction in research and test reactor designs currently using 90-93% enriched uranium is based on the practical criterion that enrichment reduction should not cause significant flux performance (flux per unit power) or burnup performance degradation relative to the unmodified reactor design. A program is now beginning in the U.S. to develop the necessary fuel technology, but several years of work will be needed. Accordingly, as an immediate interim step, the U.S. is proposing to convert existing research and test reactors (and new designs) from the use of 90-93% enriched fuel to the use of 30-45% enriched fuel wherever this can be done without unacceptable reactor performance degradation

  12. The encapsulated nuclear heat source reactor for proliferation resistant nuclear energy

    International Nuclear Information System (INIS)

    The ENHS offers a unique combination of technological barriers and material barriers to nuclear proliferation. These barriers along with adequate institutional agreements could make the ENHS nuclear energy system extremely proliferation resistant. The technical barriers are limited access to fuel, limited access to neutrons and elimination of the host-country need for construction of sensitive facilities. The material barriers include at least 26 w/o of even Pu isotopes, buildup of significant concentration of 238Pu and MA, no increase in overall Pu inventory, and possibility to imbed a very intense radiation barrier in the core without interfering with shipment and installation. An energy system based on the ENHS that uses spent fuel from LWR as the feed and multi-recycles the fuel in a proliferation-resistant manner is proposed for waste reduction and for improved uranium resource utilization. (author)

  13. Improving the proliferation resistance of research and test reactors

    International Nuclear Information System (INIS)

    Elimination, or substantial reduction, of the trade in unirradiated highly-enriched fuel elements for research and test reactors would significantly reduce the proliferation risk associated with the current potential for diversion of these materials. To this end, it is the long-term goal of U.S. policy to fuel all new and existing research and test reactors with uranium of less-than-20% enrichment (but substantially greater than natural) excepting, perhaps, only a small number of high-power, high-performance, reactors. The U.S. development program for enrichment reduction in research and test reactor designs currently using 90-93% enriched uranium is based on the practical criterion that enrichment reduction should not cause significant flux performance (flux per unit power) or burnup performance degradation relative to the unmodified reactor design. To first order, this implies the requirement that the 235U loading in the reduced-enrichment fuel elements be the same as the 235U loading in the 90-93% enriched fuel elements. This can be accomplished by substitution of higher uranium density fuel technology for currently-used fuel technology in the fuel meat volume of the current fuel element design and/or by increasing the usable fuel meat volume. For research and test reactors of power greater than 5-10 megawatts, fuel technology does not currently exist that would permit enrichment reductions to below 20% utilizing this criterion. A program is now beginning in the U.S. to develop the necessary fuel technology. Currently-proven fuel technology is capable, however, of accommodating enrichment reductions to the 30-45% range (from 90-93%) for many reactors in the 5-50MW range. Accordingly the U.S. is proposing to convert existing reactors (and new designs) in the 5-50MW range from the use of highly-enriched fuel to the use of 30-45% enriched fuel, and reactors of less that about 5MW to less-than-20% enrichment, wherever this can be done without significant performance

  14. Sweat but no gain: inhibiting proliferation of multidrug resistant cancer cells with "ersatzdroges".

    Science.gov (United States)

    Kam, Yoonseok; Das, Tuhin; Tian, Haibin; Foroutan, Parastou; Ruiz, Epifanio; Martinez, Gary; Minton, Susan; Gillies, Robert J; Gatenby, Robert A

    2015-02-15

    ATP-binding cassette (ABC) drug transporters consuming ATPs for drug efflux is a common mechanism by which clinical cancers develop multidrug resistance (MDR). We hypothesized that MDR phenotypes could be suppressed by administration of "ersatzdroges," nonchemotherapy drugs that are, nevertheless, ABC substrates. We reasoned that, through prolonged activation of the ABC pumps, ersatzdroges will force MDR cells to divert limited resources from proliferation and invasion thus delaying disease progression. We evaluated ABC substrates as ersatzdroge by comparing their effects on proliferation and survival of MDR cell lines (MCF-7/Dox and 8226/Dox40) with the effects on the drug-sensitive parental lines (MCF-7 and 8226/s, respectively) in glucose-limited condition. The changes in glucose and energy demands were also examined in vitro and in vivo. MCF-7/Dox showed higher ATP demand and susceptibility to glucose resource limitation. Ersatzdroges significantly decreased proliferation of MCF-7/Dox when the culture media contained physiological glucose concentrations (1.0 g/L) or less, but had no effect on MCF-7. Similar evidence was obtained from 8226/Dox40 and 8226/s comparison. In vivo 18F-FDG-PET imaging demonstrated that glucose uptake was increased by systemic administration of an ersatzdroge in tumors composed of MDR. These results suggest that administration of ersatzdroges, by increasing the metabolic cost of resistance, can suppress proliferation of drug-resistance phenotypes. This provides a novel and relatively simple application model of evolution-based strategy, which can exploit the cost of resistance to delay proliferation of drug-resistant cancer phenotypes. Furthermore, suggested is the potential of ersatzdroges to identify tumors or regions of tumors that express the MDR phenotype. PMID:25156304

  15. Sweat but no gain: Inhibiting proliferation of multidrug resistant cancer cells with “Ersatzdroges”

    Science.gov (United States)

    Kam, Yoonseok; Das, Tuhin; Tian, Haibin; Foroutan, Parastou; Ruiz, Epifanio; Martinez, Gary; Minton, Susan; Gillies, Robert J.; Gatenby, Robert A.

    2014-01-01

    ATP-binding cassette (ABC) drug transporters consuming ATPs for drug efflux is a common mechanism by which clinical cancers develop multidrug resistance (MDR). We hypothesized that MDR phenotypes could be suppressed by administration of “ersatzdroges”, non-chemotherapy drugs that are, nevertheless, ABC substrates. We reasoned that, through prolonged activation of the ABC pumps, ersatzdroges will force MDR cells to divert limited resources from proliferation and invasion thus delaying disease progression. We evaluated ABC substrates as ersatzdroge by comparing their effects on proliferation and survival of MDR cell lines (MCF-7/Dox and 8226/Dox40) with the effects on the drug-sensitive parental lines (MCF-7 and 8226/s, respectively) in glucose-limited condition. The changes in glucose and energy demands were also examined in vitro and in vivo. MCF-7/Dox showed higher ATP demand and susceptibility to glucose resource limitation. Ersatzdroges significantly decreased proliferation of MCF-7/Dox when the culture media contained physiological glucose concentrations (1.0 g/L) or less, but had no effect on MCF-7. Similar evidence was obtained from 8226/Dox40 and 8226/s comparison. In vivo 18F-FDG-PET imaging demonstrated that glucose uptake was increased by systemic administration of an ersatzdroge in tumors composed of MDR. These results suggest that administration of ersatzdroges, by increasing the metabolic cost of resistance, can suppress proliferation of drug-resistance phenotypes. This provides a novel and relatively simple application model of evolution-based strategy which can exploit the cost of resistance to delay proliferation of drug-resistant cancer phenotypes. Furthermore, suggested is the potential of ersatzdroges to identify tumors or regions of tumors that express the MDR phenotype. PMID:25156304

  16. 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.

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

    International Nuclear Information System (INIS)

    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

  18. Proliferation Resistance and Safeguards by Design: The Safeguardability Assessment Tool Provided by the INPRO Collaborative Project ''INPRO'' (Proliferation Resistance and Safeguardability Assessment)

    International Nuclear Information System (INIS)

    Since the INPRO Collaborative Project on Proliferation Resistance and Safeguardability Assessment Tools (PROSA) was launched in 2011, Member State experts have worked with the INPRO Section and the IAEA Department of Safeguards to develop a revised methodology for self-assessment of sustainability in the area of proliferation resistance of a nuclear energy system (NES). With the common understanding that there is ''no proliferation resistance without safeguards'' the revised approach emphasizes the evaluation of a new 'User Requirement' for ''safeguardability'', that combines metrics of effective and efficient implementation of IAEA Safeguards including ''Safeguards-by-Design'' principles. The assessment with safeguardability as the key issue has been devised as a linear process evaluating the NES against a ''Basic Principle'' in the area of proliferation resistance, answering fundamental questions related to safeguards: 1) Do a State's legal commitments, policies and practices provide credible assurance of the exclusively peaceful use of the NES, including a legal basis for verification activities by the IAEA? 2) Does design and operation of the NES facilitate the effective and efficient implementation of IAEA safeguards? To answer those questions, a questionnaire approach has been developed that clearly identifies gaps and weaknesses. Gaps include prospects for improvements and needs for research and development. In this context, the PROSA approach assesses the safeguardability of a NES using a layered ''Evaluation Questionnaire'' that defines Evaluation Parameters (EP), EP-related questions, Illustrative Tests and Screening Questions to present and structure the evidence of findings. An integral part of the assessment process is Safeguards-by-Design, the identification of potential diversion, misuse and concealment strategies (coarse diversion path

  19. Proliferation Resistance and Material Type considerations within the Collaborative Project for a European Sodium Fast Reactor

    International Nuclear Information System (INIS)

    The collaborative project for a European Sodium Fast Reactor (CP‑ESFR) is an international project where 25 European partners developed Research & Development solutions and concepts for a European sodium fast reactor. The project was funded by the 7. European Union Framework Programme and covered topics such as the reactor architectures and components, the fuel, the fuel element and the fuel cycle, and the safety concepts. Within sub‑project 3, dedicated to safety, a task addressed proliferation resistance considerations. The Generation IV International Forum (GIF) Proliferation Resistance and Physical Protection (PR and PP) Evaluation Methodology has been selected as the general framework for this work, complemented by punctual aspects of the IAEA‑INPRO Proliferation Resistance methodology and other literature studies - in particular for material type characterization. The activity has been carried out taking the GIF PR and PP Evaluation Methodology and its Addendum as the general guideline for identifying potential nuclear material diversion targets. The targets proliferation attractiveness has been analyzed in terms of the suitability of the targets’ nuclear material as the basis for its use in nuclear explosives. To this aim the PR and PP Fissile Material Type measure was supplemented by other literature studies, whose related metrics have been applied to the nuclear material items present in the considered core alternatives. This paper will firstly summarize the main ESFR design aspects relevant for PR following the structure of the GIF PR and PP White Paper template. An analysis on proliferation targets is then discussed, with emphasis on their characterization from a nuclear material point of view. Finally, a high‑level ESFR PR analysis according to the four main proliferation strategies identified by the GIF PR and PP Evaluation Methodology (concealed diversion, concealed misuse, breakout, clandestine production in clandestine facilities) is

  20. BWR type reactor

    International Nuclear Information System (INIS)

    In a coolant circulation in BWR type reactors, since the mixed stream of steam fluid undergoes a great resistance, the pressure loss due to the flow rate distribution when the coolants flow from the upper plenum into the stand pipe is increased upon passing stand pipe. Also in the spontaneous recycling reactor, pressure loss is still left upon passing the swirling blade of a gas-liquid separator. In view of the above, a plurality of vertical members each having a lower end opened to a gas-liquid two phase boundary and an upper end directly suspended from a steam dryer to the gas-liquid separator. The liquid droplets from the 2-phase boundary heated in the reactor core and formed into a mixed gas-liquid 2-phase stream is directed in the vertical direction accompanied with the steam. The liquid droplets spontaneously fallen by gravity from greater ones successively and the droplets in the steam abutted against the vertical member are fallen as a liquid membrane. Thus, the gas-liquid separation is conducted, the dry steam is directly flown into the steam dryer, thereby capable of providing a gas-liquid separator having gas-liquid separation performance with lower loss than usual. (N.H.)

  1. Proliferation Resistance for Fast Reactors and Related Fuel Cycles: Issues and Impacts

    International Nuclear Information System (INIS)

    The prospects for a nuclear renaissance, or 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 (GenIV) International Forum (GIF) and the International 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 use 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? Will new safeguards technologies and approaches need to be developed? How can the efficiency and effectiveness of 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 play a role in reducing state threats and possibly significantly reducing non-state threats. There will be an especially important role for extrinsic factors, including the various measures-from safeguards to export controls-embodied in the

  2. 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

  3. 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

  4. A status of methodology developments in France for assessing proliferation resistance of nuclear energy systems

    International Nuclear Information System (INIS)

    This presentation will provide an update of methodological developments carried in France to assess proliferation resistance of nuclear systems. In a first part we will give an outlook on the SAPRA approach (SAPRA stands for 'Simplified Approach for Proliferation Resistance Assessment'), developed by AREVA in collaboration with EDF. SAPRA is aiming at providing a simple approach which may be easily implemented to assess and make use of provisions for enhancing Proliferation Resistance (PR) of nuclear systems (that is nuclear reactors and their associated fuel cycles). In the initiatives to develop innovative nuclear energy system (Generation-IV, INPRO), PR is one of the key elements, along with economics, safety, sustainability and environment which has to be addressed. Assessment of proliferation resistance is therefore of timely importance. This method uses the classical concept of barriers. Four categories of barriers are distinguished: material, technical, institutional and specific barriers for the weapon making phase. In the proliferation process (or route), four steps are considered: diversion of materials, transformation, transport and making of the nuclear weapon using either high enriched uranium or plutonium. All steps of the fuel cycle from uranium mining to final disposals of spent fuels or nuclear waste are examined. A scale of value is then defined in order to quantify each of these elementary steps and figures are aggregated to obtain global performance indices for PR and to identify weak points. Various nuclear systems have been analyzed and general conclusions drawn from these results will be presented. This approach offers a first developed framework to derive a practical and effective use of a proven tool to the needs and specificities of proliferation resistance assessment. In a second part, we will describe the current developments undertaken at the CEA to implement the 'multi-attribute' method proposed by the Texas A and M university (W

  5. Denaturing of plutonium by transmutation of minor-actinides for enhancement of proliferation resistance

    International Nuclear Information System (INIS)

    Feasibility study for the plutonium denaturing by utilizing minor-actinide transmutation in light water reactors has been performed. And the intrinsic feature of proliferation resistance of plutonium has been discussed based on IAEA's publication and Kessler's proposal. The analytical results show that not only 238Pu but also other plutonium isotopes with even-mass-number have very important role for denaturing of plutonium due to their relatively large critical mass and noticeably high spontaneous fission neutron generation. With the change of the minor-actinide doping ratio in U-Pu mix oxide fuel and moderator to fuel ratio, it is found that the reactor-grade plutonium from conventional light water reactors can be denatured to satisfy the proliferation resistance criterion based on the Kessler's proposal but not to be sufficient for the criterion based on IAEA's publication. It has been also confirmed that all the safety coefficients take negative value throughout the irradiation. (author)

  6. AIROX dry pyrochemical processing of oxide fuels: a proliferation-resistant reprocessing method

    International Nuclear Information System (INIS)

    Potential diversion of nuclear material from power production to weapons production by national or subnational groups has resulted in a reevaluation of the proliferation resistance of various fuel cycles. The low-contamination fuel cycle, utilizting AIROX dry processing, is proliferation resistant due to the retention of fission products with the fuel and to the low concentration of fissile material in all process steps. In the AIROX process, UO2 is oxidized with air to U3O8 to expand the fuel volume which simultaneously declads and pulverizes the fuel; the fuel is subsequently reenriched, repelletized, and recycled to the reactor. Fuel cycles utilizing this method of reprocessing will extend our uranium reserves, decrease the spent fuel storage requirements, and decrease the amount of waste requiring storage in a Federal Repository for environmental isolation. AIROX reprocessing is applicable to both light-water reactor fuel cycles as well as fast-breeder fuel cycles

  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. Methodology Development and Applications of Proliferation Resistance and Physical Protection Evaluation.

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-04-11

    We present an overview of the program on the evaluation methodology for proliferation resistance and physical protection (PR&PP) of advanced nuclear energy systems (NESs) sponsored by the Generation IV International Forum (GIF). For a proposed NES design, the methodology defines a set of challenges, analyzes system response to these challenges, and assesses outcomes. The challenges to the NES are the threats posed by potential actors (proliferant States or sub-national adversaries). The characteristics of Generation IV systems, both technical and institutional, are used to evaluate the response of the system and to determine its resistance against proliferation threats and robustness against sabotage and terrorism threats. The outcomes of the system response are expressed in terms of a set of measures, which are the high-level PR&PP characteristics of the NES. The methodology is organized to allow evaluations to be performed at the earliest stages of system design and to become more detailed and more representative as the design progresses. It can thus be used to enable a program in safeguards by design or to enhance the conceptual design process of an NES with regard to intrinsic features for PR&PP.

  9. MAAP BWR application guidelines

    International Nuclear Information System (INIS)

    The MAAP Thermal-Hydraulic Qualification and Application Project has as its objective to identify those thermal-hydraulic phenomena modeled in MAAP which are important in predicting severe accident sequences, to qualify those models and to provide guidelines for use of the code. This report provides user guidelines for use of the BWR version of MAAP. The report includes a discussion of the important features of the BWR that are modeled in MAAP, the MAAP modeling of phenomena important to predicting severe accidents and user guidelines for several accident sequences

  10. Proliferation resistance of a hypothetical sodium fast reactor under an assumed breakout scenario

    International Nuclear Information System (INIS)

    The Proliferation Resistance and Physical Protection (PR and PP) Working Group of the Generation IV International Forum (GIF) conducted a high-level pathway analysis of a hypothetical sodium fast reactor and integral fuel processing facility (called collectively the 'Example Sodium Fast Reactor' or ESFR), as a test of the effectiveness of its analysis methodology. From a common set of assumed host-state capabilities and objectives, a number of threat scenarios emerge (Concealed Diversion, Concealed Misuse, Breakout or Overt Misuse, and Theft/Sabotage). This paper presents the results of the analysis based on the Breakout scenario. A distinguishing aspect of Breakout scenario consideration concerns the optimal use of the time from breakout to weapons readiness, which is related to the Proliferation Time measure. The goal of analyzing the breakout scenario was therefore to complement other analyses involving the Concealed Misuse and Diversion scenarios by exploring the minimum post-breakout time to weapons readiness. Four target strategies were chosen for analysis: (1) Diversion of LEU feed material at front-end of the ESFR facility; (2) Misuse of the reactor facility to irradiate fertile material; (3) Misuse of the reactor facility to irradiate material in the in-core fuel storage basket; and (4) Misuse of the fuel processing facility to higher-purity TRU. The investigation identified several general 'sub-strategies' within the Breakout scenario, dependent upon the aggressiveness with which a State pursues its intent to break out (including its aversion to the risk of detection). The sub-strategy chosen by a proliferant state will affect both the time available and potential complexity for proliferation activities. The sub-strategy chosen is itself affected by political factors (foreign relations agenda of state, probability of external intervention after breakout, external dependence of proliferant state's supply chain, etc.) These factors, although of interest

  11. BWR internal cracking issues

    International Nuclear Information System (INIS)

    The regulatory issues associated with cracking of boiling water reactor (BWR) internals is being addressed by the Nuclear Regulatory Commission (NRC) staff and is the subject of a voluntary industry initiative. The lessons learned from this effort will be applied to pressurized water reactor (PWR) internals cracking issues

  12. Development of advanced BWR

    International Nuclear Information System (INIS)

    The Japanese technology and domestic production of BWR type nuclear power plants have been established through the experiences in the construction and operation of BWRs in addition to the technical agreement with the General Electric Co. In early days, the plants experienced some trouble such as stress corrosion cracking and some inconvenience in the operation and maintenance. The government, electric power companies and BWR manufacturers have endeavored to standardize and improve the design of LWRs for the purpose of improving the safety, reliability and the rate of operation and reducing the radiation exposure dose of plant workers. The first and second stages of the standardization and improvement of LWRs have been completed. Five manufacturers of BWRs in the world have continued the conceptual design of a new version of BWR power plants. It was concluded that this is the most desirable version of BWR nuclear power stations, but the technical and economic evaluation must be made before the commercial application. Six electric power companies and three manufacturers of BWRs in Japan set up the organization to develop the technology in cooperation. The internal pump system, the new control rod drive mechanism and others are the main features. (Kako, I.)

  13. Proliferation Resistance and Physical Protection (PR and PP) Evaluation Methodology: Objectives and Accomplishments

    International Nuclear Information System (INIS)

    This paper presents an overview of the objectives and accomplishments of the program on the evaluation methodology for proliferation resistance and physical protection (PR and PP) of advanced nuclear energy systems (NESs). The Generation IV Road-map recommended the development of an evaluation methodology to define measures for PR and PP and to develop a methodology for evaluating them for the six NESs proposed within the Gen IV program. Accordingly, the Generation IV International Forum (GIF) formed a Working Group in December 2002 to develop a methodology. The current version of the methodology (Revision 5) has been approved by GIF for open distribution and is available at the GIF web site (*). For a proposed NES design, the methodology defines a set of challenges, analyzes system response to these challenges, and assesses outcomes. The challenges to the NES are the threats posed by potential actors (proliferant States or sub-national adversaries). The characteristics of Generation IV systems, both technical and institutional, are used to evaluate the response of the system and to determine its resistance against proliferation threats and robustness against sabotage and terrorism threats. The outcomes of the system response are expressed in terms of a set of measures, which are the high-level PR and PP characteristics of the NES. The methodology is organized to allow evaluations to be performed at the earliest stages of system design and to become more detailed and more representative as the design progresses. It can thus be used to enable a program in safeguards by design or to enhance the conceptual design process of an NES with regard to intrinsic features for PR and PP. The results are intended for three types of users: system designers, program policy makers, and external stakeholders. The methodology has been illustrated in a series of demonstration and case studies and these will be summarized in the paper. Within GIF there is any ongoing activity to

  14. 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

  15. Update of the INPRO Collaborative Project, Proliferation Resistance and Safeguard ability Assessment (Prosta) Tools

    International Nuclear Information System (INIS)

    The objectives of the INPRO Collaborative Project, Proliferation Resistance and Safeguard ability Assessment (PROSA) Tools are to make the INPRO proliferation resistance (PR) assessment methodology simpler and easier to use, to allow for different users and depths of analysis, to demonstrate the value and its usefulness of the refined assessment methodology to potential users, through a test with a reference case, and to provide input to a revision of the INPRO PR assessment manual. A summary of the project is described herein, including the procedure of PR assessment process and a case study using a SFR metal fuel manufacturing facility (SFMF) which is currently in the conceptual design phase at KAERI. The PROSA process with questionnaire approach is simpler and easier to perform that the original INPRO PR methodology with qualitative scale from 'weak' to 'very strong' to be determined by expert judgment. The PROSA process can be applied from the early stage of design showing the relationship of PR assessment to the SBD process

  16. Pragmatic approaches for assessing and implementing proliferation resistance of nuclear systems

    International Nuclear Information System (INIS)

    This paper is aiming at discussing practical approaches which may be easily implemented to assess and make use of provisions for enhancing Proliferation Resistance (PR) of nuclear systems (that is nuclear reactors and their associated fuel cycles). In the initiatives to develop innovative nuclear energy system (Generation 4, INPRO), PR is one of the key elements, along with economics, safety, sustainability and environment which has to be addressed. Assessment of proliferation resistance is therefore of timely importance. In a first part, we discuss the potential benefits that nuclear safety approach may bring to PR assessment. As a matter of fact, PR design and assessment are calling for the use of the principle of Defence in depth. This concept is widely and successfully used in nuclear safety. In the field of PR, we believe that Probability Risk Assessment (PRA) can be a valuable complement to the attribute methodology. More generally, we show how the general principles and methods of nuclear safety may be applied to PR. In a second part, as an illustration of this approach, we present a simple method for assessing PR, which uses the classical concept of barriers. In this method, four categories of barriers are distinguished: material, technical, institutional and specific barriers for the weapon making phase. In the proliferation process (or route), four steps are considered: diversion of materials, transformation, transport and making of the nuclear weapon using either high enriched uranium or plutonium. All steps of the fuel cycle from uranium mining to final disposals of spent fuels or nuclear waste are examined. A scale of value is then defined in order to quantify each of these elementary steps and figures and aggregated to obtain global performance indices for PR and to identify weak points. Various nuclear systems have been analyzed and general conclusions drawn from these results are presented. (author)

  17. 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

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

    International Nuclear Information System (INIS)

    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) p21Waf1/Cip1 and p27Kip1 descended in PPARγ1S84D stable HT1080 cell, whereas the expression of p18INK4C was not changed. Moreover, compared to the PPARγ1S84A, PPARγ1S84D 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 cells growth

  19. Trends in BWR transient analysis

    International Nuclear Information System (INIS)

    While boiling water reactor (BWR) analysis methods for transient and loss of coolant accident analysis are well established, refinements and improvements continue to be made. This evolution of BWR analysis methods is driven by the new applications. This paper discusses some examples of these trends, specifically, time domain stability analysis and analysis of the simplified BWR (SBWR), General Electric's design approach involving a shift from active to passive safety systems and the elimination/simplification of systems for improved operation and maintenance

  20. BWR stability analysis

    International Nuclear Information System (INIS)

    The objective of this study has been to examine TVO-I oscillation incident, which occured in February 22.1987 and to find out safety implications of oscillations in ATWS incidents. Calculations have been performed with RAMONA-3B and TRAB codes. RAMONA-3B is a BWR transient analysis code with three-dimencional neutron kinetics and nonequilibrium, nonhomogeneous thermal hydraulics. TRAB code is a one-dimencional BWR transient code which uses methods similar to RAMONA-3B. The results have shown that both codes are capable of analyzing of the oscillation incidents. Both out-of-phase and in-phase oscillations are possible. If the reactor scram fails (ATWS) during oscillations the severe fuel failures are always possible and the reactor core may exceed the prompt criticality

  1. 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. PMID:26372743

  2. Modernising Sweden's oldest BWR

    International Nuclear Information System (INIS)

    A new range of digital, programmable process control systems has been used to upgrade the instrumentation and control room of Oskarshamn 1's power station. It will enable Sweden's oldest BWR reactor to comply with higher safety levels, improving the operator's overview of production, and reducing the risk of human error. The ''Advant Power'' range of systems will also be used for future planned improvements to the control room. (UK)

  3. BWR AXIAL PROFILE

    International Nuclear Information System (INIS)

    The purpose of this calculation is to develop axial profiles for estimating the axial variation in burnup of a boiling water reactor (BWR) assembly spent nuclear fuel (SNF) given the average burnup of an assembly. A discharged fuel assembly typically exhibits higher burnup in the center and lower burnup at the ends of the assembly. Criticality safety analyses taking credit for SNF burnup must account for axially varying burnup relative to calculations based on uniformly distributed assembly average burnup due to the under-burned tips. Thus, accounting for axially varying burnup in criticality analyses is also referred to as accounting for the ''end effect'' reactivity. The magnitude of the reactivity change due to ''end effect'' is dependent on the initial assembly enrichment, the assembly average burnup, and the particular axial profile characterizing the burnup distribution. The set of bounding axial profiles should incorporate multiple BWR core designs and provide statistical confidence (95 percent confidence that 95 percent of the population is bound by the profile) that end nodes are conservatively represented. The profiles should also conserve the overall burnup of the fuel assembly. More background on BWR axial profiles is provided in Attachment I

  4. BWR AXIAL PROFILE

    Energy Technology Data Exchange (ETDEWEB)

    J. Huffer

    2004-09-28

    The purpose of this calculation is to develop axial profiles for estimating the axial variation in burnup of a boiling water reactor (BWR) assembly spent nuclear fuel (SNF) given the average burnup of an assembly. A discharged fuel assembly typically exhibits higher burnup in the center and lower burnup at the ends of the assembly. Criticality safety analyses taking credit for SNF burnup must account for axially varying burnup relative to calculations based on uniformly distributed assembly average burnup due to the under-burned tips. Thus, accounting for axially varying burnup in criticality analyses is also referred to as accounting for the ''end effect'' reactivity. The magnitude of the reactivity change due to ''end effect'' is dependent on the initial assembly enrichment, the assembly average burnup, and the particular axial profile characterizing the burnup distribution. The set of bounding axial profiles should incorporate multiple BWR core designs and provide statistical confidence (95 percent confidence that 95 percent of the population is bound by the profile) that end nodes are conservatively represented. The profiles should also conserve the overall burnup of the fuel assembly. More background on BWR axial profiles is provided in Attachment I.

  5. The secure, transportable, autonomous reactor (STAR): a small proliferation-resistant reactor system for developing countries

    Energy Technology Data Exchange (ETDEWEB)

    Brown, N W; Hassberger, J A; Smith, C F

    1999-05-27

    The Secure, Transportable, Autonomous Reactor (STAR), is an integrated concept for a small, proliferation-resistant nuclear power system capable of meeting the growing power demands of many regions of the developing world. The STAR approach builds on earlier work investigating the features required for implementation of such a system. The STAR approach includes establishing overall system requirements, conducting research into issues common to four reactor concepts (gas, liquid metal, light water and molten salt), and defining and performing the down-selection to a preferred concept that will serve as the basis for continued development leading to an eventual prototype. The paper indicates that a number of unique and distinguishing innovations are needed to both meet the energy demands of most of the world's developing regions and address growing nuclear proliferation concerns. These technical innovations form much of the basis underlying the STAR concept and include: eliminating on-site refueling and fuel access; incorporating a systems approach to nuclear energy supply and infrastructure design, with all aspects of equipment life, fuel and waste cycles included; small unit size enabling transportability; replaceable standardized modular design; resilient and robust design concepts leading to large safety margins, high reliability and reduced maintenance; simplicity in operation with reliance on autonomous control and remote monitoring; and waste minimization and waste form optimization.

  6. Potential of CANDLE reactor on sustainable development and strengthened proliferation resistance

    International Nuclear Information System (INIS)

    CANDLE reactor generates energy by using only natural or depleted uranium as make up fuel and achieves about 40% burn up without fuel recycling (Sekimoto, H., Ryu, K., Yoshimura, Y., 2001. CANDLE: the new burnup strategy. Nucl. Sci. Eng. 139 (3), 306-317). These distinctive characteristics eliminate the necessity of both enrichment and reprocessing processes that are recognized as essentially inevitable parts in the conventional nuclear energy concept. This paper describes that the potential performance of CANDLE reactor to meet a projected energy demand growth of the world with stabilizing carbon dioxide concentration in the atmosphere and simultaneously minimize the risk of nuclear material proliferation. A type of CANDLE reactor with moderate initial fissile inventory is feasible to be deployed with prompt enough introduction pace to satisfy the worldwide energy growth and limit the carbon dioxide concentration to about 550 ppm in the next century. Ultimately high proliferation resistance performance as a fission energy system is found for the CANDLE system due to the elimination of most vulnerable processes in the conventional fuel cycle. (author)

  7. The status of proliferation resistance evaluation methodology development in GEN IV international forum

    International Nuclear Information System (INIS)

    The Generation IV Nuclear Energy Systems International Forum (GIF) Proliferation Resistance and Physical Protection Working Group (PR and PP WG) was established in December 2002 in order to develop the PR and PP evaluation methodology for GEN IV nuclear energy systems. The methodology has been studied and established by international consensus. The PR and PP WG activities include development of the measures and metrics; establishment of the framework of PR and PP evaluation, the demonstration study using Example Sodium Fast Reactor (ESFR), which included the development of three evaluation approaches; the Case Study using ESFR and four kinds of threat scenarios; the joint study with GIF System Steering Committees (SSCs) of the six reactor design concepts; and the harmonization study with the IAEA's International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO). This paper reviews the status of GIF PR and PP studies and identifies the challenges and directions for applying the methodology to evaluate future nuclear energy systems in Japan. (author)

  8. Evaluating Proliferation Resistance and Safeguards in the Nuclear Fuel Cycle through a Fuzzy Logic Barrier Framework

    International Nuclear Information System (INIS)

    With the renewed global interest in nuclear power, modeling proliferation resistance (PR) has become more critical than ever to the maintaining and improving the security of the nuclear fuel cycle. One of the most prevalent approaches to this problem has been Probabilistic Risk Assessment (PRA), in which specific proliferation scenarios and pathways are considered and developed. The disadvantage of this approach is its resource-intensiveness; often PRA methods require access to sensitive (or even classified) information. A proposed alternative to PRA approaches is a Fuzzy Logic Barrier Model (FLBM), in which each fuel cycle stage is broken up into attributes which represent 'barriers' to a proliferation attempt. Fuzzy logic naturally allows for a means of assessing linguistic variables (such as those found in multi-attribute utility analysis models), thus allowing for a method of qualitative analysis of different barriers. Likewise, by establishing a consistent process for establishing barrier weights and 'calibrating' the weights given a well-characterized system, the envelope of subjectivity for this method can be significantly constrained. The advantage of this approach is in its ability to function as an open, configurable model which can be used to characterize PR for various fuel cycle technologies. The approach can also be extended to identify both critical needs areas for proliferation safeguards as well as the efficacy of current and proposed safeguards. Such a model can be applied to analyze the PR performance of a proposed fuel cycle system at the individual stage level (making a relative comparison of the vector of the PR value between stages), and by decomposing individual stages to quantify the effect of individual barriers upon the stage (and therefore system) PR behavior. The obvious application of this technique would thus be to identify and characterize 'weak links' in the intrinsic properties of the fuel throughout different fuel cycles; in this

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

    OpenAIRE

    Xiaoyan Sheng; Yuebo Zhang; Zhenwei Gong; Cheng Huang; Ying Qin Zang

    2008-01-01

    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 increase...

  10. Disruption of insulin receptor function inhibits proliferation in endocrine-resistant breast cancer cells.

    Science.gov (United States)

    Chan, J Y; LaPara, K; Yee, D

    2016-08-11

    The insulin-like growth factor (IGF) system is a well-studied growth regulatory pathway implicated in breast cancer biology. Clinical trials testing monoclonal antibodies directed against the type I IGF receptor (IGF1R) in combination with estrogen receptor-α (ER) targeting have been completed, but failed to show benefits in patients with endocrine-resistant tumors compared to ER targeting alone. We have previously shown that the closely related insulin receptor (InsR) is expressed in tamoxifen-resistant (TamR) breast cancer cells. Here we examined if inhibition of InsR affected TamR breast cancer cells. InsR function was inhibited by three different mechanisms: InsR short hairpin RNA, a small InsR-blocking peptide, S961 and an InsR monoclonal antibody (mAb). Suppression of InsR function by these methods in TamR cells successfully blocked insulin-mediated signaling, monolayer proliferation, cell cycle progression and anchorage-independent growth. This strategy was not effective in parental cells likely because of the presence of IGFR /InsR hybrid receptors. Downregulation of IGF1R in conjunction with InsR inhibition was more effective in blocking IGF- and insulin-mediated signaling and growth in parental cells compared with single-receptor targeting alone. Our findings show TamR cells were stimulated by InsR and were not sensitive to IGF1R inhibition, whereas in tamoxifen-sensitive parental cancer cells, the presence of both receptors, especially hybrid receptors, allowed cross-reactivity of ligand-mediated activation and growth. To suppress the IGF system, targeting of both IGF1R and InsR is optimal in endocrine-sensitive and -resistant breast cancer. PMID:26876199

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

    International Nuclear Information System (INIS)

    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

  12. 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.

  13. Status of the Gen-IV Proliferation Resistance and Physical Protection (PRPP) Evaluation Methodology

    International Nuclear Information System (INIS)

    Methodologies have been developed within the Generation IV International Forum (GIF) to support the assessment and improvement of system performance in the areas safeguards, security, economics and safety. Of these four areas, safeguards and security are the subjects of the GIF working group on Proliferation Resistance and Physical Protection (PRPP). Since the PRPP methodology (now at Revision 6) represents a mature, generic, and comprehensive evaluation approach, and is freely available on the GIF public website, several non-GIF technical groups have chosen to utilize the PRPP methodology for their own goals. Indeed, the results of the evaluations performed with the methodology are intended for three types of generic users: system designers, programme policy makers, and external stakeholders. The PRPP Working Group developed the methodology through a series of demonstration and case studies. In addition, over the past few years various national and international groups have applied the methodology to inform nuclear energy system designs, as well as to support the development of approaches to advanced safeguards. A number of international workshops have also been held which have introduced the methodology to design groups and other stakeholders. In this paper we summarize the technical progress and accomplishments of the PRPP evaluation methodology, including applications outside GIF, and we outline the PRPP methodology's relationship with the IAEA's INPRO methodology. Current challenges with the efficient implementation of the methodology are outlined, along with our path forward for increasing its accessibility to a broader stakeholder audience - including supporting the next generation of skilled professionals in the nuclear non-proliferation field. (author)

  14. Introduction to the Revision 5 of evaluation methodology for proliferation resistance and physical protection of Generation IV nuclear energy systems

    International Nuclear Information System (INIS)

    The Generation IV (GEN IV) International Forum (GIF) Proliferation Resistance and Physical Protection Working Group (PRPP WG) was established in December, 2002, as one of the crosscut groups under GIF, in order to develop a methodology for evaluating Proliferation Resistance and Physical Protection of potential GEN IV options. The group currently consists of the experts from the U.S. national laboratories and universities, from Canada, France, Republic of Korea (ROK), Japan, the International Atomic Energy Agency (IAEA), and European Union(EU). Revision 5 of the 'Evaluation Methodology for Proliferation Resistance and Physical Protection of Generation IV Nuclear Energy Systems' is also known as the Rev. 5 report. The Rev. 5 report provides an important evaluation framework that was developed with consensus-based discussions, and published by the OECD-NEA in November, 2006. The activities of PRPP WG has contributed to establish PR and PP culture, in addition, they have affected the discussion on the proliferation resistance of the future nuclear fuel cycle. Japan has to develop the future nuclear cycle system with sufficient proliferation resistance and physical protection features and demonstrate and explain about its effective NESs to the domestic and international society. For these reasons, and recognizing the usefulness of the Revision-5 Report, it was translated and published here as a Japanese-language edition with the concurrence of the OECD-NEA. The original report in English language can be downloaded at the OECD-NEA website. The translation sticks as closely as possible to the original, and that of technical terms were carefully done. If a term cannot be translated appropriately into Japanese, the original English wording is also Put down. Safeguards terms were translated with reference to 'IAEA Safeguards Glossary 2001 Edition' (Japanese), Published by the Nuclear Material Control Center Japan (NMCC). The authors are grateful to the GIF PRPP WG Co

  15. BWR 90 and BWR 90+: Two advanced BWR design generations from ABB

    International Nuclear Information System (INIS)

    ABB has two evolutionary advanced light water reactors available today - the BWR 90 boiling water reactor and the System 80+ pressurised water reactor. The BWR 90 is based on the design, construction, commissioning and operation of the BWR 75 plants. The operation experience of the six plants of this advanced design has been very good. The average annual energy availability is above 90%, and total power generation costs have been low. When developing the BWR 90 specific changes were introduced to a reference design, to adapt to technological progress, new safety requirements and to achieve cost savings. The thermal power rating of BWR 90 is 3800 MWth (providing a nominal 1374 MWe net), slightly higher than that of the reference plant ABB Atom has taken advantage of margins gained using a new generation of its SVEA fuel to attain this power rating without major design modifications. The BWR 90 design was completed and offered to the TVO utility in Finland in 1991, as one of the contenders for the fifth Finnish nuclear power plant project. Hence, the design is available today for deployment in new plant projects. Utility views were incorporated through co-operation with the Finnish utility TVO, owner and operator of the two Olkiluoto plants of BWR 75 design. A review against the European Utility Requirement (EUR) set of requirements has been performed, since the design, in 1997, was selected by the EUR Steering Committee to be the first BWR to be evaluated against the EUR documents. The review work was completed in 1998. It will be the subject of an 'EUR Volume 3 Subset for BWR 90' document. ABB is continuing its BWR development work with an 'evolutionary' design called BWR 90+, which aims at developing the BWR as a competitive option for the anticipated revival of the market for new nuclear plants beyond the turn of the century, as well as feeding ideas and inputs to the continuous modernisation efforts at operating plants. The development is performed by ABB Atom

  16. Expected new role of IAEA in the area of transparency and proliferation resistance in advanced nuclear fuel cycle

    International Nuclear Information System (INIS)

    International confidence of nuclear nonproliferation in the peaceful nuclear program has been maintained by 'nuclear material control and accountancy' and 'international safeguards' under International Atomic Energy Agency (IAEA) and Nuclear Nonproliferation Treaty (NPT) framework. International nuclear non-proliferation efforts have been changing from the early 1990s. United Nation inspections in Iraq after the 1st Gulf War raised doubts about the scope of international safeguards, such as how to deal with undeclared activities. Recent Iran and North Korea issues showed the limitations of international safeguards systems. Incompletely effective non-proliferation regime arouses further proliferation concerns and decreases the confidence of peaceful nuclear program. Especially, a country like Japan, which has full scale nuclear fuel cycle program sometimes could be exposed to criticism. Corresponding to this serious situation, JAEA has conducted studies on 'transparency' and 'proliferation resistance' to build international confidence in its peaceful nuclear programs. Generally 'transparency' has a role to increase confidence in activities and 'proliferation resistance' increases confidence in the capability of the facility. Several efforts to improve the 'transparency' such as 'Guidelines for the Management of Plutonium (INFCIRC/549 of 16 March 1998)' are going on. Joyo Remote Monitoring System has been developed under joint study with Sandia National Laboratory to test the technology to improve transparency. Several discussions on 'proliferation resistance' under international program such as International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) and Generation IV nuclear energy system (GEN-IV) are in progress. As proactive measures in response to the September eleven terrorist attacks, the IAEA Board of Governors approved an action plan designed to upgrade worldwide protection against acts of terrorism involving nuclear and other radioactive

  17. The concept of proliferation-resistant, environment-friendly, accident-tolerant, continual, and economical reactor (PEACER)

    International Nuclear Information System (INIS)

    As an effort to ameliorate generic concerns with current power reactors such as the risk of proliferation, radiological hazard of the spent fuel, and the vulnerability to core-melt accidents, the concept of a revolutionary reactor, named as PEACER, has been developed as a proliferation-resistant waste transmutation reactor with its technical footing on proven technologies of critical reactors and heavy liquid metal coolant. In this paper, results of PEACER conceptual design are summarized with the focus on the neutronic characteristics and the expected general system performance. The proliferation resistance of PEACER is built by installing both institutional and technical barriers. The latter includes denaturing of fissile materials, Pu in particular, as well as the intense radiation field associated with the pyrochemical partitioning method. When the fuel volume fraction and the core aspect ratio(L/D) are optimized, the transmutation capability of PEACER for long-living wastes from LWR spent fuels is found to exceed their production rate of two LWRs each at the same electric rating. In contrast with current power reactor design principles, the lower power density and the higher neutron leakage rate lead to the higher performance on the proliferation-resistance, transmutation capability and the accident-tolerance. The conceptual design result has shown promising characteristics in all the five target areas defined by it s name PEACER, warranting more detailed studies

  18. Thermal-hydraulic development a small, simplified, proliferation-resistant reactor

    International Nuclear Information System (INIS)

    This paper addresses thermal-hydraulics related criteria and preliminary concepts for a small (300 MWt), proliferation-resistant, liquid-metal-cooled reactor system. A main objective is to assess what extent of simplification is achievable in the concepts with the primary purpose of regaining economic competitiveness. The approach investigated features lead-bismuth eutectic (LBE) and a low power density core for ultra-long core lifetime (goal 15 years) with cartridge core replacement at end of life. This potentially introduces extensive simplifications resulting in capital cost and operating cost savings including: (1) compact, modular, pool-type configuration for factory fabrication, (2) 100+% natural circulation heat transport with the possibility of eliminating the main coolant pumps, (3) steam generator modules immersed directly in the primary coolant pool for elimination of the intermediate heat transport system, and (4) elimination of on-site fuel handling and storage provisions including rotating plug. Stage 1 natural circulation model and results are presented. Results suggest that 100+% natural circulation heat transport is readily achievable using LBE coolant and the long-life cartridge core approach; moreover, it is achievable in a compact pool configuration considerably smaller than PRISM A (for overland transportability) and with peak cladding temperature within the existing database range for ferritic steel with oxide layer surface passivation. Stage 2 analysis follows iteration with core designers. Other thermal hydraulic investigations are underway addressing passive, auxiliary heat removal by air cooling of the reactor vessel and the effects of steam generator tube rupture

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

    International Nuclear Information System (INIS)

    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

  20. 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.

  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. Natural convection type BWR reactor

    International Nuclear Information System (INIS)

    In a natural convection type BWR reactor, a mixed stream of steams and water undergo a great flow resistance. In particular, pressure loss upon passing from an upper plenum to a stand pipe and pressure loss upon passing through rotational blades are great. Then, a steam dryer comprising laminated dome-like perforated plates and a drain pipe for flowing down separated water to a downcomer are disposed above a riser. The coolants heated in the reactor core are boiled, uprise in the riser as a gas-liquid two phase flow containing voids, release steams containing droplets from the surface of the gas-liquid two phase, flow into the steam dryer comprising the perforated plates and are separated into a gas and a liquid. The dried steams flow to a turbine passing through a main steam pipe and the condensated droplets flow down through the drain pipe and the downcomer to the lower portion of the reactor core. In this way, the conventional gas-liquid separator can be saved without lowering the quality of steam drying to reduce the pressure loss and to improve the operation performance. (N.H.)

  3. BWR fuel performance

    International Nuclear Information System (INIS)

    The General Electric experience base on BWR fuel includes over 29,000 fuel assemblies which contain 1,600,000 fuel rods. Over the last five years, design, process and operating changes have been introduced which have had major effects in improving fuel performance. Monitoring this fuel performance in BWRs has been accomplished through cooperative programs between GE and utilities. Activities such as plant fission product monitoring, fuel sipping and fuel and channel surveillance programs have jointly contributed to the value of this extensive experience base. The systematic evaluation of this data has established well-defined fuel performance trends which provide the assurance and confidence in fuel reliability that only actual operating experience can provide

  4. Proliferation resistant fuel cycle system for the transition from light water reactors to fast reactors

    International Nuclear Information System (INIS)

    Full text: Introduction of commercial fast reactors (FR) is predicted to start around 2050 in Japan. Effective utilization of plutonium in FR is important for the sustainable electricity generation by nuclear. Successive replacement of light water reactors (LWR) to FR will take more than 60-years and reasonable fuel cycle management is necessary during this period. The transition scenario has various unpredictable factors such as introduction speed and time of FR, and flexible fuel cycle system was proposed to respond to these factors. The system consists of LWR and FR spent fuels reprocessing for reduction of LWR spent fuel volume and FR fuel fabrication. LWR fuel reprocessing only carries out about 90% uranium removal from LWR spent fuel, then the composition of residual spent fuel called recycle material is about 50% uranium, 15% plutonium and 35% fission products + minor actinides. Recycle material is transferred to FR fuel reprocessing to recover plutonium and uranium followed by mixed oxide (MOX) fuel fabrication for FR with radioactive impurities. Depending on the introduction time of FR, recycle material (about 1/10 volume of original spent fuel) may be stored for future use. The system has some characteristics compared with ordinary system that consists of full reprocessing facilities for LWR and FR spent fuels to produce FR fresh fuels. The LWR reprocessing facility becomes much smaller due to no Pu-U recovery and fabrication. The recycle materials can supply higher content of plutonium to FR and be compactly stored in case of FR introduction delay. Plutonium always contains uranium and impurities (fission products and minor actinides), thus the system maintains high proliferation resistance. The plutonium balance was calculated under several conditions, which revealed that the system could supply enough and no excess plutonium to FR

  5. miR-421 induces cell proliferation and apoptosis resistance in human nasopharyngeal carcinoma via downregulation of FOXO4

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Liang [Neurosurgery Institute, Key Laboratory on Brain Function Repair and Regeneration of Guangdong, Zhujiang Hospital of Southern Medical University, Guangzhou 510282 (China); Department of Otolaryngology, Guangzhou General Hospital of PLA Guangzhou Command, Guangzhou 510010 (China); Tang, Yanping [Neurosurgery Institute, Key Laboratory on Brain Function Repair and Regeneration of Guangdong, Zhujiang Hospital of Southern Medical University, Guangzhou 510282 (China); Wang, Jian [Department of Otolaryngology, Guangzhou General Hospital of PLA Guangzhou Command, Guangzhou 510010 (China); Yan, Zhongjie [Affiliated Bayi Brain Hospital, The Military General Hospital of Beijing PLA,The Bayi Clinical Medical Institute of Southern Medical University, Beijing 100700 (China); Xu, Ruxiang, E-mail: RuxiangXu@yahoo.com [Affiliated Bayi Brain Hospital, The Military General Hospital of Beijing PLA,The Bayi Clinical Medical Institute of Southern Medical University, Beijing 100700 (China)

    2013-06-14

    Highlights: •miR-421 is upregulated in nasopharyngeal carcinoma. •miR-421 induces cell proliferation and apoptosis resistance. •FOXO4 is a direct and functional target of miR-421. -- Abstract: microRNAs have been demonstrated to play important roles in cancer development and progression. Hence, identifying functional microRNAs and better understanding of the underlying molecular mechanisms would provide new clues for the development of targeted cancer therapies. Herein, we reported that a microRNA, miR-421 played an oncogenic role in nasopharyngeal carcinoma. Upregulation of miR-421 induced, whereas inhibition of miR-421 repressed cell proliferation and apoptosis resistance. Furthermore, we found that upregulation of miR-421 inhibited forkhead box protein O4 (FOXO4) signaling pathway following downregulation of p21, p27, Bim and FASL expression by directly targeting FOXO4 3′UTR. Additionally, we demonstrated that FOXO4 expression is critical for miR-421-induced cell growth and apoptosis resistance. Taken together, our findings not only suggest that miR-421 promotes nasopharyngeal carcinoma cell proliferation and anti-apoptosis, but also uncover a novel regulatory mechanism for inactivation of FOXO4 in nasopharyngeal carcinoma.

  6. BWR type nuclear reactor

    International Nuclear Information System (INIS)

    Purpose: To simplify the structure of an emergency core cooling system while suppressing the flow out of coolants upon rapture accidents in a coolant recycling device of BWR type reactors. Constitution: Recirculation pumps are located at a position higher than the reactor core in a pressure vessel, and the lower plenum is bisected vertically by a partition plate. Further, a gas-liquid separator is surrounded with a wall and the water level at the outer side of the wall is made higher than the water level in the inside of the wall. In this structure, coolants are introduced from the upper chamber in the lower plenum into the reactor core, and the steams generated in the reactor core are separated in the gas-liquid separator, whereby the separated liquid is introduced as coolants by way of the inner chamber into the lower chamber of the lower plenum and further sent by way of the outer chamber into the reactor core. Consequently, idle rotation of the recycling pumps due to the flow-in of saturated water is prevented and loss of coolants in the reactor core can also be prevented upon raptures in the pipeway and the driving section of the pump connected to the pressure vessel and in the bottom of the pressure vessel. (Horiuchi, T.)

  7. Malignant T cells exhibit CD45 resistant Stat3 activation and proliferation in cutaneous

    DEFF Research Database (Denmark)

    Krejsgaard, Thorbjørn Frej; Helvad, Rikke; Ralfkiaer, Elisabeth;

    2010-01-01

    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......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...

  8. Implication of protein tyrosine phosphatase 1B in MCF-7 cell proliferation and resistance to 4-OH tamoxifen

    Energy Technology Data Exchange (ETDEWEB)

    Blanquart, Christophe; Karouri, Salah-Eddine [Institut Cochin, Universite Paris Descartes, CNRS (UMR 8104), Paris (France); Inserm, U567, Paris (France); Issad, Tarik, E-mail: tarik.issad@inserm.fr [Institut Cochin, Universite Paris Descartes, CNRS (UMR 8104), Paris (France); Inserm, U567, Paris (France)

    2009-10-02

    The protein tyrosine phosphatase 1B (PTP1B) and the T-cell protein tyrosine phosphatase (TC-PTP) were initially thought to be mainly anti-oncogenic. However, overexpression of PTP1B and TC-PTP has been observed in human tumors, and recent studies have demonstrated that PTP1B contributes to the appearance of breast tumors by modulating ERK pathway. In the present work, we observed that decreasing the expression of TC-PTP or PTP1B in MCF-7 cells using siRNA reduced cell proliferation without affecting cell death. This reduction in proliferation was associated with decreased ERK phosphorylation. Moreover, selection of tamoxifen-resistant MCF-7 cells, by long-term culture in presence of 4-OH tamoxifen, resulted in cells that display overexpression of PTP1B and TC-PTP, and concomitant increase in ERK and STAT3 phosphorylation. siRNA experiments showed that PTP1B, but not TC-PTP, is necessary for resistance to 4-OH tamoxifen. Therefore, our work indicates that PTP1B could be a relevant therapeutic target for treatment of tamoxifen-resistant breast cancers.

  9. Insulin, pioglitazone and Zingiber officinale administrations improve proliferating cell nuclear antigen immunostaining effects on diabetic and insulin resistant rat testis

    OpenAIRE

    DARAMOLA, Adetola Olubunmi; OLATUNJI-BELLO, Ibiyemi Ibitola; OBIKA, Leonard Fidelis

    2013-01-01

    This study accessed the effects of hypoglycaemic drugs on spermatogenesis in diabetic and insulin resistant rat testis following proliferating cell nuclear antigen (PCNA) immunostaining. Male adult Sprague-Dawley rats (120-140 g) were randomly divided into 5 groups. Group 1 served as control group; fed on normal rat pellets. Group 2 served as streptozotocin-insulin treated group; received a single dose IP Injection of streptozotocin 45 mg/kg BW in Na+ citrate buffer pH 4.5 and treated with in...

  10. 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.

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

    International Nuclear Information System (INIS)

    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

  12. Optimization of Heterogeneous Utilization of Thorium in PRWs to Enhance Proliferation Resistance & Reduce Waste

    Energy Technology Data Exchange (ETDEWEB)

    Mujid Kazimi

    2003-12-18

    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 aside from the alpha (n reaction on the 240 Pu isotope) does not present any significant intrinsic barrier to weapon assembly

  13. Optimization of Heterogeneous Utilization of Thorium in PRWs to Enhance Proliferation Resistance and Reduce Waste

    International Nuclear Information System (INIS)

    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 aside from the alpha (n reaction on the 240 Pu isotope) does not present any significant intrinsic barrier to weapon assembly

  14. 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.

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

    International Nuclear Information System (INIS)

    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

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

    Science.gov (United States)

    Han, Seong-Su; Han, Sangwoo; Kamberos, Natalie L

    2014-09-26

    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. PMID:25193702

  18. 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

  19. Improvement for BWR operator training

    International Nuclear Information System (INIS)

    BWR Operator Training Center Corporation (BTC) was established in April 1971 for the purpose of training the operators from all BWR utilities in Japan. Since April 1974, more than 2600 operators and 1000 shift teams have been trained with the full-scope simulators in BTC up to the end of March 1988. To get the satisfactory results of the training, BTC has been making every effort to improve the facilities, the training materials, the instruction methods and the curricula. In this paper, such a series of recent improvements in the instruction methods and the curricula are presented that are effective to expand the knowledge and to improve the skills of middle or senior class operators. (author)

  20. 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.

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

    International Nuclear Information System (INIS)

    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. BWR radiation control: plant demonstration

    International Nuclear Information System (INIS)

    The first year's progress is presented for a four-year program intended to implement and evaluate BRAC radiation reduction operational guidelines at the Vermont Yankee BWR and to document the results in sufficient detail to provide guidance to other BWR owners. Past operational, chemistry and radiation level data have been reviewed to provide a historical base of reference. Extensive sampling and chemistry monitoring systems have been installed to evaluate plant chemistry status and the effects of program implemented changes. Radiation surveys and piping gamma scans are being performed at targeted locations to quantify radiation level trends and to identify and quantify piping isotopics. Contact radiation levels on the recirculation line at Vermont Yankee have been increasing at a rate of 175 mR/h-EFPY since 1978. A materials survey of feedwater and reactor components in contact with the process liquid has been performed to identify sources of corrosion product release, particularly cobalt and nickel. A feedwater oxygen injection system has been installed to evaluate the effects of oxygen control on feedwater materials corrosion product releases. A baseline performance evaluation of the condensate treatment and reactor water cleanup systems has been completed. Data on organics and ionics at Vermont Yankee have been obtained. A methodology of BWR feedwater system layup during extended outages was developed, and an evaluation performed of layup and startup practices utilized at Vermont Yankee during the fall 1980 and 1981 refueling outages

  4. BWR control blade replacement strategies

    International Nuclear Information System (INIS)

    The reactivity control elements in a BWR, the control blades, perform three significant functions: provide shutdown margin during normal and accident operating conditions; provide overall core reactivity control; and provide axial power shaping control. As such, the blades are exposed to the core's neutron flux, resulting in irradiation of blade structural and absorber materials. Since the absorber depletes with time (if B4C is used, it also swells) and the structural components undergo various degradation mechanisms (e.g., embrittlement, corrosion), the blades have limits on their operational lifetimes. Consequently, BWR utilities have implemented strategies that aim to maximize blade lifetimes while balancing operational costs, such as extending a refuelling outage to shuffle high exposure blades. This paper examines the blade replacement strategies used by BWR utilities operating in US, Europe and Asia by assembling information related to: the utility's specific blade replacement strategy; the impact the newer blade designs and changes in core operating mode were having on those strategies; the mechanical and nuclear limits that determined those strategies; the methods employed to ensure that lifetime limits were not exceeded during operation; and blade designs used (current and replacement blades). (author)

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

    International Nuclear Information System (INIS)

    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.

  6. 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; Denning, Richard; Catalyurek, Umit; Yilmaz, Alper; Yue, Meng; Cheng, Lap-Yan

    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.

  7. Argentina: CAREM [Implementation of Proliferation Resistance and Physical Protection Measures in Innovative Small and Medium Sized Reactors

    International Nuclear Information System (INIS)

    CAREM is a project of Argentina’s National Atomic Energy Commission, whose purpose is to develop, design and construct an innovative, simple and small nuclear power plant. This integral type PWR has an indirect cycle with distinctive features that simplify the design and support the objective of achieving a higher level of safety. Argentina has much experience in the surveillance and control of nuclear material. For safeguards implementation in the CAREM reactor, all the refuelling tasks will be developed in the reactor hall, which is designed to allow remote monitoring of all nuclear material handling. An important feature is that there is only one entrance–exit point for controlling and monitoring. This design simplifies accountability and surveillance of nuclear items during movement and transfer. Provisions are considered to cover and to seal the spent fuel pool during operation cycles. Intrinsic proliferation resistance features that facilitate the implementation of extrinsic measures are considered in the CAREM design. Long life cores do not provide cost efficient proliferation resistance for CAREM. The CAREM nuclear power plant has four main handling and storage fuel areas: (1) Fuel reception area; (2) Fresh fuel storage room; (3) Reactor pressure vessel; (4) Spent fuel storage pool. Two isolated material balance areas for spent fuel (pressure vessel core and spent fuel pool) have been included to facilitate safeguards implementation and to reduce costs. All the spent fuel will be located in these two material balance areas, and remote monitoring systems will conduct the surveillance. It is important to note that during nuclear commissioning and operation, there will be no possibility of physical access to fissile material

  8. 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

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

    International Nuclear Information System (INIS)

    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) × 103 to (3.68 ± 0.27) × 106 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) × 107 to (4.30 ± 0.81) × 1010 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 7–308 times higher than that

  10. BWR 90 - the advanced BWR of the 1990s

    International Nuclear Information System (INIS)

    The future global role of nuclear power will be determined by its ability to provide economical and safe energy. Nuclear power, like any other substantial contributor to the world's energy needs, must be generated at an acceptable cost and with negligible environmental effects. Besides, it must achieve and maintain a socially reasonable level of public acceptance, which in turn is not necessarily governed by rational assessments of the true safety and environmental impact of nuclear power. The ABB Atom approach to this situation can best be characterized as a 'cautious evolution'; for the next decade the company will largely base its offerings to the market on its 'evolutionary' light water reactor design, the BWR 90. This design builds closely on the experience from successful construction and operation of its predecessor, the BWR 75 design. In 1995 and 1996, plants of this design achieved an average load factor greater than the 87% set by EUR; the two BWR units at Olkiluoto in Finland are among the very best performing plants in the world, with an average load factor of 94% over the last 7 years. The continued LWR design development focuses on meeting requirements from utilities as well as new regulatory requirements. A particular emphasis is put on the consequences of severe accidents; there shall be no large releases to the environment. Other design improvements involve: all-digital I and C systems and enhanced human factors engineering to improve work environment for operators, optimization of buildings and containment design to decrease construction time and costs, and selection of materials as well as maintenance and operating procedures to even further reduce occupational radiation exposures. (orig.)

  11. Sorghum Dw1, an agronomically important gene for lodging resistance, encodes a novel protein involved in cell proliferation.

    Science.gov (United States)

    Yamaguchi, Miki; Fujimoto, Haruka; Hirano, Ko; Araki-Nakamura, Satoko; Ohmae-Shinohara, Kozue; Fujii, Akihiro; Tsunashima, Masako; Song, Xian Jun; Ito, Yusuke; Nagae, Rie; Wu, Jianzhong; Mizuno, Hiroshi; Yonemaru, Jun-Ichi; Matsumoto, Takashi; Kitano, Hidemi; Matsuoka, Makoto; Kasuga, Shigemitsu; Sazuka, Takashi

    2016-01-01

    Semi-dwarfing genes have contributed to enhanced lodging resistance, resulting in increased crop productivity. In the history of grain sorghum breeding, the spontaneous mutation, dw1 found in Memphis in 1905, was the first widely used semi-dwarfing gene. Here, we report the identification and characterization of Dw1. We performed quantitative trait locus (QTL) analysis and cloning, and revealed that Dw1 encodes a novel uncharacterized protein. Knockdown or T-DNA insertion lines of orthologous genes in rice and Arabidopsis also showed semi-dwarfism similar to that of a nearly isogenic line (NIL) carrying dw1 (NIL-dw1) of sorghum. A histological analysis of the NIL-dw1 revealed that the longitudinal parenchymal cell lengths of the internode were almost the same between NIL-dw1 and wildtype, while the number of cells per internode was significantly reduced in NIL-dw1. NIL-dw1dw3, carrying both dw1 and dw3 (involved in auxin transport), showed a synergistic phenotype. These observations demonstrate that the dw1 reduced the cell proliferation activity in the internodes, and the synergistic effect of dw1 and dw3 contributes to improved lodging resistance and mechanical harvesting. PMID:27329702

  12. 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......-sections with a monoclonal antibody against neural cell adhesion molecule (NCAM) and counterstaining with Mayer's hematoxylin. Compared with the pre-training values, there was a significant increase (P<0.05) in the number of NCAM-positively stained cells per fiber post-training in males (from 0.11+/-0.03 to 0...

  13. Advanced Construction of Compact Containment BWR

    International Nuclear Information System (INIS)

    The reactor concept considered in this paper has a mid/small power output, a compact containment and a simplified BWR configuration with comprehensive safety features. Compact Containment BWR (CCR) is being developed with matured BWR technologies together with innovative systems/components, will provide attractiveness for the energy market in the world due to its flexibility in energy demands as well as in site conditions, its high potential in reducing investment risk and its safety feature facilitating public acceptance. The flexibility is achieved by CCR's mid/small power output of 400 MWe class and capability of long operating cycle (refueling intervals). The high investment potential is expected from CCR's simplification/innovation in design such as natural circulation core cooling with the bottom located short core, top mounted upper entry control rod drives (CRDs) with ring-type dryers and simplified safety system with high pressure resistible primary containment vessel (PCV) concept. The natural circulation core eliminates recirculation pumps as well as needs for maintenance of such pumps. The top mounted upper entry CRDs enable the bottom located short core in RPV. The safety feature mainly consists of large water inventory above the core without large penetration below the top of the core, passive cooling system by isolation condenser (IC), high pressure resistible PCV and in-vessel retention (IVR) capability. The large inventory increases the system response time in case of design base accidents including loss of coolant accidents. The IC suppresses PCV pressure by steam condensation without any AC power. Cooling the molten core inside the RPV if the core should be damaged by loss of core coolability could attain the IVR. CCR's specific self-standing steel high pressure resistible PCV is designed to contain minimum piping and valves inside with reactor pressure vessel (RPV), only 13 m in diameter and 24 m in height. This compact PCV makes it possible to

  14. Long operating cycle simplified BWR

    International Nuclear Information System (INIS)

    Considering next generation requirement for nuclear plants, a long cycle operating simplified BWR (LSBWR) concept is proposed. The major features of LSBWR are; 1) Long cycle operation core using uranium fuels; 2) Simplified system and component as well as passive systems; 3) Combined building concept with ship hull structure. This concept have potential to reduce construction cost and to Increase availability. Safety feature of LSBWR makes possible to attain no evacuation capability in case of a severe accident. Further research and development is underway. (author)

  15. BWR water chemistry impurity studies

    International Nuclear Information System (INIS)

    Laboratory studies were made on the effect of water impurities on environmental cracking in simulated BWR water of stainless steel, low alloy steel and nickel-base alloys. Constant elongation rate tensile (CERT) tests were run in simulated normal water chemistry (NWC), hydrogen water chemistry (HWC), or start-up environment. Sulfate, chloride and copper with chloride added to the water at levels of a fraction of a ppM were found to be extremely deleterious to all kinds of materials except Type 316 NG. Other detrimental impurities were fluoride, silica and some organic acids, although acetic acid was beneficial. Nitrate and carbon dioxide were fairly inoccuous. Corrosion fatigue and constant load tests on compact tension specimens were run in simulated normal BWR water chemistry (NWC) or hydrogen water chemistry (HWC), without impurities or with added sulfate or carbon dioxide. For sensitized Type 304 SS in NWC, 0.1 ppM sulfate increased crack propagation rates in constant load tests by up to a factor of 100, and in fatigue tests up to a factor of 10. Also, cracking in Type 316 nuclear grade SS and Alloy 600 was enhanced, but to a smaller degree. Carbon dioxide was less detrimental than sulfate. 3 figs., 4 tabs

  16. Proliferation inhibition and apoptosis induction of imatinib-resistant chronic myeloid leukemia cells via PPP2R5C down-regulation

    OpenAIRE

    Shen, Qi; Liu, Sichu; Chen, Yu; Yang, Lijian; CHEN, SHAOHUA; Wu, Xiuli; Li, Bo; Lu, Yuhong; Zhu, Kanger; Li, Yangqiu

    2013-01-01

    Despite the success of imatinib and other tyrosine kinase inhibitors (TKIs), chronic myeloid leukemia (CML) remains largely incurable, and a number of CML patients die due to Abl mutation-related drug resistance and blast crisis. The aim of this study was to evaluate proliferation inhibition and apoptosis induction by down-regulating PPP2R5C gene expression in the imatinib-sensitive and imatinib-resistant CML cell lines K562, K562R (imatinib resistant without an Abl gene mutation), 32D-Bcr-Ab...

  17. Targeting AMP-activated protein kinase in adipocytes to modulate obesity-related adipokine production associated with insulin resistance and breast cancer cell proliferation

    Directory of Open Access Journals (Sweden)

    Grisouard Jean

    2011-07-01

    Full Text Available Abstract Background Adipokines, e.g. TNFα, IL-6 and leptin increase insulin resistance, and consequent hyperinsulinaemia influences breast cancer progression. Beside its mitogenic effects, insulin may influence adipokine production from adipocyte stromal cells and paracrine enhancement of breast cancer cell growth. In contrast, adiponectin, another adipokine is protective against breast cancer cell proliferation and insulin resistance. AMP-activated protein kinase (AMPK activity has been found decreased in visceral adipose tissue of insulin-resistant patients. Lipopolysaccharides (LPS link systemic inflammation to high fat diet-induced insulin resistance. Modulation of LPS-induced adipokine production by metformin and AMPK activation might represent an alternative way to treat both, insulin resistance and breast cancer. Methods Human preadipocytes obtained from surgical biopsies were expanded and differentiated in vitro into adipocytes, and incubated with siRNA targeting AMPKalpha1 (72 h, LPS (24 h, 100 μg/ml and/or metformin (24 h, 1 mM followed by mRNA extraction and analyses. Additionally, the supernatant of preadipocytes or derived-adipocytes in culture for 24 h was used as conditioned media to evaluate MCF-7 breast cancer cell proliferation. Results Conditioned media from preadipocyte-derived adipocytes, but not from undifferentiated preadipocytes, increased MCF-7 cell proliferation (p Conclusions Adipocyte-secreted factors enhance breast cancer cell proliferation, while AMPK and metformin improve the LPS-induced adipokine imbalance. Possibly, AMPK activation may provide a new way not only to improve the obesity-related adipokine profile and insulin resistance, but also to prevent obesity-related breast cancer development and progression.

  18. SCORPIO-BWR: status and future plans

    International Nuclear Information System (INIS)

    Full text: During the years from 2000 to 2003 a joint project has been performed by IFE, Halden and TEPCO Systems Corporation, Japan, to develop a core monitoring system for BWRs based on the their existing core monitoring system TiARA and the SCORPIO framework. It has been emphasised to develop a reliable, flexible, adaptable and user-friendly system, which is easy to maintain. Therefore, a rather general framework (SCORPIO Framework) has been used which facilitates easy software modifications as well as adding/ replacing physics modules. The software modules is integrated in the SCORPIO framework using the Software Bus as the communication tool and with the Picasso UIMS tool for MMI. The SCORPIO-BWR version is developed on a Windows-PC platform. The SCORPIO-BWR version provides all functions, which are necessary for all analyses and operations performed on a BWR plant and comprises functions for on-line core monitoring, predictive analysis and core management with interfaces to plant instrumentation and physics codes. Functions for system initialisation and maintenance are also included. A SCORPIO-BWR version adapted for ABWR was installed in TEPSYS facilities in Tokyo in January 2003, where the final acceptance tests were carried out and accepted. The ABWR version of the system is now in the verification and validation phase. In the period from April 2003 until March 2004 a project for realizing an offline-version of SCORPIO-BWR system, which supports the offline tasks of BWR in-core fuel management for ABWR and BWR-5 type of reactors, was developed. The offline-version of the SCORPIO-BWR system for ABWR and BWR-5 type of reactors was installed at TEPSYS in March 2003, where the final acceptance tests were carried out and accepted. Plans for the next version of this system is to study the possibility of adapting SCORPIO-BWR to work with 'mobile technology'. This means that it should be possible to access and display information from the SCORPIO-BWR system on a

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

    International Nuclear Information System (INIS)

    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

  20. 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.

  1. Challenges to Integration of Safety and Reliability with Proliferation Resistance and Physical Protection for Generation IV Nuclear Energy Systems

    Energy Technology Data Exchange (ETDEWEB)

    H. Khalil; P. F. Peterson; R. Bari; G. -L. Fiorini; T. Leahy; R. Versluis

    2012-07-01

    The optimization of a nuclear energy system's performance requires an integrated consideration of multiple design goals - sustainability, safety and reliability (S&R), proliferation resistance and physical protection (PR&PP), and economics - as well as careful evaluation of trade-offs for different system design and operating parameters. Design approaches motivated by each of the goal areas (in isolation from the other goal areas) may be mutually compatible or in conflict. However, no systematic methodology approach has yet been developed to identify and maximize synergies and optimally balance conflicts across the possible design configurations and operating modes of a nuclear energy system. Because most Generation IV systems are at an early stage of development, design, and assessment, designers and analysts are only beginning to identify synergies and conflicts between PR&PP, S&R, and economics goals. The close coupling between PR&PP and S&R goals has motivated early attention within the Generation IV International Forum to their integrated consideration to facilitate the optimization of their effects and the minimization of potential conflicts. This paper discusses the status of this work.

  2. Program of enhancing the Korea-USA cooperation research for the development of proliferation resistant fuel cycle technology

    International Nuclear Information System (INIS)

    The objective of the Program is to develop the fuel cycle technology of GEN-IV SFR (Sodium Fast Reactor) system through the Korea-USA cooperation research in order to improve the efficiency of the technology development and to increase the transparency of the research. Since the pyroprocessing research by using actual spent nuclear fuel can not be performed in Korea at present, the active demonstration research will be performed by using the USA national research facilities under the Korea-USA cooperation. Moreover, the development of safeguards technology and the methodology for the evaluation of the proliferation resistance will also be performed under the cooperation. The current cooperation national laboratories of the safeguards and pyroprocessing technology development are LANL (Los Alamos National Lab.) and INL (Idaho National Lab.), respectively. Practical research experience and technical data for the pyroprocessing technology can be achieved through the demonstration of the inactive research results, which was performed in Korea, by using actual spent nuclear fuel. The scope of the cooperation study encompass the electrolytic reduction of oxide spent fuel, electrorefining, liquid cadmium cathode process, TRU fuel fabrication, fuel performance evaluation and related safeguards technology development

  3. Advances in BWR water chemistry

    International Nuclear Information System (INIS)

    This paper reviews recent advances in Boiling Water Reactor (BWR) water chemistry control with examples of plant experiences at U.S. designed BWRs. Water chemistry advances provide some of the most effective methods for mitigating materials degradation, reducing fuel performance concerns and lowering radiation fields. Mitigation of stress corrosion cracking (SCC) of materials remains a high priority and improved techniques that have been demonstrated in BWRs will be reviewed, specifically hydrogen injection combined with noble metal chemical addition (NMCA) and the newer on-line noble metal application process (OLNC). Hydrogen injection performance, an important part of SCC mitigation, will also be reviewed for the BWR fleet, highlighting system improvements that have enabled earlier injection of hydrogen including the potential for hydrogen injection during plant startup. Water chemistry has been significantly improved by the application of pre-filtration and optimized use of ion exchange resins in the CP (condensate polishing) and reactor water cleanup (RWCU) systems. EPRI has monitored and supported water treatment improvements to meet water chemistry goals as outlined in the EPRI BWR Water Chemistry Guidelines, particularly those for SCC mitigation of reactor internals and piping, minimization of fuel risk due to corrosion and crud deposits and chemistry control for radiation field reduction. In recent years, a significant reduction has occurred in feedwater corrosion product input, particularly iron. A large percentage of plants are now reporting <0.1 ppb feedwater iron. The impacts to plant operation and chemistry of lower feedwater iron will be explored. Depleted zinc addition is widely practiced across the fleet and the enhanced focus on radiation reduction continues to emphasize the importance of controlling radiation source term. In addition, shutdown chemistry control is necessary to avoid excessive release of activated corrosion products from fuel

  4. 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).

  5. 44 BWR Waste Package Loading Curve Evaluation

    International Nuclear Information System (INIS)

    The objective of this calculation is to evaluate the required minimum burnup as a function of average initial boiling water reactor (BWR) assembly enrichment that would permit loading of fuel into a potential 44 BWR waste package (WP). The potential WP design is illustrated in Attachment I. The scope of this calculation covers a range of initial enrichments from 1.5 through 5.0 weight percent U-235, and a burnup range of 0 through 50 GWd/mtU

  6. 44 BWR Waste Package Loading Curve Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    J.M. Scaglione

    2001-11-05

    The objective of this calculation is to evaluate the required minimum burnup as a function of average initial boiling water reactor (BWR) assembly enrichment that would permit loading of fuel into a potential 44 BWR waste package (WP). The potential WP design is illustrated in Attachment I. The scope of this calculation covers a range of initial enrichments from 1.5 through 5.0 weight percent U-235, and a burnup range of 0 through 50 GWd/mtU.

  7. Integration of Safety and Reliability with Proliferation Resistance and Physical Protection for Generation IV Nuclear Energy Systems

    Energy Technology Data Exchange (ETDEWEB)

    Khalil, H. [Argonne National Laboratory (United States); Bari, R. [Brookhaven National Laboratory (United States); Fiorini, G.L. [CEA-Cadarache (France); Leahy, T. [Idaho National Laboratory (United States); Peterson, P.F. [UC Berkeley (United States); Versluis, R. [Department of Energy (United States)

    2009-06-15

    The Generation IV International Forum (GIF) is carrying out research and development for several Generation IV systems with the aim of meeting goals in the areas of sustainability, safety and reliability (S and R), proliferation resistance and physical protection (PR and PP), and economics. It has formed three working groups to help define design guidelines and to develop methodologies to assess progress in meeting the S and R, PR and PP and economics goals: the Risk and Safety Working Group, the PR and PP Working Group, and the Economics Modeling Working Group. Their methodologies [1,2] are at different stages of development, testing, and application for system development and assessment. The optimization of a nuclear energy system's performance requires an integrated consideration of all the goal areas and careful evaluation of tradeoffs for different system design and operating parameters. Design approaches motivated by each of the goal areas (in isolation of the other goal areas) may be mutually compatible or in conflict. However no systematic methodology approach has yet been developed to identify and maximize synergies and optimally balance conflicts across the possible design configurations and operating modes of a nuclear energy system. Because most Generation IV systems are at an early stage of development, design, and assessment, GIF is only beginning to identify synergies and conflicts between PR and PP, S and R, and economics goals. The coupling between PR and PP and R and S goals is perhaps most apparent. For example, passive systems and structures tend to be synergistic for both safety and security, if designed for both functions. Likewise, it is important to know where nuclear materials are and how their protection against hazards and threats is organized from both PR and PP and S and R standpoints. Human performance programs can also be designed to be synergistic for both areas. In emergency response there are clear areas of potential conflict

  8. Effects of a discoloration-resistant calcium aluminosilicate cement on the viability and proliferation of undifferentiated human dental pulp stem cells

    OpenAIRE

    Li-na Niu; Devon Watson; Kyle Thames; Carolyn M. Primus; Bergeron, Brian E.; Kai Jiao; Bortoluzzi, Eduardo A.; Cutler, Christopher W.; Ji-hua Chen; PASHLEY David H.; Franklin R Tay

    2015-01-01

    Discoloration-resistant calcium aluminosilicate cement has been formulated to overcome the timely problem of tooth discoloration reported in the clinical application of bismuth oxide-containing hydraulic cements. The present study examined the effects of this experimental cement (Quick-Set2) on the viability and proliferation of human dental pulp stem cells (hDPSCs) by comparing the cellular responses with commercially available calcium silicate cement (white mineral trioxide aggregate; WMTA)...

  9. Proliferation inhibition and apoptosis induction of imatinib-resistant chronic myeloid leukemia cells via PPP2R5C down-regulation.

    Science.gov (United States)

    Shen, Qi; Liu, Sichu; Chen, Yu; Yang, Lijian; Chen, Shaohua; Wu, Xiuli; Li, Bo; Lu, Yuhong; Zhu, Kanger; Li, Yangqiu

    2013-01-01

    Despite the success of imatinib and other tyrosine kinase inhibitors (TKIs), chronic myeloid leukemia (CML) remains largely incurable, and a number of CML patients die due to Abl mutation-related drug resistance and blast crisis. The aim of this study was to evaluate proliferation inhibition and apoptosis induction by down-regulating PPP2R5C gene expression in the imatinib-sensitive and imatinib-resistant CML cell lines K562, K562R (imatinib resistant without an Abl gene mutation), 32D-Bcr-Abl WT (imatinib-sensitive murine CML cell line with a wild type Abl gene) and 32D-Bcr-Abl T315I (imatinib resistant with a T315I Abl gene mutation) and primary cells from CML patients by RNA interference. PPP2R5C siRNAs numbered 799 and 991 were obtained by chemosynthesis. Non-silencing siRNA scrambled control (SC)-treated, mock-transfected, and untreated cells were used as controls. The PPP2R5C mRNA and protein expression levels in treated CML cells were analyzed by quantitative real-time PCR and Western blotting, and in vitro cell proliferation was assayed with the cell counting kit-8 method. The morphology and percentage of apoptosis were revealed by Hoechst 33258 staining and flow cytometry (FCM). The results demonstrated that both siRNAs had the best silencing results after nucleofection in all four cell lines and primary cells. A reduction in PPP2R5C mRNA and protein levels was observed in the treated cells. The proliferation rate of the PPP2R5C-siRNA-treated CML cell lines was significantly decreased at 72 h, and apoptosis was significantly increased. Significantly higher proliferation inhibition and apoptosis induction were found in K562R cells treated with PPP2R5C-siRNA799 than K562 cells. In conclusion, the suppression of PPP2R5C by RNA interference could inhibit proliferation and effectively induce apoptosis in CML cells that were either imatinib sensitive or resistant. Down-regulating PPP2R5C gene expression might be considered as a new therapeutic target strategy

  10. 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.

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

    International Nuclear Information System (INIS)

    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

  12. The boiling water reactor BWR 90

    International Nuclear Information System (INIS)

    During the next decade a rise in the energy demand is expected worldwide, and this will in particular call for electricity generation capacity. A number of old generating plants, both nuclear and other plants, will probably have to be shut down for aging reasons, and their replacement will enhance the need for new generating capacity. The ABB Atom considers this situation to be met with a 'cautious evolution'. The offerings will largely be based on 'evolutions' of the successful light water reactor BWR 75. The new, evolutionary plant design of ABB Atom is the BWR 90. It can be designed, licensed and constructed in accordance with any safety regulations now in force or envisaged in the Western world. Emphasis has been, and will be, placed on features that facilitates licensing, shortens construction time and keeps electricity generation costs favourable. ABB also continues to develop a design of the 'passive' type, such as the 'passive' PIUS system, for possible deployment in the future. These efforts are more long-term activities, since development, verification and licensing of distinctly 'new' reactor concepts will have an extensive lead time. This paper presents the BWR 90 and its current status. The design is based on that of its forerunner, the BWR 75 standard design, taking into account the experiences gained from design and engineering, construction, commissioning, and operation of BWR 75 plants, the needs for adapting to new technologies and new safety requirements, as well as possibilities for simplifications and cost savings. (author) 4 figs

  13. ARES - a new BWR simulator

    International Nuclear Information System (INIS)

    Coupling the three-dimensional Analytic Function Expansion Nodal (AFEN) nodal model developed within the READY project in 2000-2001 with a four-equation based stationary-state thermalhydraulics module and a new cross section model, a basis has been created for a sophisticated BWR simulator code. Motivation for distilling the efforts into a new simulator, named ARES (AFEN Reactor Simulator), can be summarized in three main points: Stationary-state analyses required by the safety authorities must be independent from the calculations made by the power utilities. Using a different simulator for some calculations is an effective method for obtaining independent results; In order to keep up with the development in the core analysis field, a 'test bench' is required for testing and evaluating new ideas and models. In addition, accuracy of the commercial codes and the models incorporated in them can be evaluated by benchmarking them against the new simulator; and Writing a new program from scratch is potentially a good way to transfer experience from the first generation of Finnish nuclear engineers. It also gives the opportunity to re-evaluate some of the ideas used in the older codes written in times of significantly smaller computer capacity. (orig.)

  14. Thermal-hydraulics in BWR

    International Nuclear Information System (INIS)

    In the heat transferring flow in BWRs, the heightening of heat transfer performance accompanying the development of new fuel for the purpose of reducing spent fuel generation and the improvement of fuel economy, the heightening of performance and the reduction of size of various heat exchangers, the development of the safety devices, of which the constitution is simple, the reliability is high, and the operation is easy, and so on are expected. As for ABWRs, thermal output is 3926 MW, and electricity output is 1356 MW. The system constitution of ABWR is shown. The main change from BWR to ABWR is the adoption of internal pumps, reinforced concrete containment vessels and electric control rod drive. For evaluating the limit output of high burnup fuel assemblies, the subchannel analysis and the effect that spacers exert to the limit output are explained. The heat transferring flow in moisture separation heater, condenser and feed water heater is reported. The heat transferring flow in passive containment vessel cooling system of water wall type and condensing type is described. (K.I.)

  15. Improvement for BWR operator training

    International Nuclear Information System (INIS)

    BWR Operator Training Center was founded in April, 1971, and in April, 1974, training was begun, since then, 13 years elapsed. During this period, the curriculum and training facilities were strengthened to meet the training needs, and the new training techniques from different viewpoint were developed, thus the improvement of training has been done. In this report, a number of the training techniques which have been developed and adopted recently, and are effective for the improvement of the knowledge and skill of operators are described. Recently Japanese nuclear power stations have been operated at stable high capacity factor, accordingly the chance of experiencing the occurrence of abnormality and the usual start and stop of plants decreased, and the training of operators using simulators becomes more important. The basic concept on training is explained. In the standard training course and the short period fundamental course, the development of the guide for reviewing lessons, the utilization of VTRs and the development of the techniques for diagnosing individual degree of learning were carried out. The problems, the points of improvement and the results of these are reported. (Kako, I.)

  16. BWR pipe crack remedies evaluation

    International Nuclear Information System (INIS)

    This paper presents results on: (a) the influence of simulated BWR environments on the stress-corrosion-craking (SCC) susceptibility of Types 304, 316NG, and 347 stainless (SS); (b) fracture-mechanics crack-growth-rate measurements on these materials and weld overlay specimens in different environments; and (c) residual stress measurements and metallographic evaluations of conventional pipe weldments treated by a mechanical-stress-improvement process (MSIP) as well as those produced by a narrow-gap welding procedure. Crack initiation studies on Types 304 and 316NG SS under crevice and non-crevice conditions in 2890C water containing 0.25 ppM dissolved oxygen with low sulfate concentrations indicate that SCC initiates at very low strains (0 in both directions, and then grew at high rate (parallel to the nominal applied load). Residual stress measurements on MSIP-treated weldments and those produced by a narrow-gap welding procedure indicate that these techniques produce compressive stresses over most of the inner surface near the weld and heat-affected zones

  17. Full system decontamination experience in BWR

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, N.; Sugai, K.; Katayouse, N.; Fujimori, A.; Iida, K.; Hayashi, K. [Tokyo Electric Power Company, Tokyo (Japan); Kanasaki, T.; Inami, I. [Toshiba Corporation, Yokohama (Japan); Strohmer, F. [Framatome ANP Gmbh, Eelangen (Germany)

    2002-07-01

    At the Fukushima Daiichi Nuclear Power Station unit 3, unit 2, unit 5 and unit 1 of Tokyo Electric Power Company (TEPCO), the replacement of the core shroud and internals has been conducted since 1997 in this order. The welded core internals in operating BWR plants were replaced to improve stress corrosion cracking (SCC) resistance. At present these units are operating smoothly. The developed technology concept is to restore those internals in open air inside the reactor pressure vessel (RPV). To reduce the radiation dose rate inside the RPV, not only a shielding method was applied to cut the radiation from the irradiated structures but also a chemical decontamination method was applied to dissolve the radioactive crud deposited on the surface by using chemical agents. The calculated decontamination factor (DF) at the RPV bottom reached 35-117. As result, the dose rate decreased to approximately 0.1 mSv/h under water. Before and after the installation of the in-vessel shielding, a mechanical cleaning was extensively applied inside the RPV to remove the residual crud as well as the swarf, chips from cutting. As a result, the dose rate at the RPV bottom decreased to ranging from 0.2 to 0.4 mSv/h in air. A working environment for human access, which was better than expected, was established inside the RPV, resulting in 70, 140, 50 and 70 man-Sv (estimated) saving respectively at unit 3 (1F-3), unit 2(1F-2), unit 5(1F-5) and unit 1(1F-1). All four full system decontamination (FSDs) contributed to the successful realization of the core shroud replacement project under the dry condition in RPV.

  18. Effects of varied interferons in combination with all-trans retinoic acid (ATRA) on proliferation and differentiation of ATRA-resistent APL cell

    Institute of Scientific and Technical Information of China (English)

    HE Peng-cheng; ZHANG Mei; LI Jing; CAI Rui-bo; LIU Ya-lin; CAO Yun-xin

    2006-01-01

    Objective:To investigate the effects and mechanisms of interferon in combination with alltrans retinoic acid (ATRA) on proliferation and differentiation of ATRA-resistent APL cell. Methods :After MR2 cells (ATRA-resistance cell line) were treated with IFN-α, IFN-γ and ATRA alone or IFN-α and IFN-γ in combination with ATRA respectively. The cell roliferation was tested by MTT test and the cell differentiation was tested through light microscope by NBT test and flow cytometry (FCM). The expression of promyelocytic leukemia (PML) protein was observed by indirect immune fluorescent method. Results: Both IFN-α and IFN-γ could inhibit the proliferation and induce the differentiation of MR2 cells to some extent. The effects were more obvious after both interferons in combination with ATRA respectively (P<0. 05). Moreover, the maturation of MR2 cells induced by IFN-γ+ATRA group was more higher than that by IFN-α+ATRA group (P<0.05). Both interferons could induce the expressions of PML protein. Conclusion :Both interferons can inhibit MR2 cells proliferation, which may be related to the expression of PML protein induced by both interferons. The inducing differentiation effects of IFN-γ+ATRA group on MR2 cells are more powerful than those of IFN-α+ATRA group, which may be related to the different signal transduction pathway of both interferons.

  19. A BWR fuel channel tracking system

    International Nuclear Information System (INIS)

    A relational database management system with a query language, Reference 1, has been used to develop a Boiling Water Reactor (BWR) fuel channel tracking system on a microcomputer. The software system developed implements channel vendor and Nuclear Regulatory Commission recommendations for in-core channel movements between reactor operating cycles. A BWR Fuel channel encloses the fuel bundle and is typically fabricated using Ziracoly-4. The channel serves three functions: (1) it provides a barrier to separate two parallel flow paths, one inside the fuel assembly and the other in the bypass region outside the fuel assembly and between channels; (2) it guides the control rod as it moves between fuel assemblies and provides a bearing surface for the blades; and (3) it provides rigidity for the fuel bundle. All of these functions are necessary in typical BWR core designs. Fuel channels are not part of typical Pressurized Water Reactor (PWR) core designs

  20. Protecting AREVA ATRIUM™ BWR fuel from debris fretting failure

    International Nuclear Information System (INIS)

    Historically, debris fretting has been the leading cause of fuel rod failure in BWR fuel assemblies, costing the industry millions of dollars in lost generation and negatively impacting the working area of plant site personnel. In this paper the focus will be on recent BWR fuel product innovation designed to eliminate debris related failures. Experience feedback from more than three decades of operation history with non-line-of-sight FUELGUARD™ lower tie plate debris filters will be presented. The development and relative effectiveness of successive generations of filtration technology will be discussed. It will be shown that modern, state of the art debris filters are an effective defense against debris fretting failure. Protective measures extend beyond inlet nozzle debris filters. The comprehensive debris resistance features built into AREVA’s newest fuel design, the ATRIUM™ 11, reduce the overall risk of debris entrapment as well as providing a degree of protection from debris that may fall down on the fuel assembly from above, e.g., during refueling operations. The positive recent experience in a debris sensitive plant will be discussed showing that the combination of advanced fuel technology and a robust foreign material exclusion program at the reactor site can eliminate the debris fretting failure mechanism. (author)

  1. Efforts for optimization of BWR core internals replacement

    International Nuclear Information System (INIS)

    The core internal components replacement of a BWR was successfully completed at Fukushima-Daiichi Unit 3 (1F3) of the Tokyo Electric Power Company (TEPCO) in 1998. The core shroud and the majority of the internal components made by type 304 stainless steel (SS) were replaced with the ones made of low carbon type 316L SS to improve Intergranular Stress Corrosion Cracking (IGSCC) resistance. Although this core internals replacement project was completed, several factors combined to result in a longer-than-expected period for the outage. It was partly because the removal work of the internal components was delayed. Learning a lesson from whole experience in this project, some methods were adopted for the next replacement project at Fukushima-Daiichi Unit 2 (1F2) to shorten the outage and reduce the total radiation exposure. Those are new removal processes and new welding machine and so on. The core internals replacement work was ended at 1F2 in 1999, and both the period of outage and the total radiation exposure were the same degree as expected previous to starting of this project. This result shows that the methods adopted in this project are basically applicable for the core internals replacement work and the whole works about the BWR core internals replacement were optimized. The outline of the core internals replacement project and applied technologies at 1F3 and 1F2 are discussed in this paper. (author)

  2. General Electric's training program for BWR chemists

    International Nuclear Information System (INIS)

    This paper describes the development and implementation of the General Electric boiling water reactor chemistry training program from 1959 to the present. The original intention of this program was to provide practical hands on type training in radiochemistry to BWR chemistry supervisors with fossil station experience. This emphasis on radiochemistry has not changed through the years, but the training has expanded to include the high purity water chemistry of the BWR and has been modified to include new commission requirements, engineering developments and advanced instrumentation. Student and instructor qualifications are discussed and a description of the spin off courses for chemistry technicians and refresher training is presented

  3. Cutting BWR feedwater crud levels further

    International Nuclear Information System (INIS)

    Reducing iron input to the BWR primary system is an important first step in decreasing radiation fields and occupational exposure. For feedwater iron, the specified optimum concentration is 0.1-0.5 ppb, as demonstrated by the newest ''low crud'' plants operating in Japan. Recent advances in condensate filtration will achieve the levels needed for optimised water chemistry and promise great benefits. Of particular interest is a newly developed filterdemineraliser septum that separates the filtration and ion exchange functions to allow each to be specifically optimised for BWR conditions. (author)

  4. BWR pipe crack remedies evaluation

    International Nuclear Information System (INIS)

    This paper presents results on: (a) the influence of simulated BWR environments on the stress-corrosion-cracking (SCC) susceptibility of types 304, 316 NG, and 347 stainless steel (SS), (b) fracture-mechanics crack growth rate measurements on these materials and weld overlay specimens in different environments, and (c) residual stress measurements and metallographic evaluations of conventional pipe weldments treated by a mechanical-stress-improvement process (MSIP) as well as those produced by a narrow-gap welding procedure. Crack initiation studies on types 304 and 316 NG SS under crevice and non-crevice conditions in 2890C water containing 0.25 ppm dissolved oxygen with low sulfate concentrations indicate that SCC initiates at low strains (3%) in the nuclear grade material. Crack growth measurements on fracture-mechanics-type specimens, under low-frequency cyclic loading, show that the type 316 NG steel cracks at a somewhat lower rate (≅ 40%) than sensitized type 304 SS in an impurity environment with 0.25 ppm dissolved oxygen; however, the latter material stops cracking when sulfate is removed from the water. Crack growth in both materials ceases under simulated hydrogen-water chemistry conditions (6 ppb oxygen) even with 100 ppb sulfate present in the water. An unexpected results was obtained in the test on a weld overlay specimen in the impurity environment, viz., the crack grew to the overlay interface at a nominal rate, branched at 900 in both directions, and then grew at a high rate (parallel to the nominal applied load). Residual stress measurements on MSIP-treated weldments and those produced by a narrow-gap welding procedure indicate that these techniques produce compressive stresses over most of the inner surface near the weld and heat-affected zones. (orig.)

  5. 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

  6. Validation of BWR advanced core and fuel nuclear designs with power reactor measurements

    International Nuclear Information System (INIS)

    Power reactor measurements have been important in validating the reliability, performance characteristics and economics of BWR advanced core and fuel designs. Such measurements go beyond the data obtainable from normal reactor operation and provide detailed benchmark data necessary to verify design and licensing computer design and simulation models. In some cases, such as in the validation of the performance of zirconium barrier pellet-cladding-interaction (PCI) resistant cladding, the BWR power reactor measurements have subjected the advanced fuel design to operating conditions more severe than normal operating conditions, thereby providing nuclear-thermal-mechanical-corrosion performance data for accelerated or extended conditions of operation. In some cases destructive measurements have been carried out on BWR power reactor fuel to provide microscopic and macroscopic data of importance in validating design and licensing analysis methods. There is not uniform agreement among core and fuel designers on the needs for special power reactor core and fuel measurements for validation of advanced designs. The General Electric approach has been to error on the side of extensive, detailed measurements so as to assure reliable performance licensing and economic design and predictive capability. This paper is a summary of some of the validative power reactor measurements that have been carried out on advanced BWR core and fuel designs. Some comparisons of predictions with the data are summarized

  7. A User requirement2 (UR2) Evaluation of Proliferation Resistance of a 600MWe TRU Burner using ORIGEN2.1

    International Nuclear Information System (INIS)

    Proliferation Resistance (PR) should be assessed during design, operation of innovative nuclear energy system (INS) and management of spent fuel. For developing a TRU burner design an evaluation of PR is necessary to quantitatively evaluate a degree of PR and possibility of misuse and diversion of nuclear material (NM). In this paper, PR is evaluated based on the assessment methodology of the INPRO user manual against a 600MWe class TRU burner. Furthermore, its result is compared with the result of PR assessment of PWR and CANDU. The attractiveness of nuclear material is also assessed by using the ORIGEN2.1 code

  8. Feasibility study on development of plate-type heat exchanger for BWR plants

    International Nuclear Information System (INIS)

    In order to apply plate-type heat exchanger to RCW, TCW and FPC system in BWR plants, heat test and seismic test of RCW system heat exchanger sample were carried out. The results of these tests showed new design plate-type heat exchanger satisfied the fixed pressure resistance and seismic resistance and keep the function. The evaluation method of seismic design was constructed and confirmed by the results of tests. As anti-adhesion measure of marine organism, an ozone-water circulation method, chemical-feed method and combination of circulation of hot water and air bubbling are useful in place of the chlorine feeding method. Application of the plate-type heat exchanger to BWR plant is confirmed by these investigations. The basic principles, structure, characteristics, application limit and reliability are stated. (S.Y.)

  9. 44-BWR WASTE PACKAGE LOADING CURVE EVALUATION

    International Nuclear Information System (INIS)

    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

  10. Osteoblast-derived sphingosine 1-phosphate to induce proliferation and confer resistance to therapeutics to bone metastasis-derived prostate cancer cells.

    Science.gov (United States)

    Brizuela, Leyre; Martin, Claire; Jeannot, Pauline; Ader, Isabelle; Gstalder, Cécile; Andrieu, Guillaume; Bocquet, Magalie; Laffosse, Jean-Michel; Gomez-Brouchet, Anne; Malavaud, Bernard; Sabbadini, Roger A; Cuvillier, Olivier

    2014-10-01

    Sphingosine 1-phosphate (S1P) plays important roles in cell proliferation, differentiation or survival mainly through its surface G-protein-coupled receptors S1P1-5. Bone represents the major site of metastasis for prostate cancer (CaP) cells, which rely on bone-derived factors to support their proliferation and resistance to therapeutics. In the present work we have found that conditioned medium (CM) from the MC3T3 osteoblastic cell line or primary murine and human osteoblast-like cells, as well as co-culture with MC3T3 stimulate proliferation of CaP lines in S1P-dependent manner. In addition, osteoblastic-derived S1P induces resistance of CaP cells to therapeutics including chemotherapy and radiotherapy. When S1P release from osteoblastic cells is decreased (inhibition of SphK1, knock-down of SphK1 or the S1P transporter, Spns2 by siRNA) or secreted S1P neutralized with anti-S1P antibody, the proliferative and survival effects of osteoblasts on CaP cells are abolished. Because of the paracrine nature of the signaling, we studied the role of the S1P receptors expressed on CaP cells in the communication with S1P secreted by osteoblasts. Strategies aimed at down-regulating S1P1, S1P2 or S1P3 (siRNA, antagonists), established the exclusive role of the S1P/S1P1 signaling between osteoblasts and CaP cells. Bone metastases from CaP are associated with osteoblastic differentiation resulting in abnormal bone formation. We show that the autocrine S1P/S1P3 signaling is central during differentiation to mature osteoblasts by regulating Runx2 level, a key transcription factor involved in osteoblastic maturation. Importantly, differentiated osteoblasts exhibited enhanced secretion of S1P and further stimulated CaP cell proliferation in a S1P-dependent manner. By establishing the dual role of osteoblast-borne S1P on both osteoblastic differentiation and CaP cell proliferation and survival, we uncover the importance of S1P in the bone metastatic microenvironment, which may open

  11. Proliferation resistance and energy security advantages of a thorium-uranium dioxide once-through fuel cycle for light water reactors

    International Nuclear Information System (INIS)

    This study analyzes whether spent light reactor (LWR) thorium-uranium dioxide fuel poses a significantly lower risk for nuclear weapon proliferation than spent uranium-dioxide fuel, based on the isotopic composition of the contained uranium and plutonium. Mixed Th/U fuel with an initial enrichment of 19.5% U235 can achieve an average burnup of 70,000 MWd/tHM in a PWR using 30% UO2 and 70% ThO2. To get the equivalent burnup, LEU fuel requires an initial enrichment of 8.0% U235. Two computer codes, MCNP and ORIGEN2, are used to perform the depletion calculation. The spent mixed thorium-uranium dioxide fuel discharged from a pressurized-water reactor has a plutonium isotopic composition and higher decay heat production per kilogram of plutonium more proliferation resistant than spent low enriched uranium dioxide fuel, while significantly reducing the quantity of plutonium produced. The U233 + U235 mixture in spent thorium-uranium fuel is low enriched and contaminated with gamma-emitting U232. With respect to energy security, the introduction of a thorium-uranium fuel cycle could reduce concern over uranium fuel supply of a resource-poor nation since thorium reserve is much larger, compared to fuel cycles using 4.5% LEU, while its uranium saving is almost equivalent to plutonium recycling. Overall, spent thorium-uranium fuel appears significantly more proliferation resistant in terms of the weapons-usability of the contained fissile material than spent low enriched uranium fuel, although use of 19.5% enriched uranium in fresh fuel would facilitate production of weapons-grade uranium at a higher rate in countries with clandestine enrichment facilities. (S.Y.)

  12. Combination of PIM and JAK2 inhibitors synergistically suppresses cell proliferation and overcomes drug resistance of myeloproliferative neoplasms

    OpenAIRE

    Huang, Shih-Min A.; Wang, Anlai; Greco, Rita; LI, Mrs Zhifang; Sun, Fangxian; Barberis, Claude; Tabart, Michel; Patel, Vinod; Schio, Laurent; Hurley, Raelene; Chen, Bo; Cheng, Hong; Lengauer, Christoph; Pollard, Jack; Watters, James

    2014-01-01

    Inhibitors of JAK2 kinase are emerging as an important treatment modality for myeloproliferative neoplasms (MPN). However, similar to other kinase inhibitors, resistance to JAK2 inhibitors may eventually emerge through a variety of mechanisms. Effective drug combination is one way to enhance therapeutic efficacy and combat resistance against JAK2 inhibitors. To identify potential combination partners for JAK2 compounds in MPN cell lines, we performed pooled shRNA screen targeting 5,000 genes ...

  13. 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 BACKGROUND: 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. METHODOLOGY AND PRINCIPAL FINDINGS: 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. CONCLUSION/SIGNIFICANCE: 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.

  14. Process inherent ultimate safety/boiling-water reactor PIUS/BWR

    International Nuclear Information System (INIS)

    This document is a series of viewgraphs on: design basis of PIUS/BWR, definition of PIUS/BWR, mechanisms of safe shutdown and afterheat cooling, advantages of PIUS/BWR, and research and development requirements

  15. BWR NSSS design basis documentation

    International Nuclear Information System (INIS)

    programs that GE has participated in and describes the different options and approaches that have been used by various utilities in their design basis programs. Some of these variations deal with the scope and depth of coverage of the information, while others are related to the process (how the work is done). Both of these topics can have a significant effect on the program cost. Some insight into these effects is provided. The final section of the paper presents a set of lessons learned and a recommendation for an optimum approach to a design basis information program. The lessons learned reflect the knowledge that GE has gained by participating in design basis programs with nineteen domestic and international BWR owner/operators. The optimum approach described in this paper is GE's attempt to define a set of information and a work process for a utility/GE NSSS Design Basis Information program that will maximize the cost effectiveness of the program for the utility. (

  16. EPRI BWR Water Chemistry Guidelines Revision

    International Nuclear Information System (INIS)

    BWRVIP-190: BWR Water Chemistry Guidelines – 2008 Revision has been revised. The revision committee consisted of U.S. and non-U.S. utilities (members of the BWR Vessel and Internals Protection (BWRVIP) Mitigation Committee), reactor system manufacturers, fuel suppliers, and EPRI and industry experts. The revised document, BWRVIP-190 Revision 1, was completely reformatted into two volumes, with a simplified presentation of water chemistry control, diagnostic and good practice parameters in Volume 1 and the technical bases in Volume 2, to facilitate use. The revision was developed in parallel and in coordination with preparation of the Fuel Reliability Guidelines Revision 1: BWR Fuel Cladding Crud and Corrosion. Guidance is included for plants operating under normal water chemistry (NWC), moderate hydrogen water chemistry (HWC-M), and noble metal application (GE-Hitachi NobleChem™) plus hydrogen injection. Volume 1 includes significant changes to BWR feedwater and reactor water chemistry control parameters to provide increased assurance of intergranular stress corrosion cracking (IGSCC) mitigation of reactor materials and fuel reliability during all plant conditions, including cold shutdown (≤200°F (93°C)), startup/hot standby (>200°F (93°C) and ≤ 10%) and power operation (>10% power). Action Level values for chloride and sulfate have been tightened to minimize environmentally assisted cracking (EAC) of all wetted surfaces, including those not protected by hydrogen injection, with or without noble metals. Chemistry control guidance has been enhanced to minimize shutdown radiation fields by clarifying targets for depleted zinc oxide (DZO) injection while meeting requirements for fuel reliability. Improved tabular presentations of parameter values explicitly indicate levels at which actions are to be taken and required sampling frequencies. Volume 2 provides the technical bases for BWR water chemistry control for control of EAC, flow accelerated corrosion

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

    International Nuclear Information System (INIS)

    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

  18. 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. PMID:26825545

  19. 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.

  20. Improving proliferation resistance of high breeding gain generation 4 reactors using blankets composed of light water reactor waste

    Energy Technology Data Exchange (ETDEWEB)

    Hellesen, C.; Grape, S.; Haakanson, A.; Jacobson Svaerd, S.; Jansson, P. [Division of Applied Nuclear Physics, Uppsala University, Aangstroemlaboratoriet Laegerhyddsvaegen 1, 751 20 Uppsala (Sweden)

    2013-07-01

    Fertile blankets can be used in fast reactors to enhance the breeding gain as well as the passive safety characteristics. However, such blankets typically result in the production of weapons grade plutonium. For this reason they are often excluded from Generation IV reactor designs. In this paper we demonstrate that using blankets manufactured directly from spent light water (LWR) reactor fuel it is possible to produce a plutonium product with non-proliferation characteristics on a par with spent LWR fuel of 30-50 MWd/kg burnup. The beneficial breeding and safety characteristics are retained. (authors)

  1. CANDLE reactor: an option for simple, safe, high nuclear proliferation resistant , small waste and efficient fuel use reactor

    International Nuclear Information System (INIS)

    The innovative nuclear energy systems have been investigated intensively for long period in COE-INES program and CRINES activities in Tokyo Institute of Technology. Five requirements; sustainability, safety, waste, nuclear-proliferation, and economy; are considered as inevitable requirements for nuclear energy. Characteristics of small LBE cooled CANDLE fast reactor developed in this Institute are discussed for these requirements. It satisfies clearly four requirements; safety, nonproliferation and safeguard, less wastes and sustainability. For the remaining requirement, economy, a high potential to satisfy this requirement is also shown

  2. Service experience of BWR pressure vessels

    International Nuclear Information System (INIS)

    The overall service experience with Boiling Water Reactor (BWR) pressure vessels has been excellent. The only significant factor that impacted the service performance has been thermal fatigue cracking of feedwater inlet nozzle. This concern has been mitigated by eliminating the source of thermal cycling stress through design and operational changes. Although stress corrosion cracking has occurred in early atypical steam generator vessel designs, analysis and field experience has indicated that this mechanism is not expected in the BWR reactor pressure vessel (RPV). Other limited materials related cracking problems have been associated with RPV stainless steel and nickel-base alloy attachments and penetrations. Solutions to these problems have involved design and materials modifications. Finally, due to the low end of life fluence resulting from the large core-to-RPV-wall water annulus, irradiation embrittlement effects are minimal

  3. Examination of overlay repaired BWR pipe joints

    International Nuclear Information System (INIS)

    Intergranular stress corrosion cracking (IGSCC) in a large number of austenitic stainless steel girth welds in boiling water reactor (BWR) piping has prompted the development of the weld overlay for repair (WOR) as a short-term remedy. It is necessary to examine the deposited overlay weld material for adequate definition of its condition and to monitor the overlaid IGSCC to determine if it grows past the bounds assumed in the design of the repair. This paper reports on NDE techniques evaluated using weld overlaid pipe samples containing known defects, overlaid samples removed from BWR service, and overlaid weld joints in plant. These samples included overlays containing fabrication defects and overlaid pipes containing deep and shallow laboratory- and service-induced IGSCC

  4. Crud deposition modeling on BWR fuel rods

    International Nuclear Information System (INIS)

    Deposition of boiling water reactor (BWR) system corrosion products (crud) on operating fuel rods has resulted in performance-limiting conditions in a number of plants. The operational impact of performance-limiting conditions involving crud deposition can be detrimental to a BWR operator, resulting in unplanned or increased frequency of fuel inspections, fuel failure and associated radiological consequences, operational restrictions including core power derate and/or forced shutdowns to remove failed fuel, premature discharge of individual bundles or entire reloads, and/or undesirable core design restrictions. To facilitate improved management of crud-related fuel performance risks, EPRI has developed the CORAL (Crud DepOsition Risk Assessment ModeL) tool. This paper presents a summary of the CORAL elements and benchmarking results. Applications of CORAL as a tool for fuel performance risk assessment are also discussed. (author)

  5. The design of large natural circulation BWR's

    International Nuclear Information System (INIS)

    Boiling water reactors (BWR) with natural circulation are applied for capacities up to 60 MWe. Based on scale studies, however, it appears that larger production units are more efficient. It is recommended to investigate the bottlenecks in realizing larger reactors (>1000 MWe). The aim of the study on the title subject is to study to what extent the production capacity of BWRs with natural circulation can be increased. Based on data from the literature a simple analytic method has been chosen and existing BWR designs were compared. Capacities of 1300 MWe appear to be possible. These reactors will have a smaller pin diameter and a lower water supply temperature. Also steam separators with a minor pressure reduction must be available. The reliability of the stability measurement must be increased. Based on the results of this investigation the priorities for research on the design of future BWRs have been determined

  6. BWR radiation buildup control with ionic zinc

    International Nuclear Information System (INIS)

    In 1983 a hypothesis was disclosed which suggested that the presence of ionic zinc in the reactor water of the BWR could reduce radiation buildup. This hypothesis was developed from correlations of plant data, and subsequently, from laboratory experiments which demonstrated clearly that ionic zinc inhibits the corrosion of stainless steel. The benefits of zinc addition have been measured at the Vallecitos Nuclear Center under and EPRI/GE project. Experimentation and analyses have been performed to evaluate the impact of intentional zinc addition on the IGSCC characteristics of primary system materials and on the performance of the nuclear fuel. It has been concluded that no negative effects are expected. The author conclude that the intentional addition of ionic zinc to the BWR reactor water at a concentration of approximately 10 ppb will provide major benefits in controlling the Co-60 buildup on primary system stainless steel surfaces. The intentional addition of zinc is now a qualified technique for use in BWRs

  7. Recent developments in BWR water chemistry

    International Nuclear Information System (INIS)

    Water chemistry is of critical importance to the operation and economic viability of the Boiling Water Reactor (BWR). A successful water chemistry program will satisfy the following goals: - Minimize the incidence and growth of SCC/IASCC, - Minimize plant radiation fields controllable by chemistry, -Maintain fuel integrity by minimizing cladding corrosion, - Minimize flow-accelerated corrosion (FAC) in balance-of-plant components. The impact of water chemistry on each of these goals is discussed in more detail in this paper. It should be noted that water chemistry programs also include surveillance and operating limits for other plant water systems (e.g., service water, closed cooling water systems, etc.) but these are out of the scope of this paper. This paper reviews developments in water chemistry guidelines for U.S. BWR nuclear power plants. (author). 2 figs., 2 tabs., 7 refs

  8. Parallel channel effects under BWR LOCA conditions

    International Nuclear Information System (INIS)

    Due to parallel channel effects, different flow patterns such as liquid down-flow and gas up-flow appear simultaneously in fuel bundles of a BWR core during postulated LOCAs. Applying the parallel channel effects to the fuel bundle, water drain tubes with a restricted bottom end have been developed in order to mitigate counter-current flow limiting and to increase the falling water flow rate at the upper tie plate. The upper tie plate with water drain tubes is an especially effective means of increasing the safety margin of a reactor with narrow gaps between fuel rods and high steam velocity at the upper tie plate. The characteristics of the water drain tubes have been experimentally investigated using a small-scaled steam-water system simulating a BWR core. Then, their effect on the fuel cladding temperature was evaluated using the LOCA analysis program SAFER. (orig.)

  9. EASY 5 BWR simulation model for digital feedwater control design

    International Nuclear Information System (INIS)

    The development of a BWR simulation model in support of a program to design and evaluate the digital feedwater control system for the Monticello Boiling Water Reactor (BWR) is described. This model was developed in the EASY5 simulation language in conjunction with EPRI's Modular Modeling System (MMS) two-phase Library. The model consists of three main elements: the BWR reactor vessel module, the feedwater system model, and the steamline model. Transient results for the BWR vessel module and the feedwater system model are presented

  10. 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. PMID:27397581

  11. Swedes repair BWR thermal fatigue cracks

    International Nuclear Information System (INIS)

    The discovery of cracks in the feedwater and shutdown cooling systems of Sweden's Barseback 2 BWR in 1980 led to investigations in other Swedish nuclear power stations. Similar cracks were found and the defective parts repaired or replaced before being returned to service. The cause of the cracks has been evaluated and efforts are being made to prevent a recurrence. Experience with Ringhals 1, Orkarsham 2 and Forsmark 1 systems are also described. (author)

  12. Assessment of two BWR accident management strategies

    International Nuclear Information System (INIS)

    Candidate mitigative strategies for the management of in-vessel events during the late phase (after-core degradation has occurred) of postulated boiling water reactor (BWR) severe accidents were considered at Oak Ridge National Laboratory (ORNL) during 1990. The identification of new strategies was subject to the constraint that they should, to the maximum extent possible, make use of the existing equipment and water resources of the BWR facilities, and not require major equipment modifications or additions. As a result of this effort, two of these candidate strategies were recommended for further assessment. The first was a strategy for containment flooding to maintain the core and structural debris within the reactor vessel in the event that vessel injection cannot be restored to terminate a severe accident sequence. The second strategy pertained to the opposite case, for which vessel injection would be restored after control blade melting had begun; its purpose was to provide an injection source of borated water at the concentration necessary to preclude criticality upon recovering a damaged BWR core. Assessments of these two strategies were performed during 1991 and this paper provides a discussion of the motivation for and purpose of these strategies, and the potential for their success. ((orig.))

  13. Advanced Neutronics Tools for BWR Design Calculations

    International Nuclear Information System (INIS)

    This paper summarizes the developments implemented in the new APOLLO2.8 neutronics tool to meet the required target accuracy in LWR applications, particularly void effects and pin-by-pin power map in BWRs. The Method Of Characteristics was developed to allow efficient LWR assembly calculations in 2D-exact heterogeneous geometry; resonant reaction calculation was improved by the optimized SHEM-281 group mesh, which avoids resonance self-shielding approximation below 23 eV, and the new space-dependent method for resonant mixture that accounts for resonance overlapping. Furthermore, a new library CEA2005, processed from JEFF3.1 evaluations involving feedback from Critical Experiments and LWR P.I.E, is used. The specific '2005-2007 BWR Plan' settled to demonstrate the validation/qualification of this neutronics tool is described. Some results from the validation process are presented: the comparison of APOLLO2.8 results to reference Monte Carlo TRIPOLI4 results on specific BWR benchmarks emphasizes the ability of the deterministic tool to calculate BWR assembly multiplication factor within 200 pcm accuracy for void fraction varying from 0 to 100%. The qualification process against the BASALA mock-up experiment stresses APOLLO2.8/CEA2005 performances: pin-by-pin power is always predicted within 2% accuracy, reactivity worth of B4C or Hf cruciform control blade, as well as Gd pins, is predicted within 1.2% accuracy. (authors)

  14. Analysis of BWR lattices to recycle americium

    International Nuclear Information System (INIS)

    This study was carried out to assess the ability to eliminate meaningful quantities of americium in a primarily thermal neutron flux by 'spiking' modern BWR fuel with this minor actinide (MA). The studies carried out so far include the simulation of modern 10 x 10 BWR lattices employing the Westinghouse lattice physics code PHOENIX-4 alongside validation studies using MCNP5 models of the same lattices that were spatially depleted via the MONTEBURNS code coupling to ORIGEN. When considering the total inventory of minor actinides in Am-spiked pins, excluding isotopes of uranium and plutonium, the results indicate that a reduction of approximately 50% or more in the total mass inventory of these minor actinides is viable within the selected pins. Therefore, these preliminary results have encouraged the extension of this work to the development of improved lattice designs to help optimize the transmutation rates as well as absolute MA inventory reductions. The ultimate goal being to design batches of these advanced BWR bundles alongside multi-cycle core reload strategies. (authors)

  15. BWR recirculation pump diagnostic expert system

    International Nuclear Information System (INIS)

    At General Electric (GE), an on-line expert system to support maintenance decisions for BWR recirculation pumps for nuclear power plants has been developed. This diagnostic expert system is an interactive on-line system that furnishes diagnostic information concerning BWR recirculation pump operational problems. It effectively provides the recirculation pump diagnostic expertise in the plant control room continuously 24 hours a day. The expert system is interfaced to an on-line monitoring system, which uses existing plant sensors to acquire non-safety related data in real time. The expert system correlates and evaluates process data and vibration data by applying expert rules to determine the condition of a BWR recirculation pump system by applying knowledge based rules. Any diagnosis will be automatically displayed, indicating which pump may have a problem, the category of the problem, and the degree of concern expressed by the validity index and color hierarchy. The rules incorporate the expert knowledge from various technical sources such as plant experience, engineering principles, and published reports. These rules are installed in IF-THEN formats and the resulting truth values are also expressed in fuzzy terms and a certainty factor called a validity index. This GE Recirculation Pump Expert System uses industry-standard software, hardware, and network access to provide flexible interfaces with other possible data acquisition systems. Gensym G2 Real-Time Expert System is used for the expert shell and provides the graphical user interface, knowledge base, and inference engine capabilities. (author)

  16. Scaling and uncertainty in BWR instability problems

    International Nuclear Information System (INIS)

    This paper deals with a critical review of activities, performed at the DCMN of Pisa University, in relation to the thermo-hydraulic oscillations in two-phase systems. Stability analyses, including model development and achievement of experimental data, are generally performed for BWRs in order to achieve the following objectives: to reach a common understanding in relation to the predictive capabilities of system codes and to the influence of various parameters on the instability; to establish a data base for the qualification of the analytical tools already or becoming available; to set-up qualified tools (code/models + nodalization + user assumption) suitable for predicting the unstable behaviour of the nuclear plants of interest (current BWR, SBWR, ABWR and RBMK). These considerations have been the basis for the following researches: 1) proposal of the Boiling Instability Program (BIP) (1) 2) evaluation of stability tests in PIPER-ONE apparatus (2) 3) coupled thermal-hydraulic and neutronic instabilities in the LaSalle-2 BWR plant (3) 4) participation to the NEA-OECD BWR Benchmark (4) The RELAP/MOD2 and RELAP5/MOD3 codes have been used. (author)

  17. Decontamination techniques for BWR power generation plant

    International Nuclear Information System (INIS)

    The present report describes various techniques used for decontamination in BWR power generation plants. Objectives and requirements for decontamination in BWR power plants are first discussed focusing on reduction in dose, prevention of spread of contamination, cleaning of work environments, exposure of equipment parts for inspection, re-use of decontaminated resources, and standards for decontamination. Then, the report outlines major physical, chemical and electrochemical decontamination techniques generally used in BWR power generation plants. The physical techniques include suction of deposits in tanks, jet cleaning, particle blast cleaning, ultrasonic cleaning, coating with special paints, and flushing cleaning. The chemical decontamination techniques include the use of organic acids etc. for dissolution of oxidized surface layers and treatment of secondary wastes such as liquids released from primary decontamination processes. Other techniques are used for removal of penetrated contaminants, and soft and hard cladding in and on equipment and piping that are in direct contact with radioactive materials used in nuclear power generation plants. (N.K.)

  18. Advanced neutronics tools for BWR design calculations

    International Nuclear Information System (INIS)

    This paper summarizes the developments implemented in the new APOLLO2.8 neutronics tool to meet the required target accuracy in LWR applications, particularly void effects and pin-by-pin power map in BWRs. The Method of Characteristics was developed to allow efficient LWR assembly calculations in 2D-exact heterogeneous geometry; resonant reaction calculation was improved by the optimized SHEM-281 group mesh, which avoids resonance self-shielding approximation below 23 eV, and the new space-dependent method for resonant mixture that accounts for resonance overlapping. Furthermore, a new library CEA2005, processed from JEFF3.1 evaluations involving feedback from Critical Experiments and LWR P.I.E, is used. The specific '2005-2007 BWR Plan' settled to demonstrate the validation/qualification of this neutronics tool is described. Some results from the validation process are presented: the comparison of APOLLO2.8 results to reference Monte Carlo TRIPOLI4 results on specific BWR benchmarks emphasizes the ability of the deterministic tool to calculate BWR assembly multiplication factor within 200 pcm accuracy for void fraction varying from 0 to 100%. The qualification process against the BASALA mock-up experiment stresses APOLLO2.8/CEA2005 performances: pin-by-pin power is always predicted within 2% accuracy, reactivity worth of B4C or Hf cruciform control blade, as well as Gd pins, is predicted within 1.2% accuracy

  19. Reliability innovations for AREVA NP BWR fuel

    International Nuclear Information System (INIS)

    AREVA NP is a supplier of nuclear fuel assemblies and associated core components to light water reactors worldwide, representing today more than 185,000 fuel assemblies on the world market including more than 63,000 fuel assemblies for boiling water reactors (BWRs). ATRIUM trademark 10 fuel assemblies have been supplied to a total of 32 BWR plants worldwide resulting in an operating experience over 20,250 fuel assemblies. ATRIUM trademark 10XP and ATRIUM trademark 10XM are AREVA NP's most recent fuel assembly designs featuring improved fuel utilization and achieving high margins to operating limits while maintaining very good reliability. Nevertheless, fuel failures are still encountered in all modern and advanced fuel assembly designs leading to significant operating limitations or unplanned shutdowns of nuclear power plants. The majority of fuel failures in BWR plants are caused by debris fretting, with PCI induced failures being a second leading cause. AREVA NP runs programs to study these root causes and to develop product solutions as part of the continuous improvement process within the Zero Tolerance for Failure (ZTF) initiative. The focus of the ZTF initiative is to further upgrade BWR fuel assembly reliability to achieve the goal of failure free fuel. In the following, two major product improvements are described that will significantly contribute to this goal: - Improved FUELGUARD trademark Lower Tie Plate - Chamfered Fuel Pellet Design (orig.)

  20. Optimization of Heterogeneous Fuel Designs for Utilization of Thorium In PWRs To Enhance Proliferation Resistance and Reduce Waste

    International Nuclear Information System (INIS)

    This paper presents a summary of the first stage of the project aimed to examine heterogeneous core design options for the implementation of the thorium-233U fuel cycle in pressurized water reactors (PWRs) and to identify the core design and fuel management strategies that will maximize the benefits from inclusion of thorium in the fuel. The project is carried out within a framework of Nuclear Energy Research Initiative (NERI) supported by the US Department of Energy (1). Principal investigators are M. Todosow from Brookhaven National Laboratory and M. Kazimi from Massachusetts Institute of Technology with contributions from Kurchatov Institute (Russia) and Ben-Gurion University of the Negev (Israel). The fuel cycle assessment concentrates on key measures of performance in several important areas including proliferation characteristics of the spent fuel, reliability, safety, cost, environmental impact, and licensing issues

  1. 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.

  2. 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.

  3. 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.

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

    International Nuclear Information System (INIS)

    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

  5. TARMS2, New-generation BWR core management system

    International Nuclear Information System (INIS)

    Toshiba has developed a prototype of a new-generation core management system for the boiling water reactor (BWR) TARMS2. It contains function modules for core monitoring and core prediction analysis. TARMS2 is equipped with an advanced three-dimensional BWR core physics model, LOGOS

  6. BWR Full Integral Simulation Test (FIST) Phase II test results and TRAC-BWR model qualification

    International Nuclear Information System (INIS)

    Eight matrix tests were conducted in the FIST Phase I. These tests investigated the large break, small break and steamline break LOCA's, as well as natural circulation and power transients. There are nine tests in Phase II of the FIST program. They include the following LOCA tests: BWR/6 LPCI line break, BWR/6 intermediate size recirculation break, and a BWR/4 large break. Steady state natural circulation tests with feedwater makeup performed at high and low pressure, and at high pressure with HPCS makeup, are included. Simulation of a transient without rod insertion, and with controlled depressurization, was performed. Also included is a simulation of the Peach Bottom turbine trip test. The final two tests simulated a failure to maintain water level during a postulated accident. A FIST program objective is to assess the TRAC code by comparisons with test data. Two post-test predictions made with TRACB04 are compared with Phase II test data in this report. These are for the BWR/6 LPCI line break LOCA, and the Peach Bottom turbine trip test simulation

  7. 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.

  8. High CHMP4B expression is associated with accelerated cell proliferation and resistance to doxorubicin in hepatocellular carcinoma.

    Science.gov (United States)

    Hu, Baoying; Jiang, Dawei; Chen, Yuyan; Wei, Lixian; Zhang, Shusen; Zhao, Fengbo; Ni, Runzhou; Lu, Cuihua; Wan, Chunhua

    2015-04-01

    Charged multivesicular body protein 4B (CHMP4B), a subunit of the endosomal sorting complex required for transport (ESCRT)-III complex, plays an important part in cytokinetic membrane abscission and the late stage of mitotic cell division. In this study, we explored the prognostic significance of CHMP4B in human hepatocellular carcinoma (HCC) and its impact on the physiology of HCC cells. Western blot and immunohistochemistrical analyses showed that CHMP4B was significantly upregulated in HCC tissues, compared with adjacent non-tumorous tissues. Meanwhile, clinicopathological analysis revealed that high CHMP4B expression was correlated with multiple clinicopathological variables, including AFP, cirrhosis, AJCC stage, Ki-67 expression, and poor prognosis. More importantly, univariate and multivariate survival analyses demonstrated that CHMP4B served as an independent prognostic factor for survival of HCC patients. Using HCC cell cultures, we found that the expression of CHMP4B was progressively upregulated after the release from serum starvation. To verify whether CHMP4B could regulate the proliferation of HCC cells, CHMP4B was knocked down through the transfection of CHMP4B-siRNA oligos. Flow cytometry and CCK-8 assays indicated that interference of CHMP4B led to cell cycle arrest and proliferative impairment of HCC cells. Additionally, depletion of CHMP4B expression could increase the sensitivity to doxorubicin in HepG2 and Huh7 cells. Taken together, our results implied that CHMP4B could be a promising prognostic biomarker as well as a potential therapeutic target of HCC. PMID:25874485

  9. Investigation of burnup credit implementation for BWR fuel

    International Nuclear Information System (INIS)

    Burnup Credit allows considering the reactivity decrease due to fuel irradiation in criticality studies for the nuclear fuel cycle. Its implementation requires to carefully analyze the validity of the assumptions made to define the axial profile of the burnup and void fraction (for BWR), to determine the composition of the irradiated fuel and to compute the criticality simulation. In the framework of Burnup Credit implementation for BWR fuel, this paper proposes to investigate part of these items. The studies presented in this paper concern: the influence of the burnup and of the void fraction on BWR spent fuel content and on the effective multiplication factor of an infinite array of BWR assemblies. A code-to-code comparison for BWR fuel depletion calculations relevant to Burnup Credit is also performed. (authors)

  10. Local instability in BWR reactor simulator

    International Nuclear Information System (INIS)

    ''Local'' oscillations in thermal-hydraulic quantities have been observed while performing system code calculations related to PIPER-ONE BWR simulator; void fraction oscillations were detected at the channel middle elevation but not at channel bottom and top. This paper deals with an experimental investigation of ''local'' instability starting from experimental data obtained with the PIPER-ONE facility. The considered test is PO-SD-5B; the post-test analysis has been carried out by RELAP5/MOD2 code. In particular the analysis concerns the experimental and calculated oscillations of pressure drop in the core region and the calculated oscillations of channel void fraction

  11. Level 2 PRA for a German BWR

    International Nuclear Information System (INIS)

    A concept for a Level 2 Probabilistic Risk Assessment (L2 PRA) for a German Boiling Water Reactor (BWR) has been developed taking into account the role of L2 PRA within the German regulatory landscape. According to this concept, a plant specific evaluation of the severe accident phenomenology as well as analyses of the accident progression for the severe accident scenarios has been performed. Furthermore a plant specific MELCOR 1.8.6 model has been developed and special MELCOR source term calculations have been performed for the different release paths. This paper will present examples from the different areas described above. (author)

  12. Development of ECP models for BWR applications

    International Nuclear Information System (INIS)

    The electrochemical corrosion potential (ECP) of stainless steel has been measured under simulated Boiling Water Reactor (BWR) coolant circuit conditions using a rotating cylinder electrode. Based on the results of measurements an empirical model has been developed to predict the ECP of structure materials in a BVTR primary circuit as a function of H2, O2, and H2O2 concentrations in reactor coolant and water flow velocity. The ECP modeling results using the H2, O2, and H2O2 concentrations calculated by the radiolysis model are compared with the available reactor internal ECP data obtained in an operating reactor

  13. Fracture assessment of a BWR pump nozzle

    International Nuclear Information System (INIS)

    Fracture mechanics calculations are performed to support the non-destructive testing (NDT) qualification programs for pump nozzle investigations of boiling water reactor (BWR) nozzles of reactor pressure vessels (RPVs), with the aim of the determination of qualification defects, which are located in the Inconel 182 weld of the pump nozzle at the bottom of the RPV. The ferritic nozzle and housing have an Inconel buttering and each part is cladded with Inconel 182 before it is mounted. All theses weldments are heat treated after welding; only the connecting weldment between pump housing and nozzle, which is also an Inconel 182 weld, performed on site, is in the as welded condition. (author)

  14. A BWR licensing experience in the USA

    International Nuclear Information System (INIS)

    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)

  15. CECP, Decommissioning Costs for PWR and BWR

    International Nuclear Information System (INIS)

    1 - Description of program or function: The Cost Estimating Computer Program CECP, designed for use on an IBM personal computer or equivalent, was developed for estimating the cost of decommissioning boiling water reactor (BWR) and light-water reactor (PWR) power stations to the point of license termination. 2 - Method of solution: Cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial volume and costs; and manpower staffing costs. Using equipment and consumables costs and inventory data supplied by the user, CECP calculates unit cost factors and then combines these factors with transportation and burial cost algorithms to produce a complete report of decommissioning costs. In addition to costs, CECP also calculates person-hours, crew-hours, and exposure person-hours associated with decommissioning. 3 - Restrictions on the complexity of the problem: The program is designed for a specific waste charge structure. The waste cost data structure cannot handle intermediate waste handlers or changes in the charge rate structures. The decommissioning of a reactor can be divided into 5 periods. 200 different items for special equipment costs are possible. The maximum amount for each special equipment item is 99,999,999$. You can support data for 10 buildings, 100 components each; ESTS1071/01: There are 65 components for 28 systems available to specify the contaminated systems costs (BWR). ESTS1071/02: There are 75 components for 25 systems available to specify the contaminated systems costs (PWR)

  16. Dynamic analysis of BWR scram reactivity characteristics

    International Nuclear Information System (INIS)

    An extensive study of BWR scram reactivity behavior is presented. It is based on a space-time analysis of a BWR/4 code using the two-dimensional (R, Z) dynamics code BNL-TWIGL which includes a two-phase thermal-hydraulic model. Calculations were made of the sensitivity of scram to physical quantities such as initial control rod position and power distribution, scram speed, system pressure and varying inlet flow rate and temperature. The end-of-cycle Haling operating condition was found to give rise to the limiting scram reactivity function. Even with scram a power surge was found to be possible with severely decreasing inlet temperature. Calculations were also made to find the effect on scram of commonly used modeling approximations. These included the effect of neglecting delayed neutrons (conservative), using a time invariant void distribution (non-conservative) and defining point kinetics parameters such as reactivity, amplitude function and generation time in terms of different weighting functions. The importance of defining point kinetics parameters consistent with their use in plant transient analyses was demonstrated with particular emphasis on the role of ''residual reactivity''

  17. Vibration characteristics of BWR primary containment vessel

    International Nuclear Information System (INIS)

    This paper reports on a part of a seismic proving test of a BWR Primary Containment Vessel (PCV), a vibration analysis that is carried out considering the effect of reinforcement around attached masses such as equipment hatches and a personnel airlock. A deflected shape with the local attached masses and reinforcement is expanded using free vibration modes of an axisymmetric PCV without attached masses and reinforcement. The free vibration modes are calculated considering a coupling effect between the PCV shell and suppression pool water using conical shell finite elements and ring fluid elements. The same expanding functions and free vibration modes are used to express strain energy, kinetic energy and work done by inertia forces due to input acceleration considering locally varying thickness of the PCV and attached masses. The equations of motion are obtained by substituting the strain energy, kinetic energy, and work done by inertia forces into the Lagrange's equation. Calculated natural frequencies, free vibration modes and frequency response functions are compared with numerical results obtained by a general shell finite element analysis and with test results using the large-scale high-performance vibration table of Tadotsu Engineering Laboratory, Nuclear Power Engineering Test Center (NUPEC). The comparison shows a reasonable agreement, leading to better understanding of the dynamic characteristics of the BWR PCV

  18. BWR normal water chemistry guidelines: 1986 revision

    International Nuclear Information System (INIS)

    Boiling water reactors (BWRs) have experienced stress corrosion cracking in the reactor cooling system piping resulting in adverse impacts on plant availability and personnel radiation exposure. The BWR Owners Group and EPRI have sponsored a major research and development program to provide remedies for this stress corrosion cracking problem. This work shows that the likelihood of cracking depends on the plant's water chemistry performance (environment) as well as on material condition and stress level. Plant experience and other research demonstrate that water quality also affects fuel performance and radiation field buildup in BWRs. This report,''BWR Normal Water Chemistry Guidelines: 1986 Revision,'' presents suggested generic water chemistry specifications, justifies the proposed water chemistry limits, suggests responses to out-of-specification water chemistry, discusses available chemical analysis methods as well as data management and surveillance schemes, and details the management philosophy required to successfully implement a water chemistry control program. An appendix contains recommendations for water quality of auxiliary systems. 73 refs., 20 figs., 9 tabs

  19. 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.

  20. 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)

  1. BWR Assembly Optimization for Minor Actinide Recycling

    International Nuclear Information System (INIS)

    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). A top-level objective of the Advanced Fuel Cycle Systems Analysis program element of the DOE NERI program is to investigate spent fuel treatment and recycling options for current light water reactors (LWRs). Accordingly, this project targets to expand the traditional scope of nuclear fuel management optimization into the following two complementary specific objectives: (1) To develop a direct coupling between the pin-by-pin within-bundle loading control variables and core-wide (bundle-by-bundle) optimization objectives, (2) to extend the methodology developed to explicitly encompass control variables, objectives, and constraints designed to maximize minor actinide incineration in BWR bundles and cycles. The first specific objective is projected to 'uncover' dormant thermal margin made available by employing additional degrees of freedom within the optimization process, while the addition of minor actinides is expected to 'consume' some of the uncovered thermal margin. Therefore, a key underlying goal of this project is to effectively invest some of the uncovered thermal margin into achieving the primary objective.

  2. Thermal aging evaluation of casting stainless steel under BWR environment

    International Nuclear Information System (INIS)

    Effect of thermal aging under BWR condition on material properties of casting stainless steel were evaluated by such as Charpy impact test, using replaced BWR component material. Solution heat treatment was performed to the same material and the material properties were obtained. Comparing each material test results, impact value of thermal aging material was lower than solution heat treatment material. By the results, thermal aging effect on material properties under BWR condition was confirmed. The material properties were compared with model equation using PLM evaluation and conservativeness of model equation was confirmed. (author)

  3. Development of new irradiation facility for BWR safety research

    International Nuclear Information System (INIS)

    In JAEA (Japan Atomic Energy Agency), about the irradiation embrittlement of the reactor pressure vessel and the stress corrosion cracking of reactor core composition apparatus concerning the long-term use of the light water reactor (BWR), in order to check the influence of the temperature, pressure, and water quality, etc on BWR condition. The water environmental control facility which performs irradiation assisted stress corrosion-cracking (IASCC) evaluation under BWR irradiation environment was fabricated in JMTR (Japan Materials Testing Reactor). This report is described the outline of manufacture of the water environmental control facility for doing an irradiation test using the saturation temperature capsule after JMTR re-operation. (author)

  4. Up-regulation of Hsp27 by ERα/Sp1 facilitates proliferation and confers resistance to apoptosis in human papillary thyroid cancer cells.

    Science.gov (United States)

    Mo, Xiao-Mei; Li, Li; Zhu, Ping; Dai, Yu-Jie; Zhao, Ting-Ting; Liao, Ling-Yao; Chen, George G; Liu, Zhi-Min

    2016-08-15

    17β-estradiol (E2) has been suggested to play a role in the development and progression of papillary thyroid cancer. Heat shock protein 27 (Hsp27) is a member of the Hsp family that is responsible for cell survival under stressful conditions. Previous studies have shown that the 5'-promoter region of Hsp27 gene contains a specificity protein-1 (Spl) and estrogen response element half-site (ERE-half), which contributes to Hsp27 induction by E2 in breast cancer cells. However, it is unclear whether Hsp27 can be up-regulated by E2 and which estrogen receptor (ER) isoform and tethered transcription factor are involved in this regulation in papillary thyroid cancer cells. In the present study, we demonstrated that Hsp27 can be effectively up-regulated by E2 at mRNA and protein levels in human K1 and BCPAP papillary thyroid cancer cells which have more than two times higher level of ERα than that of ERβ. The up-regulation of Hsp27 by E2 is mediated by ERα/Sp1 and ERβ has repressive effect on this ERα/Sp1-mediated up-regulation of Hsp27. Moreover, we showed that the up-regulation of Hsp27 by ERα/Sp1 facilitates proliferation and confers resistance to apoptosis through interaction with procaspase-3. Targeting this pathway may be a potential strategy for therapy of papillary thyroid cancer. PMID:27179757

  5. YME1L controls the accumulation of respiratory chain subunits and is required for apoptotic resistance, cristae morphogenesis, and cell proliferation.

    Science.gov (United States)

    Stiburek, Lukas; Cesnekova, Jana; Kostkova, Olga; Fornuskova, Daniela; Vinsova, Kamila; Wenchich, Laszlo; Houstek, Josef; Zeman, Jiri

    2012-03-01

    Mitochondrial ATPases associated with diverse cellular activities (AAA) proteases are involved in the quality control and processing of inner-membrane proteins. Here we investigate the cellular activities of YME1L, the human orthologue of the Yme1 subunit of the yeast i-AAA complex, using stable short hairpin RNA knockdown and expression experiments. Human YME1L is shown to be an integral membrane protein that exposes its carboxy-terminus to the intermembrane space and exists in several complexes of 600-1100 kDa. The stable knockdown of YME1L in human embryonic kidney 293 cells led to impaired cell proliferation and apoptotic resistance, altered cristae morphology, diminished rotenone-sensitive respiration, and increased susceptibility to mitochondrial membrane protein carbonylation. Depletion of YME1L led to excessive accumulation of nonassembled respiratory chain subunits (Ndufb6, ND1, and Cox4) in the inner membrane. This was due to a lack of YME1L proteolytic activity, since the excessive accumulation of subunits was reversed by overexpression of wild-type YME1L but not a proteolytically inactive YME1L variant. Similarly, the expression of wild-type YME1L restored the lamellar cristae morphology of YME1L-deficient mitochondria. Our results demonstrate the importance of mitochondrial inner-membrane proteostasis to both mitochondrial and cellular function and integrity and reveal a novel role for YME1L in the proteolytic regulation of respiratory chain biogenesis. PMID:22262461

  6. Recent experience and development of BWR fuel at NFI

    International Nuclear Information System (INIS)

    This paper describes the results of recent investigations by Nuclear Fuel Industries, Ltd. (NFI) conducted in cooperation with BWR electric power companies in Japan regarding high burnup fuel behavior, i.e. fuel cladding corrosion and hydrogen pickup, degradation of pellet thermal conductivity with burnup, and fission gas release. The authors confirmed by pool inspection that 9x9 assemblies irradiated up to 53 GWd/t, which is the maximum burnup in our experience, showed good performance without any harmful phenomena. With respect to the advanced Zr alloy HiFi, it was confirmed that HiFi retained high corrosion resistance and showed low hydrogen pick up and good mechanical properties after six cycles of irradiation. Regarding the high burnup fuel behavior, it was confirmed that the thermal behavior of the fuel, such as pellet thermal conductivity degradation and fission gas release behavior beyond 80 GWd/t, was stable in the extrapolation range of the burnup fuel behavior between about 60-70 GWd/t. In addition, a fuel performance analysis code developed by NFI was verified to predict the data measured beyond 80 GWd/t well. (author)

  7. Reduced moderation BWR with advanced recycle system (BARS)

    International Nuclear Information System (INIS)

    lattice. Glass rods coated with transparent electric resistant heater material are adopted in this experiment. The two-phase (steam and water) flow behavior in the test channel has been recorded by video camera and analyzed. These experiments have been done under the atmospheric pressure in order to make it easy to observe. The second one is establishing of the BT correlation equation on tight lattice bundle. Mini tight lattice bundle experiments were planned to study the correlation. These experiment include 7-rods hexagonal bundle experiment and 14-rods square bundle experiment. Various rod configurations on tight lattice have been examined in 7-rods bundle experiments for the parametric study on tight lattice. The 14-rods bundle experiment aims the measurement of square channel effect. One of the important purposes of these experiments is to verify the existing the BT correlation equation such as Arai's equation on tight lattice bundle. These experiments have been done in Toshiba's high temperature and high-pressure BWR experimental facility for stability and transient test (BEST). The third one is cooling property for tight lattice in hypothetical accident evaluation. Counter-current-flow-limitation (CCFL) experiment was planned for this purpose. (author)

  8. ATRIUMTM Fuel - Continuous Upgrading for High Duty BWR Plants

    International Nuclear Information System (INIS)

    of the U.S. BWR plants do or will operate on 120 % of the initial rated power and the same trend with up-rates between 10-35 % is valid for most European BWR plants. For the individual fuel bundle, the average power is hence increased proportionally which reduces the core loading flexibility, the margins to PCI and the critical bundle dryout power. AREVA NP has developed a BWR fuel assembly to meet the enhanced reliability and operating demands of the next decade of all BWR plants. In this challenge, the decision to move to an 11x11 fuel rod array was a clear consequence driven by our customers' expectation of a high performing fuel at a maximum of reliable, flexible and problem-free operation. A major design optimization goal was to increase the heated length in the fuel assembly to fully or partly compensate the reactor power increase. Less power load on a fuel rod enables high burnup with less operating restrictions and much lower risk of damage. Today, the dryout performance of our available ATRIUMTM 10XM product is excellent and an ambitious benchmark for any new fuel design. More heated length will generally raise the critical power. The design goal was set to achieve the fuel utilization improvement goal while not degrading dryout performance compared to this existing fuel product. Besides the lattice design itself, the spacer grid design plays the key role for achieving high critical power and, at the same time, is strongly impacting the total fuel assembly pressure drop. Since 2005, various spacer grid design concepts were investigated and tested. Finally, an evolution of the egg-crate ULTRAFLOWTM spacer design featuring integrated springs and improved vane geometry will provide an optimum of performance, resistance against debris capture and robustness. The ATRIUMTM 11 will also feature a fully symmetric rod array with a centrally positioned square internal water channel. It provides the load-carrying structure as in our current ATRIUMTM fuel family as well

  9. Sunitinib significantly suppresses the proliferation, migration, apoptosis resistance, tumor angiogenesis and growth of triple-negative breast cancers but increases breast cancer stem cells.

    Science.gov (United States)

    Chinchar, Edmund; Makey, Kristina L; Gibson, John; Chen, Fang; Cole, Shelby A; Megason, Gail C; Vijayakumar, Srinivassan; Miele, Lucio; Gu, Jian-Wei

    2014-01-01

    The majority of triple-negative breast cancers (TNBCs) are basal-like breast cancers. However there is no reported study on anti-tumor effects of sunitinib in xenografts of basal-like TNBC (MDA-MB-468) cells. In the present study, MDA-MB-231, MDA-MB-468, MCF-7 cells were cultured using RPMI 1640 media with 10% FBS. Vascular endothelia growth factor (VEGF) protein levels were detected using ELISA (R & D Systams). MDA-MB-468 cells were exposed to sunitinib for 18 hours for measuring proliferation (3H-thymidine incorporation), migration (BD Invasion Chamber), and apoptosis (ApopTag and ApoScreen Anuexin V Kit). The effect of sunitinib on Notch-1 expression was determined by Western blot in cultured MDA-MB-468 cells. 10(6) MDA-MB-468 cells were inoculated into the left fourth mammary gland fat pad in athymic nude-foxn1 mice. When the tumor volume reached 100 mm(3), sunitinib was given by gavage at 80 mg/kg/2 days for 4 weeks. Tumor angiogenesis was determined by CD31 immunohistochemistry. Breast cancer stem cells (CSCs) isolated from the tumors were determined by flow cytometry analysis using CD44(+)/CD24(-) or low. ELISA indicated that VEGF was much more highly expressed in MDA-MB-468 cells than MDA-MB-231 and MCF-7 cells. Sunitinib significantly inhibited the proliferation, invasion, and apoptosis resistance in cultured basal like breast cancer cells. Sunitinib significantly increased the expression of Notch-1 protein in cultured MDA-MB-468 or MDA-MB-231 cells. The xenograft models showed that oral sunitinib significantly reduced the tumor volume of TNBCs in association with the inhibition of tumor angiogeneisis, but increased breast CSCs. These findings support the hypothesis that the possibility should be considered of sunitinib increasing breast CSCs though it inhibits TNBC tumor angiogenesis and growth/progression, and that effects of sunitinib on Notch expression and hypoxia may increase breast cancer stem cells. This work provides the groundwork for an

  10. Small scale BWR core debris eutectics formation and melting experiment

    International Nuclear Information System (INIS)

    A small scale experiment has recently been performed at Oak Ridge under the auspices of the BWR Severe Accident Technology (BWRSAT) program to provide information concerning the formation of mixtures during heatup of representative BWR reactor vessel bottom head debris and to determine the composition and melting temperatures of these mixtures. The initial structure of the bottom head debris layers modeled in the experiment was taken from the results of recent BWR Accident Response (BWRSAR) code predictions for the short-term station blackout accident sequence. The experimental results provide useful information concerning the mixtures formed and their proportions and properties. The observed run-off of a stainless steel-zirconium eutectic alloy supports the contention that the initial pour from a BWR reactor vessel would consists of molten metals at relatively low temperatures. (orig.)

  11. BWR stability: history and state-of-the-art

    International Nuclear Information System (INIS)

    The paper briefly recalls the historical developments, reviews the important phenomena, the analytical and simulation tools that are used for the analysis of BWR stability focussing on the linear, frequency domain methods

  12. Swedes repair BWR thermal fatigue cracks

    International Nuclear Information System (INIS)

    In connection with an accident at the Barsebaeck-2 NPP the causes of cracking in steel pipe fittings of the BWR type reactor cooling system are investigated. In the course of testing carried out by the methods of gamma radiography and liquid penetrant inspection the cracks 10-100 mm long and with depth up to 10 mm are found. The most of the cracks is concentrated in regions near pipe fittings in the direction of water stream flow. The cause of crack formation is the thermal stress arising during mixing the water with different temperatures in particular, the feedwater having at normal operational conditions temperature of 180 deg C and the emergency cooling system water with the temperature of 270 deg C. The conclusion is drawn on the necessity of designing the new configurations of joints which are able to withstand the temperature gradients

  13. Transportation activities for BWR fuels at NFI

    International Nuclear Information System (INIS)

    Nuclear Fuel Industries, LTD. (NFI) supplies fuel assemblies for both PWR and BWR in Japan. We can also manage transportation of the fuel assemblies from our fabrication facilities to the nuclear power plants of Japanese utilities. For the transportation of fuel assembly, we designed and fabricated the transportation containers to meet the requirements of the IAEA regulations, and licensed in Japan. This paper introduced the recent activity and R and D of NFI concerning transportation of BWR fuel assembly. NT-XII transportation container was developed for fresh BWR fuel assemblies. NT-XII container consists of inner container and outer container. Two BWR fuel assemblies with up to 5 wt.-% 235-U enriched are enclosed in an inner container. In the concept of NT-XII container design, we made the best priority to transportation efficiency, as well as ensuring fuel integrity during transportation. NT-XII has been used since 2002 in Japan. Thanks to the lightening weight of containers, the number of containers to be loaded to one transportation truck was increased up to 9 containers (equivalent to 18 fuel assemblies) compared with former type container (NT-IV transportation container) which can be loaded up to 6 containers (equivalent to 12 fuel assemblies). In addition to the design of brand-new container, we promote the improvement of the packaging methods. In Japan, in order to reduce the damage to the fuel rod and fuel spacer while transporting, polyethylene sleeves which are called 'packing separators' are inserted in the rod-to-rod gap of fuel assembly. However, packing separators requires time and cost for the installation at fuel fabrication facility. In the same way, huge time and cost are needed for removal of packing separators at nuclear power plant. For the improvement of preparation efficiency before and after transportation, we investigated the influence of vibration to fuel integrity in case of transportation without packing separators. Based on the above

  14. Radiation source term reduction in BWR plants

    International Nuclear Information System (INIS)

    This series of slides presents: the collective radiation exposures at US and European BWRs; the European experience with source term reduction measures (normal water chemistry - NWC): zinc addition, stellite replacement, full system decontamination; the effects of evolving water chemistries/US experience. The conclusions are summarized as follows: worldwide reduction of collective radiation exposures at BWRs by following the ALARA principle; zinc addition proven option for source term reduction for NWC and hydrogen water chemistry (HWC) plants; reducing feedwater iron has been proven to reduce dose rates - as operational observations in the US indicate; optimized feedwater iron is very important for fuel performance under all modes of water chemistry (HWC, Zn, and noble metal chemical addition (NMCA)); minimize 59Co sources/stellite, follow the ALARA principle; full system decontamination (FSD) plus zinc injection is an attractive option for reducing reactor coolant system (RCS) dose rates of mature BWR plants

  15. Improvement of BWR radioactive waste disposal system

    International Nuclear Information System (INIS)

    Description is made here about the improvement of the disposal systems for liquid and solid radioactive waste in recent BWR plants. Regarding the improvement of liquid waste disposal systems in Toshiba, emphasis was laid on crud removal and laundry drain treatment; as a result, a crud removal system utilizing a centrifugal separation mechanism and an evaporative enriching system using anti-foaming agent have been practicalized. An important problem left for solution is the method of volume reduction of solid waste. A method attracting attention of late is the plastic solidification which is far superior to the conventional cement and bitumen solidification. In Toshiba, this method is promising to be practicalized, in which the waste is dried and then solidified with thermosetting resin. (author)

  16. Reactor fueling of BWR type reactors

    International Nuclear Information System (INIS)

    Purpose: To enable the pattern exchange for control rods during burning in Control Cell Core type BWR reactors. Constitution: A plurality of control cells are divided into a plurality of groups such that the control cells is aparted from each other by way of at least two fuel assemblies other than the control cells with respect to the vertical and lateral directions of the reactor core cross section, as well as they are in adjacent with control cells of other groups with respect to the orthogonal direction. This enables to perform the pattern exchange for the control rods during burning in the control cell core with ease, and the control blade and the story effect harmful to the mechanical soundness of fuels can thus be suppressed. (Moriyama, K.)

  17. BWR stability using a reducing dynamical model

    International Nuclear Information System (INIS)

    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)

  18. Manufacturing technology and process for BWR fuel

    International Nuclear Information System (INIS)

    Following recent advanced technologies, processes and requests of the design changes of BWR fuel, Nuclear Fuel Industries, Ltd. (NFI) has upgraded the manufacturing technology and honed its own skills to complete its brand-new automated facility in Tokai in the latter half of 1980's. The plant uses various forms of automation throughout the manufacturing process: the acceptance of uranium dioxide powder, pelletizing, fuel rod assembling, fuel bundle assembling and shipment. All processes are well computerized and linked together to establish the integrated control system with three levels of Production and Quality Control, Process Control and Process Automation. This multi-level system plays an important role in the quality assurance system which generates the highest quality of fuels and other benefits. (author)

  19. TRU transmutation type BWR fuel assembly

    International Nuclear Information System (INIS)

    The BWR fuel assembly is formed by bundling a plurality of fuel rods and a water channel disposed at the center of the assembly by a plurality of spacers. An upper tie plate and a lower tie plate are disposed to upper and lower portions of the fuel rods and the water channel respectively. An upper end plug of the water channel is attached detachably to a cylindrical main body of the water channel. A zircaloy tube incorporating TRU nuclides is contained and secured in the water channel. The zircaloy tube has such a structure as capable of incorporating and sealing oxides or metal materials containing TRU nuclides. Since the zircaloy tube containing TRU nuclides is contained not in fuel region but in the water rod, the loaded uranium amount of fuels is not reduced but the reactivity can be ensured. (I.N.)

  20. Contingency programs for BWR pipe cracking

    International Nuclear Information System (INIS)

    General Electric (GE) has aggressively addressed the problem of Intergranular Stress Corrosion Cracking (IGSCC). Intergranular Stress Corrosion Cracking has occurred in boiling water reactors in less than 1% of the Type-304 stainless steel welds in operating plants. However a comprehensive program by GE, with EPRI support in many cases, has provided technical solutions to limit IGSCC. As part of this program the Nuclear Services Department (NSD) has taken new technology and applied it to programs for the prevention of IGSCC and for repair of pipe cracks should they occur. The purpose of this paper is to provide a description and the current status of the major ongoing service programs for dealing with the problems related to IGSCC in BWR operating plants

  1. BWR stability using a reduced dynamical model

    International Nuclear Information System (INIS)

    BWR stability can be treated with reduced order 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 struct-ure 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 in non-linear. Simple parametric calculat-ion of decay ratio are performed, showing why BWRs become unstable when they are operated at low flow and high power. (Author). 7 refs

  2. BWR startup and shutdown activity transport control

    International Nuclear Information System (INIS)

    This paper summarizes BWR industry experience on good practices for controlling the transport of corrosion product activity during shutdowns, particularly refueling outages, and for startup chemistry control to minimize IGSCC (intergranular stress corrosion cracking). For shutdown, overall goals are to minimize adverse impacts of crud bursts and the time required to remove activated corrosion products from the reactor coolant during the shutdown process prior to refueling, and to assist plants in predicting and controlling radiation exposure during outages. For startup, the overall goals are to highlight conditions during early heatup and startup when sources of reactor coolant oxidants are high, when there is a greater likelihood for chemical excursions associated with refueling outage work activities, and when hydrogen injection is not available to mitigate IGSCC due to system design limitations. BWR water chemistry has changed significantly in recent years with the adoption of hydrogen water chemistry, zinc addition and noble metal chemical applications. These processes have, in some instances, resulted in significant activity increases during shutdown evolutions, which together with reduced time for cleanup because of shorter outages, has consequently increased outage radiation exposure. A review several recent outages shows that adverse effects from these conditions can be minimized, leading to the set of good practice recommendations for shutdown chemistry control. Most plants lose the majority of their hydrogen availability hours during early startup because feedwater hydrogen injection systems were not originally designed to inject hydrogen below 20% power. Hydrogen availability has improved through modifications to inject hydrogen at lower power levels, some near 5%. However, data indicate that IGSCC is accelerated during early startup, when dissolved oxygen and hydrogen peroxide levels are high and reactor coolant temperatures are in the 300 to 400 oF (

  3. In-reactor ECP measurements in BWR plants

    International Nuclear Information System (INIS)

    It has previously been confirmed that the initiation and propagation of Stress Corrosion Cracking (SCC) of stainless steel and Ni-based alloy exposed in the primary water of Boiling Water Reactor (BWR) depend on water chemistry. It is known that the corrosion environment is evaluated with Electrochemical Corrosion Potential (ECP) and SCC susceptibility is high when the ECP is high. Then it also has been confirmed that hydrogen injection in feedwater and Noble Metal Chemical Addition (NMCA) are effective to reduce ECP of reactor components. ECP measurements in BWR plants, which are BWR-3 and BWR-4, were performed in Normal Water Chemistry (NWC), Hydrogen Water Chemistry (HWC) and post-NMCA environments to evaluate the mitigation effect of SCC by HWC and NMCA. ECP measurements were conducted in the lower plenum region, bottom head region, below the core plate and near the bottom of the active fuel region, by installing modified LPRM with ECP electrodes, and at bottom head drain line flange location (BHDL). It is confirmed that the ECP are reduced to less than -200mV(SHE) by 0.9 ppm H2 concentration at feedwater in BWR-3 and 1.1 ppm H2 concentration at feedwater in BWR-4, and that the ECP can be reduced to less than -200mV(SHE) by 0.3 ppm or less H2 concentration after NMCA is applied. (author)

  4. 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.

  5. Proliferation Resistant Nuclear Reactor Fuel

    International Nuclear Information System (INIS)

    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.

  6. Involvement of peroxisome proliferator-activated receptors in cardiac and vascular remodeling in a novel minipig model of insulin resistance and atherosclerosis induced by consumption of a high-fat/cholesterol diet

    OpenAIRE

    Yongming, Pan; Zhaowei, Cai; Yichao, Ma; Keyan, Zhu; Liang, Chen; Fangming, Chen; Xiaoping, Xu; Quanxin, Ma; Minli, Chen

    2015-01-01

    Background A long-term high-fat/cholesterol (HFC) diet leads to insulin resistance (IR), which is associated with inflammation, atherosclerosis (AS), cardiac sympathovagal imbalance, and cardiac dysfunction. Peroxisome proliferator-activated receptors (PPARs) and nuclear factor ĸB (NF-κB) are involved in the development of IR-AS. Thus, we elucidated the pathological molecular mechanism of IR-AS by feeding an HFC diet to Tibetan minipigs to induce IR and AS. Methods Male Tibetan minipigs were ...

  7. Thermal-hydraulic stability tests for newly designed BWR rod bundle (step-III fuel type A)

    International Nuclear Information System (INIS)

    Thermal-hydraulic stability tests have been performed on electrically heated bundles to simulate the newly designed Boiling Water Reactor (BWR) fuels in a parallel channel test loop. The objective of the current experimental program is to investigate how the newly designed bundle could improve the thermal-hydraulic stability. Measurements of the thermal-hydraulic instability thresholds in two vertical rod bundles have been conducted in steam-water two-phase flow conditions at the TOSHIBA test loop. Fluid conditions were BWR operating conditions of 7 MPa system pressure, 1.0-2.0x106 kg/m2/h inlet mass flux and 28-108 kJ/kg inlet subcooling. The parallel channel test loop consists of a main bundle of 3x3 indirectly heated rods of 1/9 symmetry of 9x9 full lattice and a bypass bundle of 8x8. These are both simulated BWR rod bundles in respect of rod diameter, heated length, rod configuration, fuel rod spacer, core inlet hydraulic resistance and upper tie plate. There are three types of the 3x3 test bundles with different configurations of a part length rod of two-thirds the length of the other rods and an axial power shape. The design innovation of the part length rod for a 9x9 lattice development, though addition of more fuel rods increases bundle pressure drop, reduces pressure drop in the two-phase portion of the bundle, and enhances the thermal hydraulic stability. Through the experiments, the parameter dependency on the channel stability threshold is obtained for inlet subcooling, inlet mass flux, inlet flow resistance, axial power shape and part length rod. The main conclusion is that the stability threshold is about 10% greater with the part length rod than without the part length rod. The new BWR bundle consisting of the part length rod has been verified in respect of thermal hydraulic stability performance. (author)

  8. Proliferation risks

    International Nuclear Information System (INIS)

    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

  9. BWR Source Term Generation and Evaluation

    International Nuclear Information System (INIS)

    This calculation is a revision of a previous calculation (Ref. 7.5) that bears the same title and has the document identifier BBAC00000-01717-0210-000061. 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 Quality

  10. 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

  11. Handling Proliferation

    Directory of Open Access Journals (Sweden)

    Pierre Laszlo

    2001-10-01

    Full Text Available The ethics of the chemist identify with those of the citizen, in principle. The observed perversions, such as proliferation of chemicals, stem from the values of a chemical community closed upon itself, and from the attendant identification of a mere know-how with a science. The epistemic degradation produces moral indifference.

  12. BWR operator training for emergency conditions using simulator in Japan

    International Nuclear Information System (INIS)

    BWR Operator Training Center Corporation (BTC) is the only training organization for all Japanese BWR utilities. BWR operator training program is constructed to maintain and upgrade the ability of the individual operator and the operation crews. The training course for the individual operator is focused on the operator's positions, from that of the main operator to the licensed shift supervisor. The target of the training for operation crews, which is called the family training, is to upgrade the team performance including shift supervisor. The training programs are combined to improve and upgrade the operator performance using simulator. BWR training simulator has been designed to demonstrate many kinds of malfunctions, modeling a single failure and multiple failure, including loss of coolant accident. It has also the alteration and scheduling function to simulate the abnormal and accident situation. These functions are utilized to easily simulate the accident scenario and enlarge the operator's knowledge in case of the abnormal condition. This paper describes the BWR operator training for emergency conditions using simulator and the advanced control board simulator performance including the abnormal situation functions. (author). 5 refs, 2 tabs

  13. BWR blowdown/emergency core cooling integral program

    International Nuclear Information System (INIS)

    The Program plan identifying the phased approach to testing has been completed and the first test phase is well underway. Test results to date show the expected lower peak cladding temperatures and slower system blowdown response for a BWR/6 LOCA simulation compared to the BWR/4. Counter-current flow limiting (CCFL) tests show that exact geometric replication was not necessary to simulate CCFL characteristics for a BWR prototype fuel bundle upper tie plate. Separate effects blowdown tests demonstrate the importance of flow length scaling in simulating break geometry for limiting the critical or blowdown flow rate. A new bundle thermal hydraulic method, MAYU04, has been completed. This method is shown to provide a substantial improvement in the prediction of bundle temperatures. The BD/ECC Program represents an important contribution to BWR safety research. Early results from the program have already provided a better understanding of the governing phenomena during hypothetical LOCA simulation tests. Future tests are expected to provide a basis for further improvements in BWR LOCA phenomena modeling

  14. Reliability of BWR high pressure core cooling

    International Nuclear Information System (INIS)

    The high pressure coolant injection system (HPCI), and the reactor core isolation cooling system (RCIC) are steam turbine driven systems that can inject water into a boiling water reactor at full operating pressure. Their purpose is to supply water during any failure that allows water to be lost while the reactor is at pressure and temperature. A large number of BWR plants are not meeting HPCI and RCIC performance goals for core cooling. NSAC considers concurrent failure of NPCI and RCIC to be the most probable potential cause of low reactor water level and possibly fuel damage in a boiling water reactor. Between January 1978 and May 1981, 169 licensee event reports were filed where HPCI or RCIC was inoperable or was declared inoperable. The present effort has shown that at least 40% of NPCI and RCIC problems might be averted by a high quality preventive maintenance program. About half of the plants do not perform cold quick-start surveillance testing of HPCI and RCIC. They do perform routine startup tests, but the equipment is first preheated and the startup is relatively gentle. However, emergency start-ups are abrupt and from the cold condition. Therefore, cold quick-start testing is the only way to assure that all components, control systems, and instruments are functioning correctly for automatic safety initiation. (author)

  15. BWR plant analyzer development at BNL

    International Nuclear Information System (INIS)

    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 speed 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. The accomplishment of detailed and accurate simulations in complex power plants at high speed and low cost are due chiefly to two reasons. The first reason is the application of five distinct modeling principles [2] which are not employed in any other simulation code. The second, and even more important reason is the utilization of a special-purpose peripheral computer with its 13 task-specific parallel processors

  16. Control method for BWR type power plant

    International Nuclear Information System (INIS)

    The present invention provides a method of controlling a BWR type plant having internal pumps capable of sufficiently utilizing the performance of a whole volume turbine bypass plant to enable stable supply of electric power upon load interruption of power generator thereof. Namely, upon occurrence of load interruption of a power generator or turbine trip, a plurality of internal pumps are tripped simultaneously to abruptly reduce a reactor core flow rate by a predetermined value or more. In this case, a reactor core flow rate abruptly reduction scram signal is prevented. Alternatively, a plurality of internal pumps are tripped simultaneously to abruptly reduce the reactor core flow rate. In this case, a reactor core flow rate abrupt reduction scram set value is changed in order to inhibit the reactor core flow rate abrupt reduction scram signal. With such procedures, upon load interruption of power generator or upon trip of turbine, reactor core flow rate is abruptly reduced by trip of internal pumps for avoiding increase of neutron fluxes due to reactor pressure change. However, since reactor scram is avoided, the operation can be continued upon load interruption of power generator. As a result, performance of whole volume turbine bypass plant can be utilized sufficiently even upon occurrence of load interruption of power generator. (I.S.)

  17. Analysis for steam separators in BWR

    International Nuclear Information System (INIS)

    Steam seperators in BWR are thin walled pipes with special cyclons for seperating steam and water arranged in a hexagonal scheme and welded to the core cover head. The pipes are connected to each other in two planes by a gridwork of small upright sheets also arranged in a hexagonal scheme. At the level of these gridworks there are also shroud rings for holding long screws. The sheets of the gridwork run concentric to the tubes and are interrupted by the tubes. The interiour forces of the gridwork-members are transferred through the tubewalls by circumferential bending which causes extreme elastic ovalizing of the pipe corss-section near to the gridwork connection. The height of these members may be constant or broader at the pipe connection (fig. 1d) for reducing the bending stresses in the pipe walls. If this ovalizing is involved in the elastic behaviour of the gridwork it works like a plate with a Poisson ratio ν approx. equal to 1. Such structures are extremely stiff against equal stresses in perpendicular directions and extremely weak for opposite stresses in both directions. Instead of this interrupted grid work the upper one may be a continues one, arranged tangentially to the tupes. The elastic behaviour of this gridwork is quite different to that of the interrupted one. (orig./GL)

  18. BWR control rod design using tabu search

    International Nuclear Information System (INIS)

    An optimization system to get control rod patterns (CRP) has been generated. This system is based on the tabu search technique (TS) and the control cell core heuristic rules. The system uses the 3-D simulator code CM-PRESTO and it has as objective function to get a specific axial power profile while satisfying the operational and safety thermal limits. The CRP design system is tested on a fixed fuel loading pattern (LP) to yield a feasible CRP that removes the thermal margin and satisfies the power constraints. Its performance in facilitating a power operation for two different axial power profiles is also demonstrated. Our CRP system is combined with a previous LP optimization system also based on the TS to solve the combined LP-CRP optimization problem. Effectiveness of the combined system is shown, by analyzing an actual BWR operating cycle. The results presented clearly indicate the successful implementation of the combined LP-CRP system and it demonstrates its optimization features

  19. Fuel assembly for BWR type reactor

    International Nuclear Information System (INIS)

    A fuel assembly of an BWR type reactor of present invention, in which a plurality of fuel rods are arranged in a regular square lattice like configuration include vibration-filled fuel rods in which granular nuclear fuel materials and granular non-nuclear fuel materials having a smaller neutron absorbing cross sectional area are mixed and filled. With such a constitution, the content of the mixed and filled non-nuclear fuel materials in the vibration filled fuel rods is at least 20% by a volume ratio in average in fuel assemblies. In addition, a burnable poison is optionally added and mixed to the granular mixture of the nuclear fuel material and the diluting granules. With such a constitution, the manufacturing cost can be reduced, and the combustion rate of the nuclear fission materials is increased to improve reactor core characteristics, thereby enabling to obtain sufficient Pu loading amount per assembly, and fuel assemblies excellent in flexibility in design and economic property can be obtained. (T.M.)

  20. Fuel assembly for BWR type reactor

    International Nuclear Information System (INIS)

    A fuel assembly has a 9 x 9 square lattice arrangement having a water channel which occupies an area of 3 x 3 lattice pattern corresponding to 9 fuel rods. Fuel pellets comprise those of not more 7 kinds which have fission products at enrichment degrees different by a spun of not less than 10%. Fuel rods comprise from 4 to 12 first type fuel rods and remaining second type fuel rods. The first type fuel rod is loaded with fuel pellets of fissionable products having an enrichment degree axially different at the upper and the lower portions. The second type fuel rod is loaded with fuel pellets of fissionable products having the same enrichment degree in the vertical direction. With such a constitution, the enrichment degree of fissionable products of fuel pellets in the fuel assembly for a BWR type reactor having different reactor constitution and operation conditions can be used in common. Accordingly, the degree of freedom for the design of the distribution of the enrichment degree is increased. (I.N.)

  1. Method of operating BWR type reactors

    International Nuclear Information System (INIS)

    Purpose: To enable reactor control depending on any demanded loads by performing control by the insertion of control rods in addition to the control by the regulation of the flow rate of the reactor core water at high power operation of a BWR type reactor. Method: The power is reduced at high power operation by decreasing the flow rate of reactor core water from the starting time for the power reduction and the flow rate is maintained after the time at which it reaches the minimum allowable flow rate. Then, the control rod is started to insert from the above time point to reduce the power to an aimed level. Thus, the insufficiency in the reactivity due to the increase in the xenon concentration can be compensated by the withdrawal of the control rods and the excess reactivity due to the decrease in the xenon concentration can be compensated by the insertion of the control rods, whereby the reactor power can be controlled depending on any demanded loads without deviating from the upper or lower limit for the flow rate of the reactor core water. (Moriyama, K.)

  2. Operation method for BWR type reactor

    International Nuclear Information System (INIS)

    In a BWR type reactor, the number of fuels at low enrichment, among initially loaded fuels, is increased greater than that of fuels to be exchanged, and the number of fuels at low enrichment remained in a reactor core after fuel exchange is decreased to smaller than that of entire control rods. Further, the fuels at low enrichment are disposed to the inner side except for the outermost circumference in the reactor core after fuel exchange. Since fuels of high reactivity are disposed at the outermost circumference in a second cycle, leakage of neutrons is increased and effective breeding factor is decreased. However, since the number of brought over fuels at low enrichment is decreased and the number of fuels at high enrichment is increased, effective average reactor core enrichment degree is increased, to compensate the lowering thereof due to the increase of neutron leakage. Since dispersion effect for the distribution of the enrichment degree can be utilized as much as possible by greatly reducing the number and the enrichment degree of fuels at low enrichment for initially loaded fuels, irrespective of the average enrichment degree and the fueling pattern in a first cycle, a burnup degree upon take-out of initially loaded fuels at ow enrichment degree can be increased to maximum. (N.H.)

  3. Sphere-pac versus pellet UO2 fuel in de Dodewaard BWR

    International Nuclear Information System (INIS)

    Comparative testing of UO2 sphere-pac and pellet fuel rods under LWR conditions has been jointly performed by the Netherlands Utilities Research Centre (KEMA) in Arnhem, the Netherlands Energy Research Foundation (ECN) at Petten and the Netherlands Joint Nuclear Power Utility (GKN) at Dodewaard. This final report summarizes the highlights of this 1968-1988 program with strong emphasis on the fuel rods irradiated in the Dodewaard BWR. The conclusion reached is that under normal LWR conditions sphere-pac UO2 in LWR fuel rods offers better resistance against stress corrosion cracking of the cladding, but that under fast, single step, power ramping conditions pellet UO2 in LWR fuel rods has a better resistance against hoop stress failure of the cladding. 128 figs., 36 refs., 19 tabs

  4. Measurement of pressure drops in prototypic BWR and PWR fuel assemblies in the laminar regime - Pressure drop measurement of laminar air flow in prototypic BWR and PWR fuel assemblies

    International Nuclear Information System (INIS)

    BWR tube bundle are partial length leaving significantly greater flow area in the top third of the bundle. With fewer grid spacers and expanded flow area in upper bundle, the BWR assembly exhibited less flow resistance at a given Reynolds number compared to the PWR assembly when located in a storage cell analogous to the BWR canister. This PWR storage cell was smaller than any used commercially in spent fuel pools or dry storage casks. When the PWR assembly was tested inside of storage cell sizes that spanned pool and cask cells available in industry, the flow resistance at a given Reynolds number was equivalent or less than that exhibited by the BWR assembly. These measurements should prove useful in independently validating CFD results or constructing numerically equivalent flow elements for use in fuel modeling efforts. (authors)

  5. 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.

  6. The BWR owners' group planning guide for life extension

    International Nuclear Information System (INIS)

    Extending the operating life of a commercial nuclear power plant has been shown to be economically beneficial to both the utility and the electric customer. As such, many utilities are planning and implementing plant life extension (PLEX) programs. A document has been developed which provides guidance to utilities in formulating a PLEX program plant for one or more boiling water reactor (BWR) plants. The guide has been developed by the BWR Owners' Group Plant Life Extension Committee. The principal bases for this guide were the BWR Pilot and Lead Plant Programs. These programs were used as models to develop the 'base plan' described in this guide. By formulating their program plant utilizing the base plan, utilities will be able to maximize the use of existing evaluations and results. The utility planner will build upon the base plan by adding any tasks or features that are unique to their programs. (author)

  7. The JAERI code system for evaluation of BWR ECCS performance

    International Nuclear Information System (INIS)

    Development of respective computer code system of BWR and PWR for evaluation of ECCS has been conducted since 1973 considering the differences of the reactor cooling system, core structure and ECCS. The first version of the BWR code system, of which developmental work started earlier than that of the PWR, has been completed. The BWR code system is designed to provide computational tools to analyze all phases of LOCAs and to evaluate the performance of the ECCS including an ''Evaluation Model (EM)'' feature in compliance with the requirements of the current Japanese Evaluation Guideline of ECCS. The BWR code system could be used for licensing purpose, i.e. for ECCS performance evaluation or audit calculations to cross-examine the methods and results of applicants or vendors. The BWR code system presented in this report comprises several computer codes, each of which analyzes a particular phase of a LOCA or a system blowdown depending on a range of LOCAs, i.e. large and small breaks in a variety of locations in the reactor system. The system includes ALARM-B1, HYDY-B1 and THYDE-B1 for analysis of the system blowdown for various break sizes, THYDE-B-REFLOOD for analysis of the reflood phase and SCORCH-B2 for the calculation of the fuel assembl hot plane temperature. When the multiple codes are used to analyze a broad range of LOCA as stated above, it is very important to evaluate the adequacy and consistency between the codes used to cover an entire break spectrum. The system consistency together with the system performance are discussed for a large commercial BWR. (author)

  8. Examination of dissimilar metal welds in BWR and PWR piping

    International Nuclear Information System (INIS)

    This paper addresses dissimilar metal weld examinations at PWRS. Surveys were conducted to document the dissimilar metal weld configurations at PWR plants and to update the information known about dissimilar metal weld configurations at BWR plants. The experiences which BWR utilities have had with dissimilar metal weld examinations are documented and include: correct identification of IGSCC, indications thought to be IGSCC but were actually fabrication flaws, and difficulties encountered with the examination of dissimilar metal welds after stress improvement. An experimental program was conducted which verified that the longitudinal wave procedures developed for BWRs are also applicable to PWR designs

  9. Evaluation of PWR and BWR pin cell benchmark results

    International Nuclear Information System (INIS)

    In order to carry out reliable reactor core calculations for a boiled water reactor (BWR) or a pressurized water reactor (PWR) first reactivity calculations have to be carried out for which several calculation programs are available. The purpose of the title project is to exchange experiences to improve the knowledge of this reactivity calculations. In a large number of institutes reactivity calculations of PWR and BWR pin cells were executed by means of available computer codes. Results are compared. It is concluded that the variations in the calculated results are problem dependent. Part of the results is satisfactory. However, further research is necessary

  10. Corrosion potential monitoring and its simulation in BWR conditions

    International Nuclear Information System (INIS)

    A simulation algorithm of corrosion potentials for BWR plant materials has been demonstrated. Cathodic and anodic electrochemical kinetic equations have been derived by analyzing kinetic models involving water radiolysis products of oxygen, hydrogen peroxide, and hydrogen. Corrosion rates of type 304 stainless steel with respect to corrosion potentials are formulated by numerical analysis as well. An electrochemical mixed potential theorem is applied to compute corrosion potentials. Flow rate effects of coolants on corrosion potentials of plant structural materials are expressed as a function of diffusion layer thickness. A fundamental technique and a theory to simulate corrosion potentials have been developed. Corrosion potentials in BWR conditions can be simulated by these results

  11. BWR plant dynamic analysis code BWRDYN user's manual

    International Nuclear Information System (INIS)

    Computer code BWRDYN has been developed for thermal-hydraulic analysis of a BWR plant. It can analyze the various types of transient caused by not only small but also large disturbances such as operating mode changes and/or system malfunctions. The verification of main analytical models of the BWRDYN code has been performed with measured data of actual BWR plant. Furthermore, the installation of BOP (Balance of Plant) model has made it possible to analyze the effect of BOP on reactor system. This report describes on analytical models and instructions for user of the BWRDYN code. (author)

  12. Impact of Plant Noise on BWR Stability Analyses

    International Nuclear Information System (INIS)

    A small amount of thermal-hydraulic noise is present in a boiling water reactor (BWR). The noise originates primarily from minor fluctuations in fluid flow and pressure distribution in the recirculation system of the BWR and manifests itself as a small fluctuation on the order of 1-2% for the average power range monitors (APRM) during normal operation. A larger noise level is observed for single-loop operation than for two-loop operation. This noise has an impact on the stability performance of the BWR. This is particularly the case when the noise contains a significant component at the resonant frequency for BWR instabilities, which is typically on the order of 0.5 Hz. For a pump trip event that can lead to instability, the noise will impact the growth rate of the reactor instability. The initial magnitude of the oscillations will be larger as the decay ratio increases above unity. For operation at low flow, such as for minimum pump speed or single loop operation where the decay ratio is larger, the impact of noise could lead to small oscillations at the resonant frequency for the APRM signals. The impact of noise on BWR instabilities is analyzed with the TRACG code. TRACG consists of a multi dimensional two-fluid thermal hydraulics model and the three-dimensional kinetics model consistent with the GE 3D core simulator, PANACEA. TRACG models the reactor primary system and has been extensively qualified against test data and BWR plant data. Thermal hydraulic instability test data, as well as data from BWR instability events and tests, have been used extensively in this qualification. This paper demonstrates the impact of noise on BWR stability response for events leading to instability, such as pump trip events, as well as operation at low core flow due to single loop operation. The impact is illustrated through sensitivity studies with the TRACG code and by comparison to plant data. The impact of reactor noise on the performance of the instability detection system

  13. BwrCrud: Development and validation of a new code for improved simulation of activity transport in BWR primary systems

    International Nuclear Information System (INIS)

    During the nineties, much reactor experience and significant new findings have been acquired. Thus, it became necessary to improve the previous code for computer model. Development of an improved code (named as BwrCrud) for modeling of activated corrosion products in BWR systems is progressing to model the influence of factors such as iron and zinc flow, HWC operation and fuel failures in a better way. The present version of the Code, BwrCrud 1.0 was tested in the Kashiwazaki Kariwa 5 (KK-5) plant and it was concluded that the cord can work satisfactorily and manage to model the KK-5 behavior appropriately in consideration of some uncertainties in input and verification data. Phase 2 of the project is to be completed in the beginning of 1999. (M.N.)

  14. 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)

  15. Non-proliferation considerations

    International Nuclear Information System (INIS)

    This paper reiterates the Indian viewpoint that consideration of ''proliferation resistance'' is outside the terms of reference of Working Group 4 as agreed at the Washington Conference. The discussions in WG4 should therefore cover only safeguards aspects. The paper goes on to critisize the various assessment factors introduced in INFCE/DEP./WG-4/104 and the various alternative technologies proposed. The Indian view is reinstated that if a country requires reprocessing based on its nuclear energy programmes and priorities, there should be no hindrance. International safeguards should be applied to all nuclear materials in all countries without discrimination or differentiation between civil and military programmes. The paper concludes that non-proliferation is essentially a political matter and has no technical solution

  16. Pool swell in a nuclear containment wetwell. [BWR

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez, T.

    1976-04-01

    A brief description is presented of scale model tests conducted to study LOCA induced wetwell pool swelling in the BWR Mk 1 containment pressure suppression system. The Mk 1 containment configuration is described together with the scale model design, the conduct of the tests, and the experimental results. (DG)

  17. 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)

  18. Transmutation of minor actinide using thorium fueled BWR core

    International Nuclear Information System (INIS)

    One of the methods to conduct transmutation of minor actinide is the use of BWR with thorium fuel. Thorium fuel has a specific behaviour of producing a little secondary minor actinides. Transmutation of minor actinide is done by loading it in the BWR with thorium fuel through two methods, namely close recycle and accumulation recycle. The calculation of minor actinide composition produced, weigh of minor actinide transmuted, and percentage of reminder transmutation was carried SRAC. The calculations were done to equivalent cell modeling from one fuel rod of BWR. The results show that minor actinide transmutation is more effective using thorium fuel than uranium fuel, through both close recycle and accumulation recycle. Minor actinide transmutation weight show that the same value for those recycle for 5th recycle. And most of all minor actinide produced from 5 unit BWR uranium fuel can transmuted in the 6th of close recycle. And, the minimal value of excess reactivity of the core is 12,15 % Δk/k, that is possible value for core operation

  19. High Pressure Boiling Water Reactor HP-BWR

    International Nuclear Information System (INIS)

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

  20. Experience and reliability of Framatome ANP's PWR and BWR fuel

    International Nuclear Information System (INIS)

    Based on three decades of fuel supply to 169 PWR and BWR plants on four continents, Framatome ANP has a very large database from operating experience feedback. The performance of Framatome PWR and BWR fuel is discussed for the period 1992-2001 with special emphasis on fuel failures, countermeasures and their effectiveness. While PWR fuel performance in most reactors has been good, the performance in some years did suffer from special circumstances that caused grid-to-rod fretting failures in few PWRs. After solving this problem, fuel of all types showed high reliability again. Especially the current PWR fuel products AFA 3G, HTP, Mark B and Mark BW showed a very good operating performance. Fuel reliability of Framatome ANP BWR fuel has been excellent over the last decade with average annual fuel rod failure rates under 1x10-5 since 1991. More than 40% of all BWR fuel failures in the 1992-2001 decade were caused by debris fretting. The debris problem has been remedied with the FUELGUARDTM lower tie plate, and by reactor operators' efforts to control the sources of debris. PCI, the main failure mechanism in former periods, affected only 10 rods. All of these rods had non-liner cladding. (author)

  1. Material operating behaviour of ABB BWR control rods

    International Nuclear Information System (INIS)

    The BWR control rods made by ABB use boron carbide (B4C and hafnium as absorber material within a cladding of stainless steel. The general behaviour under operation has proven to be very good. ABB and many of their control rod customers have performed extensive inspection programs of control rod behaviour. However, due to changes in the material properties under fast and thermal neutron irradiation defects may occur in the control rods at high neutron fluences. Examinations of irradiated control rod materials have been performed in hot cell laboratories. The examinations have revealed the defect mechanism Irradiation Assisted Stress Corrosion Cracking (IASCC) to appear in the stainless steel cladding. For IASCC to occur three factors have to act simultaneously. Stress, material sensitization and an oxidising environment. Stress may be obtained from boron carbide swelling due to irradiation. Stainless steel may be sensitized to intergranular stress corrosion cracking under irradiation. Normally the reactor environment in a BWR is oxidising. The presentation focuses on findings from hot cell laboratory work on irradiated ABB BWR control rods and studies of irradiated control rod materials in the hot cells at PSI. Apart from physical, mechanical and microstructural examinations, isotope analyses were performed to describe the local isotopic burnup of boron. Consequences (such as possible B4C washout) of a under operation in a ABB BWR, after the occurrence of a crack is discussed based on neutron radiographic examinations of control rods operated with cracks. (author)

  2. Advanced methods for BWR transient and stability analysis

    International Nuclear Information System (INIS)

    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)

  3. Age-related degradation of BWR control rod drives

    International Nuclear Information System (INIS)

    This paper reviews the major age-related degradation mechanisms for U. S. boiling water reactor (BWR) control rod drives (CRDs). Component aging caused by various types of stress corrosion cracking, fatigue, general corrosion, wear, and rubber degradation are discussed. (author)

  4. Decay ratio studies in BWR and PWR using wavelet

    International Nuclear Information System (INIS)

    The on-line stability of BWR and PWR is studied using the neutron noise signals as the fluctuations reflect the dynamic characteristics of the reactor. Using appropriate signal modeling for time domain analysis of noise signals, the stability parameters can be directly obtained from the system impulse response. Here in particular for BWR, an important stability parameter is the decay ratio (DR) of the impulse response. The time series analysis involves the autoregressive modeling of the neutron detector signal. The DR determination is strongly effected by the low frequency behaviour since the transfer function characteristic tends to be a third order system rather than a second order system for a BWR. In a PWR low frequency behaviour is modified by the Boron concentration. As a result of these phenomena there are difficulties in the consistent determination of the DR oscillations. The enhancement of the consistency of this DR estimation is obtained by wavelet transform using actual power plant data from BWR and PWR. A comparative study of the Restimation with and without wavelets are presented. (orig.)

  5. The Advanced BWR Nuclear Plant: Safe, economic nuclear energy

    Energy Technology Data Exchange (ETDEWEB)

    Redding, J.R. [GE Nuclear Energy, San Jose, CA (United States)

    1994-12-31

    The safety and economics of Advanced BWR Nuclear Power Plants are outlined. The topics discussed include: ABWR Programs: status in US and Japan; ABWR competitiveness: safety and economics; SBWR status; combining ABWR and SBWR: the passive ABWR; and Korean/GE partnership.

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

    International Nuclear Information System (INIS)

    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

  7. Development of underwater high-definition camera for the confirmation test of core configuration and visual examination of BWR fuel

    International Nuclear Information System (INIS)

    The purpose of this study is to develop underwater High-Definition camera for the confirmation test of core configuration and visual examination of BWR fuels in order to reduce the time of these tests and total cost regarding to purchase and maintenance. The prototype model of the camera was developed and examined in real use condition in spent fuel pool at HAMAOKA-2 and 4. The examination showed that the ability of prototype model was either equaling or surpassing to conventional product expect for resistance to radiation. The camera supposes to be used in the dose rate condition of under about 10 Gy/h. (author)

  8. Characteristics of fluctuating pressure generated in BWR main steam lines

    International Nuclear Information System (INIS)

    The BWR-3 steam dryer in the Quad Cities Unit 2 Nuclear Power Plant was damaged by high cycle fatigue due to acoustic-induced vibration. The dryer failure was as attributed to flow-induced acoustic resonance at the stub pipes of safety relief valves (SRVs) in the main steam lines (MSLs). The acoustic resonance was considered to be generated by interaction between the sound field and an unstable shear layer across the closed side branches with SRV stub pipes. We have started a research program on BWR dryers to develop their loading evaluation methods. Moreover, it has been necessary to evaluate the dryer integrity of BWR-5 plants which are the main type of BWR in Japan. In the present study, we used 1/10-scale BWR tests and analyses to investigate the flow-induced acoustic resonance and acoustic characteristics in MSLs. The test apparatus consisted of a steam dryer, a steam dome and 4 MSLs with 20 SRV stub pipes. A finite element method (FEM) was applied for the calculation of three-dimensional wave equations in acoustic analysis. We demonstrated that remarkable fluctuating pressures occurred in high and low frequency regions. High frequency fluctuating pressures was generated by the flow-induced acoustic resonance in the SRV stub pipes. Low frequency fluctuating pressure was generated in an MSL with the dead leg. The frequency of the latter almost coincided with the natural frequency of the MSL with the dead leg. The amplitude of the fluctuating pressures in the multiple stub pipes became more intense because of interaction between them compared with that in the single stub pipe. Acoustic analysis results showed that the multiple stub pipes caused several natural frequencies in the vicinity of the natural frequency of the single stub pipe and several modes of the standing wave in the MSLs. (author)

  9. Nuclear transmutation characteristics of reduced moderation BWR (Thesis)

    International Nuclear Information System (INIS)

    In the present thesis, the nuclear transmutation characteristics of reduced moderation BWR, which decides the spent fuel characteristics and its safety in its nuclear fuel cycle, were investigated and compared with other types of reactors. The major conclusions were obtained as follows: The decay heat and radioactivity from FPs increases in fuel burn-up. However, they which normalized with burn-up are small for the reactor with low specific power and long operation period due to the decay during the long operation period. Breeder type of reduced moderation BWR shows low decay heat and radioactivity from FPs because of the long operation period approximately 3000 days which realized by the high conversion ratio. That also shows low decay heat and radioactivity from actinide nuclides due to the hard spectrum. MA recycling reactor of high conversion type of reduced moderation BWR was designed. The neptunium, which has large impact for environmental burden from the viewpoint of nuclide transport analysis, can be incinerated approximately 40% of loaded inventory which corresponds to 22 units of LWR per year. LLFP (99Tc, 129I, 135Cs) transmutation by breeder type of reduced moderation BWR was estimated. As a result, the support factor cannot be lower than unity for each LLFP nuclides. In other words, the reduced moderation BWR cannot reduce LLFP because the LLFP target cannot be loaded inner of the reactor core due to the small margin of core specification. It is expected that these results and the characteristics of other types of reactor shown in the present study benefit the discussion for various nuclear fuel cycle options. (author)

  10. BWR 90+ - Nuclear power plant for 21st century

    International Nuclear Information System (INIS)

    BWR 90+ is a boiling water reactor, based on the previous models BWR90 and BWR75, and on the operational experiences gained with six reactors of the previous generation. The development work started in 1994 in co-operation with Teollisuuden Voima Oy (TVO). At present all the boiling water reactor owners participate the cooperation. The objectives of the development were: (1) to develop a boiling water reactor of competitive price level and short construction time, and which meets the latest safety requirements, (2) to itemize the technologies improving the security and competitivity of present plants, and (3) to maintain the expertise of the personnel of the companies participating the development work, and improving the BWR- technology. High power output and short construction time reduce the power generation costs. Large amount of fuel assemblies leads to higher safety margins. Reduction of scram groups from 18 to 16 reduces the amount of components, the assembly space and costs. The reactor technical data is as follows: Thermal power output 4250 MWth; electric power output 1500 MWe, construction time 1500 days, costs 1500 pounds/kWe, no. of fuel assemblies 872, no. of scram groups 16, turbines 1, the capacity factor 90% and the duration of service outage 3 weeks. Specific features of BWR90+ are: short construction time and low costs, risk for connection between wet and dry spaces has been minimized, reactor core remains covered by water during loss-of-coolant accident caused by fuel replacement, Passive collection and cooling of core-melt inside the containment, the containment is not the first wall against the spreading of core-melt, steam explosions and core- concrete interactions have low probabilities, high gas- volume of wet-space reduces the pressure increase during a severe accident, filter-equipped gas removal system forms the final overpressure shield, the containment is cylindrical, and the plant is equipped with digital instrumentation and control

  11. Prognostic significance of a formalin-resistant nuclear proliferation antigen in mammary carcinomas as determined by the monoclonal antibody Ki-S1.

    OpenAIRE

    Kreipe, H.; Alm, P.; Olsson, H; Hauberg, M.; L. Fischer; Parwaresch, R

    1993-01-01

    Proliferative activity is a potential prognostic indicator of neoplastic cell growth. Usually the assessment of the tumor growth fraction requires specially processed or frozen tissue. We have raised a monoclonal antibody, Ki-S1, suitable for the detection of proliferating cells in routinely processed and paraffin-embedded tissue specimens. A retrospective study was conducted on 83 mammary carcinoma patients with a median follow-up of 45.6 months. Ki-S1 positivity was significantly (P < 0.000...

  12. Guidance for the application of an assessment methodology for innovative nuclear energy systems. INPRO manual - Proliferation resistance. Vol. 5 of the final report of phase 1 of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)

    International Nuclear Information System (INIS)

    The International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) was initiated in the year 2000, based on a resolution of the IAEA General Conference (GC(44)/RES/21). The main objectives of INPRO are (1) to help to ensure that nuclear energy is available to contribute in fulfilling energy needs in the 21st century in a sustainable manner, (2) to bring together both technology holders and technology users to consider jointly the international and national actions required to achieve desired innovations in nuclear reactors and fuel cycles; and (3) to create a forum to involve all relevant stakeholders that will have an impact on, draw from, and complement the activities of existing institutions, as well as ongoing initiatives at the national and international level. The INPRO manual is comprised of an overview volume (No. 1), and eight additional volumes covering the areas of economics (Volume 2), infrastructure (Volume 3), waste management (Volume 4), proliferation resistance (laid out in this report) (Volume 5), physical protection (Volume 6), environment (Volume 7), safety of nuclear reactors (Volume 8), and safety of nuclear fuel cycle facilities (Volume 9). This volume of the INPRO manual is based on the results of an INPRO study on proliferation resistance of the DUPIC fuel cycle performed by the Republic of Korea during 2005 and 2006, recommendations from IAEA consultancy meetings, and on a special service agreement with G. Pshakin (Russian Federation). The INPRO Manual starts with an introduction in Chapter 1. In Chapter 2, the necessary information is described to perform an INPRO assessment in the area of proliferation resistance. Explanatory notes on the INPRO basic principles (BP) and user requirements (UR) in the area of proliferation resistance, are reproduced in Chapter 3 to provide context for the assessor; additionally, background of each criterion (CR) and a corresponding procedure is described how to perform an INPRO assessment. The

  13. Innovative nuclear fuel cycle technologies meeting user's requirements in the area of economy, safety, environment and waste, proliferation-resistance and cross-cutting issues

    International Nuclear Information System (INIS)

    Full text: The Technical Specialists Meeting (TSM) organized by International Atomic Energy Agency (IAEA) at Vienna during April 2-4, 2003, was part of the Agency's 'Innovative Nuclear Reactors and Fuel Cycles Programme' (INPRO), to enable Member State organisations, stakeholders and policy makers to exchange reliable up-to-date information on current and future trends in nuclear fuel cycle technologies. In the TSM, some 36 experts from Member States Argentina, Belgium, Canada, China, France, Germany, India, Japan, Republic of Korea and Russian Federation, European Union (EU), World Nuclear Association (WNA) and IAEA participated. In the Opening Session, IAEA activities in the area of nuclear fuel cycle technology were outlined. This was followed by 4 Technical Sessions covering (i) Front-end of Fuel Cycle; (ii) Energy Conversion; (iii) Back-end of Fuel Cycle; and (iv) Innovative Nuclear Fuel Cycle Concepts where 21 papers were presented. In the Panel Discussion and Concluding Session, thrust areas were identified and a few recommendations were made for consideration of IAEA. The TSM brought out: (i) the status of uranium resources worldwide in terms of deposits, demand, supply, market and current price, types of mining activities, the major uranium producers and enrichment plants; (ii) the advanced water-cooled reactors, FBRs, HTRs and ADS; (iii) the extraordinary diversity of fuel cycle options - from once-through fuel cycle involving relatively inexpensive LEU technology (SIGMA) to plutonium breeding and utilization in LWRs, PHWRs and FBRs and thorium cycles for breeding U233 and their utilization in AHWR and from direct disposal to partitioning and transportation; (iv) the intrinsic and extrinsic features of proliferation-resistance; (v) the need for economic indicators and critical inter-comparison of fossil fuel and nuclear power industries on one hand and wind, solar and other renewable sources of energy on the other; and (vi) IAEA's complementary role in US

  14. Aldo-keto reductase 1B10 promotes development of cisplatin resistance in gastrointestinal cancer cells through down-regulating peroxisome proliferator-activated receptor-γ-dependent mechanism.

    Science.gov (United States)

    Matsunaga, Toshiyuki; Suzuki, Ayaka; Kezuka, Chihiro; Okumura, Naoko; Iguchi, Kazuhiro; Inoue, Ikuo; Soda, Midori; Endo, Satoshi; El-Kabbani, Ossama; Hara, Akira; Ikari, Akira

    2016-08-25

    Cisplatin (cis-diamminedichloroplatinum, CDDP) is one of the most effective chemotherapeutic drugs that are used for treatment of patients with gastrointestinal cancer cells, but its continuous administration often evokes the development of chemoresistance. In this study, we investigated alterations in antioxidant molecules and functions using a newly established CDDP-resistant variant of gastric cancer MKN45 cells, and found that aldo-keto reductase 1B10 (AKR1B10) is significantly up-regulated with acquisition of the CDDP resistance. In the nonresistant MKN45 cells, the sensitivity to cytotoxic effect of CDDP was decreased and increased by overexpression and silencing of AKR1B10, respectively. In addition, the AKR1B10 overexpression markedly suppressed accumulation and cytotoxicity of 4-hydroxy-2-nonenal that is produced during lipid peroxidation by CDDP treatment, suggesting that the enzyme acts as a crucial factor for facilitation of the CDDP resistance through inhibiting induction of oxidative stress by the drug. Transient exposure to CDDP and induction of the CDDP resistance decreased expression of peroxisome proliferator-activated receptor-γ (PPARγ) in MKN45 and colon cancer LoVo cells. Additionally, overexpression of PPARγ in the cells elevated the sensitivity to the CDDP toxicity, which was further augmented by concomitant treatment with a PPARγ ligand rosiglitazone. Intriguingly, overexpression of AKR1B10 in the cells resulted in a decrease in PPARγ expression, which was recovered by addition of an AKR1B10 inhibitor oleanolic acid, inferring that PPARγ is a downstream target of AKR1B10-dependent mechanism underlying the CDDP resistance. Combined treatment with the AKR1B10 inhibitor and PPARγ ligand elevated the CDDP sensitivity, which was almost the same level as that in the parental cells. These results suggest that combined treatment with the AKR1B10 inhibitor and PPARγ ligand is an effective adjuvant therapy for overcoming CDDP resistance of

  15. Fissile material disposition and proliferation risk

    International Nuclear Information System (INIS)

    The proliferation risk of a facility is dependent on the material attractiveness, level of safeguards, and physical protection applied to the material in conjunction with an assessment of the impact of the socioeconomic circumstances and threat environment. Proliferation risk is a complementary extension of proliferation resistance. The authors believe a better determination of nuclear material proliferation can be achieved by establishing the proliferation risk for facilities that contain nuclear material. Developing a method that incorporates the socioeconomic circumstances and threat environment inherent to each country enables a global proliferation assessment. In order to effectively reduce the nuclear danger, a broadly based set of criteria is needed that provides the capability to relatively assess a wide range of disposition options/facilities in different countries and still ensure a global decrease in proliferation risk for plutonium

  16. BWR-GALE, Radioactive Gaseous and Liquid Waste Release from BWR

    International Nuclear Information System (INIS)

    1 - Description of program or function: 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. 2 - Method of solution: GALE calculates expected releases based on 1) standardized coolant activities derived from ANS Standards 18.1 Working Group recommendations, 2) release and transport mechanisms that result in the appearance of radioactive material in liquid and gaseous waste streams, 3) plant-specific design features used to reduce the quantities of radioactive materials ultimately released to the environs, and 4) information received on the operation of nuclear power plants. 3 - Restrictions on the complexity of the problem: The liquid release portion of GALE uses subroutines taken from the ORIGEN (CCC-217) to calculate radionuclide buildup and decay during collection, processing, and storage of liquid radwaste. Memory requirements for this part of the program are determined by the large nuclear data base accessed by these subroutines

  17. Operation and fuel design strategies to minimise degradation of failed BWR fuel

    International Nuclear Information System (INIS)

    Degradation of failed fuel may result in forced shutdown of the reactor to extract the failed fuel. If this occurs during a time when the price of electricity is high, the cost for this forced shutdown may be very costly. The objective of this paper is to point out the impact of fuel design and also operation strategy on the tendency of failed fuel degradation. The following number of items are discussed in the paper: Failure causes: The dominating causes are debris fretting, PCI and crud/water chemistry related defects. It is recommended to adopt the goal, maximum one defect per year per million rods in the core and to achieve the zero-failure goal for PCI. Models for secondary failure development: Two different secondary degradation scenarios can develop, circumferential cracks or breaks and axial cracks. Models for describing the propagation of secondary defects are given and discussed. The secondary degradation tendency can be delayed and minimized by using fuel cladding with improved corrosion resistance such as cladding with large secondary phase particles and high iron content in the liner layer. Also, the spacer design has a large impact on the tendency for transversal break formation. A spacer that catches the debris at the lower part of the fuel assembly will reduce the risk of getting transversal breaks. On the other hand a spacer that catches the debris in the upper part of the fuel assembly will result in a significant risk of developing transversal breaks in low and intermediate burnup fuel. A new model for data analyses - BwrFuelRelease: A new model, BwrFuelRelease, is presented. This model is an efficient tool for analyses of measured off-gas and reactor water data. The model can replace all currently used methods for analyses of fuel failures. By this model it is possible to detect very small defects, to quantify with high precision the amount of Fissile materials on the core surfaces during operation both with non-defected core and during

  18. Importance of momentum effects in BWR reactor vessel modeling

    International Nuclear Information System (INIS)

    The most severe class of events (other than design-basis LOCA's) for boiling water reactors (BWR's) entails rapid pressurization of the reactor vessel (RV) due to rapid valve closures in the main steamlines. The severity of these events is caused by the fact that the pressurization of the core results in a rapid decrease in the core void fraction and, coupled to a strong negative void coefficient of reactivity, in a rapid power increase if the reactor protective system is slow in responding. The DYNODE-B program models the nuclear steam supply system of BWR's. Early versions only explicitly calculated the RV dome pressure. The latest version permits optional explicit calculation of the dome, core outlet, and core average pressures. The use of the new option has shown the influence of the two-phase momentum effects on the core pressure and flow transients resulting from pressure wave reflections at the liquid (incompressible) core inlet region

  19. Condensate polishing guidelines for PWR and BWR plants

    International Nuclear Information System (INIS)

    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. Diagnosis of nonlinear BWR oscillations using TRAC/BF1

    International Nuclear Information System (INIS)

    The nonlinear nature of boiling water reactor (BWR) stability has been demonstrated in both experimental tests and lumped parameter calculational models. Point kinetic reactivity feedback is nonlinear because of its functional dependence on fuel temperature and moderator density. The TRAC/BF1 model used in this analysis differs from a lumped parameter model in its spatial extent. The model, intended to be consistent with a BWR/4, was developed with four active fuel channel components representing one hot, two average, and one peripheral bundles. The vessel internals were modeled explicitly. These internals include lower and upper plena, separator/dryers, core shroud, and dryer skirt. The jet pump/recirculation system is modeled in an azimuthally symmetric fashion. The feedwater and steam line boundary conditions are based on time-dependent data representative of that observed during the LaSalle oscillation event

  1. Coupled BWR calculations with the numerical nuclear reactor software system

    International Nuclear Information System (INIS)

    The Numerical Nuclear Reactor (NNR) is a software suite for integrated high-fidelity reactor core simulations including neutronic and thermal-hydraulic feedback. Using solution modules with formulations to reflect the multi-dimensional nature of the system, NNR offers a comprehensive core modeling capability with pin-by-pin representation of fuel assemblies and coolant channels. Originally developed for pressurized water reactors, the NNR analysis capabilities have recently been extended for boiling water reactor (BWR) applications as part of EPRI Fuel Reliability Program. The neutronics methodology is extended to treat non-periodic structure of BWR fuel assemblies, and a new Eulerian two-phase CFD boiling heat transfer model has been integrated with the software system. This paper summarizes the experience with, and results of, the first-of-a-kind coupled calculations as demonstration of a fully-integrated, high-fidelity simulation capability for assessment of margin to crud-induced failure from fuel-duty perspective. (authors)

  2. Development of internal CRD for next generation BWR

    International Nuclear Information System (INIS)

    In order to develop a competitive and high performance Next Generation BWR with fossil power plant, an internal CRD using a heatproof ceramics insulated coil is under development. In case of a 1700MWe next generation BWR, the internal CRDs are installed in a RPV whose size is equivalent to the 1356 MWe ABWR, and there will be no space required for CRDs and CRD exchange under RPV. These advantages realize a compact PCV and reduced volume of a reactor building. Moreover, the internal CRDs eliminate penetration via a bottom flange of RPV, and lower installation level of RPV in a drywell. This brings further advantages of elimination of RIA (Reactivity Induced Accidents) caused by CR withdrawing under pressure boundary broken, and easy IVR (In Vessel Retention) by vessel bottom cooling in case of a severe accidents. (author)

  3. Development of internal CRD for next generation BWR

    International Nuclear Information System (INIS)

    In order to develop a competitive and high performance Next Generation BWR with fossil power plant, an internal CRD using a heatproof ceramics insulated coil has been developed. In case of a 1700MWe next generation BWR, the internal CRDs are installed in a RPV whose size is equivalent to the 1356 MWe ABWR, and there will be no space required for CRDs and CRD exchange under RPV. These advantages realize a compact PCV and reduced volume of a reactor building. Moreover, the internal CRDs eliminate penetration via a bottom flange of RPV, and lower installation level of RPV in a drywell. This brings further advantages of elimination of RIA (Reactivity Induced Accidents) caused by CR withdrawing under pressure boundary broken, and easy IVR (In Vessel Retention) by vessel bottom cooling in case of severe accidents. (authors)

  4. Study on fast spectrum BWR core for actinide recycle

    International Nuclear Information System (INIS)

    A study has been performed on neutronic and thermal characteristics of a fast spectrum BWR core with tight fuel lattice for an innovative fuel cycle system named BARS (BWR with an Advanced Recycle System) aiming at Pu multi-recycling and MA's (Minor Actinides) burning. The coolant void reactivity tends to shift to positive direction due to the fast neutron spectrum, so the neutron streaming channel and axial enrichment distribution was introduced in the core design to overcome the tendency. It was found that negative void reactivity is attained for whole burnup cycle by using Monte Carlo code. As for the thermal characteristics, the critical power measurement test was performed in order to get the database and the result was compared with the Arai's critical power correlation. As the result, Arai's correlation can predict our critical power test data considerably. Through our core design study and benchmark tests, we will make the conceptual design of BARS core successfully. (authors)

  5. Neutron noise analysis of BWR using time series analysis

    International Nuclear Information System (INIS)

    The main purpose of this paper is to give more quantitative understanding of noise source in neutron flux and to provide a useful tool for the detection and diagnosis of reactor. The space dependent effects of distributed neutron flux signals at the axial direction of two different strings are investigated by the power contribution ratio among neutron fluxes and the incoherent noise spectra of neutron fluxes derived from autoregressive spectra. The signals are measured on the medium sized commercial BWR of 460 MWe in Japan. From the obtained results, local and global noise sources in neutron flux are discussed. This method is indicated to be a useful tool for detection and diagnosis of anomalous phenomena in BWR. (orig./RW)

  6. Corrosion of Inconel X750 in simulated BWR core environment

    International Nuclear Information System (INIS)

    The corrosion tests of Inconel X750 in the simulated core environment by using an irradiated test loop were performed. From the experiments, the followings were clarified; 1. corrosion rate was proportional to minus cubic root of the time and the value was evaluated to be 19±4 mdm after one year, 2. the value was markedly high compared to other reported experimental data, but it had a good coincidence with the value evaluated from activity balance data in actual BWR plants, 3. the most important chemical specie which affected the corrosion behavior was identified to be hydrogen peroxide, and 4. the rate of specimens pre-filmed in the high temperature atmosphere at 700deg C for 5 hrs was reduced to be one third or less than that of specimens pre-oxidized in steam at 390deg C for 13 hrs which was the same condition for actual BWR fuel springs. (author)

  7. High effective of platinum(IV) complex with adamantylamine in overcoming resistance to cisplatin and suppressing proliferation of ovarian cancer cells in vitro

    Czech Academy of Sciences Publication Activity Database

    Horváth, Viktor; Šindlerová, Lenka; Souček, Karel; Blanářová, Olga; Sova, P.; Kroutil, A.; Žák, F.; Mistr, A.; Turánek, J.; Hofmanová, Jiřina; Kozubík, Alois

    Valtice, 2005. s. 21. [Xenobiochemické symposium /23./. 16.05.2005-19.05.2005, Valtice] R&D Projects: GA MPO(CZ) PZ-Z2/29 Institutional research plan: CEZ:AV0Z50040507 Keywords : ovarian cancer * overcoming resistance * cisplatin Subject RIV: BO - Biophysics

  8. Electrochemical potential measurements under simulated BWR water chemistry conditions

    International Nuclear Information System (INIS)

    Laboratory studies have been performed to investigate the stainless steel corrosion potential under simulated BWR coolant chemistry conditions. In addition to dissolved oxygen and hydrogen, test parameters also included chemical additives, metallic ions and hydrogen peroxide at various concentrations. The effect of water flow velocity was also investigated under various water chemistry conditions. The details of test results have been described elsewhere, and the highlights of the investigation are summarized in this paper. (J.P.N.)

  9. Thermohydraulic stability coupled to the neutronic in a BWR

    International Nuclear Information System (INIS)

    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 Laguna Verde

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

    International Nuclear Information System (INIS)

    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

  11. BWR radiation control: plant demonstration. Volume 2. Appendixes

    International Nuclear Information System (INIS)

    The first year's progress is presented for a four-year program intended to implement and evaluate BRAC radiation reduction operational guidelines at the Vermont Yankee BWR and to document the results in sufficient detail to provide guidance to other BWR owners. Past operational, chemistry and radiation level data have been reviewed to provide a historical base of reference. Extensive sampling and chemistry monitoring systems have been installed to evaluate plant chemistry status and the effects of program implemented changes. Radiation surveys and piping gamma scans are being performed at targeted locations to quantify radiation level trends and to identify and quantify piping isotopics. Contact radiation levels on the recirculation line at Vermont Yankee have been increasing at a rate of 175 mR/h-EFPY since 1978. A materials survey of feedwater and reactor components in contact with the process liquid has been performed to identify sources of corrosion product release, particularly cobalt and nickel. A feedwater oxygen injection system has been installed to evaluate the effects of oxygen control on feedwater materials corrosion product releases. A baseline performance evaluation of the condensate treatment and reactor water cleanup systems has been completed. Data on organics and ionics at Vermont Yankee have been obtained. A methodology of BWR feedwater system layup during extended outages was developed, and an evaluation performed of layup and startup practices utilized at Vermont Yankee during the fall 1980 and 1981 refueling outages

  12. Valuation of BWR stability operating in natural circulation conditions

    International Nuclear Information System (INIS)

    Nowadays, the design of reactors having appropriate stability margins, the adoption of operating procedures avoiding possible unstable regions and the development of mitigation strategies to cope with inadvertent instability occurrences have strongly limited safety concerns in this regard. However, despite the obvious need for plant-specific Probabilistic Safety Assessment (PSA), BWR (boiling water reactor) transients of general interest can be identified and characterized as for example, overpressurization events, Large Break Loss of Coolant Accidents (LBLOCAs), feedwater temperature decrease, increase of core flow, main circulation pump flow rate increase, control rod withdrawal and others. Simulations of these complex scenarios have been improved by the utilization of coupled thermal-hydraulic (TH) and neutron kinetics (NK) system codes. In this work, the RELAP5/MOD3.3 thermal-hydraulic system code and the PARCS/2.4 3D neutron kinetic code have been adopted to predict the Peach Bottom BWR stability during recirculation pumps trip while the reactor is operating in a special region of power and core flow map. In the recirculation pump trip event, the stopping of the recirculation pumps causes a sharp decrease in the core flow, which generates a considerable negative reactivity insertion that tends to reduce power and, consequently, the amount of steam generated. The BWR reactor stability has been valuated during natural circulation conditions after the pumps trip event. The time evolution of the power and the related thermal-hydraulic parameters were investigated to analyse the behavior of the reactor for this special operation condition. (author)

  13. Verification of a BWR code package by gamma scan measurements

    International Nuclear Information System (INIS)

    High-burnup 8 x 8 fuel with a large central water rod (called step 2 fuel) has been recently introduced to the latest Japanese boiling water reactor (BWR) plants. Lanthanum-140 gamma intensity is almost directly related to nodal powers. By gamma scan measurement, the axial distribution of 140La in the exposed fuel was measured at the end of cycle (EOC) 1 and was compared with the calculation by a BWR code package TGBLA/LOGOS. The multienrichment fuel-type core (MEC) design was adopted for the initial cycle core of the plants. The MEC design contains three different enrichment types of fuels to simulate the equilibrium cycles, achieve much higher discharge exposure, and save fuel cycle cost, and the low-enrichment fuels are loaded in periphery and in control cells. Such MEC design could be a challenge to the BWR design methods because of the large spectrum mismatch among the fuel assemblies of the different enrichments. The aforementioned comparison has shown that the accuracy of the TGBLA/LOGOS code package is satisfactory

  14. 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.

  15. 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

  16. Matrine inhibits the proliferation, invasion and migration of castration-resistant prostate cancer cells through regulation of the NF-κB signaling pathway.

    Science.gov (United States)

    Li, Qi; Lai, Yiming; Wang, Chengbin; Xu, Guibin; He, Zheng; Shang, Xiaohong; Sun, Yi; Zhang, Fan; Liu, Leyuan; Huang, Hai

    2016-01-01

    Matrine is a naturally occurring alkaloid extracted from the Chinese herb Sophora flavescens. It has been demonstrated to exhibit antiproliferative properties, promote apoptosis and inhibit cell invasion in a number of cancer cell lines. It has also been shown to improve the efficacy of chemotherapy when it is combined with other chemotherapy drugs. However, the therapeutic efficacy of matrine for prostate cancer remains poorly understood. In the present study, we showed that matrine inhibited the proliferation, migration and invasion of both DU145 and PC-3 cells in a dose- and time-dependent manner. It also reduced the cell population at S phase and increased the cell population at sub-G1 phase. The increases in both the apoptotic cell population and cell population at S and sub-G1 phases consistently indicated a pro-apoptotic effect of matrine. Decreases in levels of P65, p-P65, IKKα/β, p-IKKα/β, IKBα and p-IKBα as detected by immunoblot analysis in the matrine-treated DU145 and PC-3 cells suggested an involvement of the NF-κB signaling pathway. Therefore, it is a novel promising addition to the current arsenal of chemotherapy drugs for the treatment of androgen-independent prostate cancer. PMID:26497618

  17. 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 ...

  18. Characterization of solidified radioactive wastes produced at Montalto di Castro BWR plant with reference to the site storage

    International Nuclear Information System (INIS)

    The cement solidification of the Montalto di Castro BWR plant radwastes has been studied both from the point of view of the mixtures of formulation and of the product characterization. Five radwaste types and mixtures of them have been taken into consideration, determining the best chemical formulations starting from the compressive strenght as leading parameter. The solidified products have been characterized from the point of view of the freeze and thawing resistance, the water immersion resistance, the leachability, the dimensional changes and the free standing water. All the tests have been performed taking into account the real site conditions, so the leaching tests and the water immersion tests have been carried out using sea water and table water as leachant

  19. Analysis of a BWR MKI utilizing advanced BMI-2104 computational techniques to calculate source terms

    International Nuclear Information System (INIS)

    The effects of incorporating code modifications to more realistically represent accident phenomenology were examined with specific focus on the BWR MKI AE γ-sequence. Results indicate that the source terms are only a fraction of those reported in BMI-2104 and that core/concrete interaction is not a significant source term contributor for the BWR MKI containments. (author)

  20. The use of the partial coherence function technique for the investigation of BWR noise dynamics

    International Nuclear Information System (INIS)

    The extensive experimental investigations, at the last time, indicate that the partial coherence function technique can be a powerful method of the investigation of BWR noise dynamics. Symple BWR noise dynamics model for the global noise study, based on different noise phenomena, is proposed in this paper. (author)

  1. Simplified distributed parameters BWR dynamic model for transient and stability analysis

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa-Paredes, Gilberto [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana-Iztapalapa, Av. San Rafael Atlixco, 186, Col. Vicentina, Mexico, DF 09340 (Mexico)]. E-mail: gepe@xanum.uam.mx; Nunez-Carrera, Alejandro [Comision Nacional de Seguridad Nuclear y Salvaguardias, Doctor Barragan 779, Col. Narvarte, Mexico, DF 03020 (Mexico); Vazquez-Rodriguez, Alejandro [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana-Iztapalapa, Av. San Rafael Atlixco, 186, Col. Vicentina, Mexico, DF 09340 (Mexico)

    2006-09-15

    This paper describes a simplified model to perform transient and linear stability analysis for a typical boiling water reactor (BWR). The simplified transient model was based in lumped and distributed parameters approximations, which includes vessel dome and the downcomer, recirculation loops, neutron process, fuel pin temperature distribution, lower and upper plenums reactor core and pressure and level controls. The stability was determined by studying the linearized versions of the equations representing the BWR system in the frequency domain. Numerical examples are used to illustrate the wide application of the simplified BWR model. We concluded that this simplified model describes properly the dynamic of a BWR and can be used for safety analysis or as a first approach in the design of an advanced BWR.

  2. 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; Nygaard, Tobias; Hvid, Lars Grøndahl; Wernbom, Mathias; Suetta, Charlotte Arneboe; Frandsen, Ulrik

    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...

  3. 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.

  4. 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.

  5. Prony's method application for BWR instabilities characterization

    International Nuclear Information System (INIS)

    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

  6. BWR and PWR chemistry operating experience and perspectives

    International Nuclear Information System (INIS)

    It is well recognized that proper control of water chemistry plays a critical role in ensuring the safe and reliable operation of Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). State-of-the-art water chemistry programs reduce general and localized corrosion of reactor coolant system, steam cycle equipment, and fuel cladding materials; ensure continued integrity of cycle components; and reduce radiation fields. Once a particular nuclear plant component has been installed or plant system constructed, proper water chemistry provides a global tool to mitigate materials degradation problems, thereby reducing the need for costly repairs or replacements. Recognizing the importance of proper chemistry control and the value in understanding the relationship between chemistry guidance and actual operating experience, EPRI continues to collect, monitor, and evaluate operating data from BWRs and PWRs around the world. More than 900 cycles of valuable BWR and PWR operating chemistry data has been collected, including online, startup and shutdown chemistry data over more than 10 years (> 20 years for BWRs). This paper will provide an overview of current trends in BWR and PWR chemistry, focusing on plants in the U.S.. Important chemistry parameters will be highlighted and discussed in the context of the EPRI Water Chemistry Guidelines requirements (i.e., those parameters considered to be of key importance as related to the major goals identified in the EPRI Guidelines: materials integrity; fuel integrity; and minimizing plant radiation fields). Perspectives will be provided in light of recent industry initiatives and changes in the EPRI BWR and PWR Water Chemistry Guidelines. (author)

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

    OpenAIRE

    Javier Ortiz-Villafuerte; Andrés Rodríguez-Hernández; Enrique Araiza-Martínez; Luis Fuentes-Márquez; Jorge Viais-Juárez

    2013-01-01

    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 liqui...

  8. Characterization studies of BWR-4 neutron noise analysis spectra

    International Nuclear Information System (INIS)

    Neutron noise analysis measurements were made in three BWR-4 reactors under full-power conditions to determine the noise characterization spectra of the reactors with two different instrument-tube cooling configurations. Both configurations were designed to prevent flow-induced vibration of the instrument tubes and subsequent damage of fuel channel boxes caused by impacts of the tubes with the boxes. Noise spectra from these three reactors were compared with spectra previously obtained prior to changing the instrument-tube cooling configuration, and no evidence of impacting was found

  9. In-situ testing of BWR closure head studs

    International Nuclear Information System (INIS)

    Mechanized ultrasonic inspection of closure head studs often is on the critical path. In German BWR's, a floodcompensator is used which allows human access to the studs despite the water is up to a much higher level. For stud inspection this provides a potential solution to get out of the critical path. However, the space restrictions around the studs due to the geometry of the floodcompensator did not allow the use of the existing manipulators. This paper describes the design of a dedicated compact manipulator of a construction which copes with the restricted space available around the studs

  10. Neutron dosimetry. Environmental monitoring in a BWR type reactor

    International Nuclear Information System (INIS)

    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)

  11. Neutron dosimetry. Environmental monitoring in a BWR type reactor

    International Nuclear Information System (INIS)

    The measurements carried out on reactor dosimetry are applied mainly at the study on the effects of the radiation in the materials of the reactor; a 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 tracks manufactured in the ININ is presented, for the environmental monitoring in penetrations around the primary container of the Unit 1 of the Laguna Verde power plant. The monitoring of neutrons carried out with ends of radiological protection, during those operational tests of the reactor. (author)

  12. Neutron flux and fluence determination for BWR reactors

    International Nuclear Information System (INIS)

    Measurements of gamma emission rates from Fe and Cu dosimeters extracted from a BWR type reactor vessel were carried out in order to determine their total activity. The dosimeter's activity is related to the neutron flux there by taking into account the reactor material's embrittlement caused by neutron bombardment. The dosimeters were taken out after the first reactor operation cycle. From gamma radioactivity measurements of these dosimeters, neutron flux and fluence were calculated. These parameters are used in the determination of shift and adjusted reference temperature values needed for the development of pressure-temperature curves used during reactor operation

  13. Void effects on BWR Doppler and void reactivity feedback

    International Nuclear Information System (INIS)

    The significance of steam voids and control rods on the Doppler feedback in a gadolinia shimmed BWR is demonstrated. The importance of bypass voids when determining void feedback is also shown. Calculations were done using a point model, i.e., feedback was expressed in terms of reactivity coefficients which were determined for individual four-bundle configurations and then appropriately combined to yield reactor results. For overpower transients the inclusion of the void effect of control rods is to reduce Doppler feedback. For overpressurization transients the inclusion of the effect of bypass void wil increase the reactivity due to void collapse. (author)

  14. Corrosion products release from steel surface into BWR water coolant

    Energy Technology Data Exchange (ETDEWEB)

    Kritsky, V.G.; Korolev, A.S.; Berezina, I.G.; Sofyin, M.V.

    1986-02-01

    Factors influencing steel corrosion product release and transfer into a BWR primary circuit have been studied and reported on in this paper. The study of corrosion kinetics and corrosion product release was carried out on the samples tested under RBMK NPP condensate-feedwater cycle conditions, as well as, under test rig conditions. The ratio of corrosion product specific mass, transferred to the water, to the whole corrosion product specific mass of steel, formed under the given conditions was determined and used as a criterion, characterizing the extent of corrosion product transfer from the steel surface into the water.

  15. Development of underwater surface inspection system for BWR suppression chamber

    International Nuclear Information System (INIS)

    In order to inspect underwater paints surface of BWR suppression chamber as an in-service inspection, the underwater surface inspection system has been developed, which consists of magnetic crawler, crud recovery nozzles and special camera unit. The system can move on wall surface of pressure suppression chamber with magnetic crawler to inspect surface paints films under the muddy water with special camera unit after removing surface cruds. Based on inspection results, the paint films inspection criteria to assess degradation of paints films were proposed. (T. Tanaka)

  16. Strengthening the non proliferation regime: French views

    International Nuclear Information System (INIS)

    3 main issues can be identified in the French policy concerning the backing of non proliferation: 1) responding resolutely to proliferation crises, 2) reinforcing substantive efforts to prevent and impede proliferation, and 3) strengthening the non-proliferation regime. The first issue is very important because combating proliferation is vital to the security of all. Concerning the second issue, France attaches particular importance to strengthening specific measures to prevent and check proliferation. Let me mention a few proposals that we put forward: exports need to be controlled more effectively, proliferation activities have to be criminalized, or the development of proliferation-resistant technologies should be supported. Concerning the third issue it means the strengthening of the non-proliferation regime, France proposes several means: -) aiming at the universalization of the additional protocol; -) ensuring that the Agency continues to have sufficient human, financial and technical resources to fulfill its verification mission effectively; -) encouraging the IAEA to make full use of the authority available to it; -) enhancing the use of information relevant to the delivery of the IAEA mandate; and -) sharing more accurate information concerning the breaches of commitments that happen. The paper is followed by the slides of the presentation. (A.C.)

  17. 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)

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

    International Nuclear Information System (INIS)

    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 1018 n/cm2) in the TRIGA Mark III Salazar reactor and separately with Ni+3 ions in a Tandetrom accelerator (E= 4.8 MeV and an ion flux rank of 0.1 to 53 ions/A2). (Author)

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

    International Nuclear Information System (INIS)

    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. BWROPT: A multi-cycle BWR fuel cycle optimization code

    Energy Technology Data Exchange (ETDEWEB)

    Ottinger, Keith E.; Maldonado, G. Ivan, E-mail: Ivan.Maldonado@utk.edu

    2015-09-15

    Highlights: • A multi-cycle BWR fuel cycle optimization algorithm is presented. • New fuel inventory and core loading pattern determination. • The parallel simulated annealing algorithm was used for the optimization. • Variable sampling probabilities were compared to constant sampling probabilities. - Abstract: A new computer code for performing BWR in-core and out-of-core fuel cycle optimization for multiple cycles simultaneously has been developed. Parallel simulated annealing (PSA) is used to optimize the new fuel inventory and placement of new and reload fuel for each cycle considered. Several algorithm improvements were implemented and evaluated. The most significant of these are variable sampling probabilities and sampling new fuel types from an ordered array. A heuristic control rod pattern (CRP) search algorithm was also implemented, which is useful for single CRP determinations, however, this feature requires significant computational resources and is currently not practical for use in a full multi-cycle optimization. The PSA algorithm was demonstrated to be capable of significant objective function reduction and finding candidate loading patterns without constraint violations. The use of variable sampling probabilities was shown to reduce runtime while producing better results compared to using constant sampling probabilities. Sampling new fuel types from an ordered array was shown to have a mixed effect compared to random new fuel type sampling, whereby using both random and ordered sampling produced better results but required longer runtimes.

  1. Managing the aging of BWR control rod drive systems

    International Nuclear Information System (INIS)

    This Phase I Nuclear Plant Aging Research (NPAR) study examines the aging phenomena associated with BWR control and rod drive mechanisms (CRDMs) and assesses the merits of various methods of ''imaging'' this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of the Nuclear Plant Reliability Data System (NPRDS) failure cases attributed to the control rod drive (CRD) system, and (4) personal information exchange with nuclear industry CRDM maintenance experts. The report documenting the findings of this research, NUREG-5699, will be published this year. Nearly 23% of the NPRDS CRD system component failure reports were attributed to the CRDM. The CRDM components most often requiring replacement due to aging are the Graphitar seals. The predominant causes of aging for these seals are mechanical wear and thermal embrittlement. More than 59% of the NPRDS CRD system failure reports were attributed to components that comprise the hydraulic control unit (HCU). The predominant HCU components experiencing the effects of service wear and aging are value seals, discs, seats, stems, packing, and diaphragms

  2. BWROPT: A multi-cycle BWR fuel cycle optimization code

    International Nuclear Information System (INIS)

    Highlights: • A multi-cycle BWR fuel cycle optimization algorithm is presented. • New fuel inventory and core loading pattern determination. • The parallel simulated annealing algorithm was used for the optimization. • Variable sampling probabilities were compared to constant sampling probabilities. - Abstract: A new computer code for performing BWR in-core and out-of-core fuel cycle optimization for multiple cycles simultaneously has been developed. Parallel simulated annealing (PSA) is used to optimize the new fuel inventory and placement of new and reload fuel for each cycle considered. Several algorithm improvements were implemented and evaluated. The most significant of these are variable sampling probabilities and sampling new fuel types from an ordered array. A heuristic control rod pattern (CRP) search algorithm was also implemented, which is useful for single CRP determinations, however, this feature requires significant computational resources and is currently not practical for use in a full multi-cycle optimization. The PSA algorithm was demonstrated to be capable of significant objective function reduction and finding candidate loading patterns without constraint violations. The use of variable sampling probabilities was shown to reduce runtime while producing better results compared to using constant sampling probabilities. Sampling new fuel types from an ordered array was shown to have a mixed effect compared to random new fuel type sampling, whereby using both random and ordered sampling produced better results but required longer runtimes

  3. 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.

  4. Logical model for the control of a BWR turbine

    International Nuclear Information System (INIS)

    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. BWR stability analysis with three-dimensional transient code

    International Nuclear Information System (INIS)

    Recently, neutron flux oscillations of two different modes were observed in several foreign BWR plants. One is core wide oscillation mode which is characterized by a phenomenon that neutron flux oscillates in-phase over a whole core. At La Salle 2 plant (U.S.A.), the amplitude of core wide neutron flux oscillation grew considerably large to result in a reactor scram, which aroused great concern about BWR stability. The other is regional oscillation mode which is characterized by the phenomenon, as typically observed at Caorso plant (Italy), that neutron flux of a half core oscillates out-of-phase to that of the other half core. These neutron flux oscillation phenomena were caused by nuclear-thermal hydraulic coupled instability and requires an evaluation study on oscillation detectability and effect on fuel integrity. Particularly, the regional oscillation mode requires three-dimensional analysis since it may bring about locally large amplitude power oscillation. For this reason, analysis was done with the three-dimensional transient code TOSDYN-2 to study reactor condition which causes the regional oscillation and also to evaluate fuel thermal margin under the neutron flux oscillations of these two instability modes. (author)

  6. Standard Technical Specifications, General Electric plants, BWR/4

    International Nuclear Information System (INIS)

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the BWR Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. 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. This document 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

  7. Analysis of NEACRP 3D BWR core transient benchmark

    International Nuclear Information System (INIS)

    NEACRP BWR cold water injection benchmark is analyzed by two codes: TRAC-BF1/SKETCH-N code system by JAERI, Japan and TRAB-3D code by VTT Energy, Finland. Basic features of the codes are described. Neutronics modules of the codes apply nodal methods; separate calculations are performed to compare their accuracy. Thermal-hydraulics modules are significantly different: TRAC-BF1 uses two-phase two-fluid model, while TRAB-3D applies drift-flux model with four separated equations. A representative set of the global and local reactor parameters is given for both the steady-state and transient conditions. TRAB-3D calculations have been performed with two slip correlations: EPRI and the simple Zuber-Findley correlation. A comparison of the two TRAB results shows the importance of the slip model on some computed reactor parameters. The results of the TRAC-BF1/SKETCH-N and TRAB-3D codes are in a close agreement, especially when the advanced EPRI correlation is used in the TRAB-3D code. The presented data can be useful for assessment of other BWR codes. (author)

  8. BWR Mark II ex-vessel corium interaction analyses

    International Nuclear Information System (INIS)

    This report describes the results of a series of studies conducted to investigate the behavior of core debris within a BWR Mark II containment. These studies focused on the interaction of core debris with concrete and steel structures (downcomers and inpedestal floor drains) within the drywell, the transport of debris through these drains and downcomers into the wetwell, and on debris-water reactions within the wetwell. Estimates of the conditions under which debris would penetrate the in-pedestal drain lines, the time-dependent behavior of the debris within the drain lines, and the amount of debris which might enter the suppression pool via these drain lines are provided. An assessment of the conditions under which the upper lip of the downcomers would be expected to fail (i.e. melt) due to exposure to hot core debris is presented. Finally, the unique characteristics of debris water interactions in Mark II containments are discussed, the existing knowledge base regarding core-concrete debris-water interactions is summarized, and an evaluation of the applicability of the MELCOR 1.80 code's debris-water interaction model to BWR Mark II's is presented

  9. TRAB, a transient analysis program for BWR. Part 1

    International Nuclear Information System (INIS)

    TRAB is a transient analysis program for BWR. The present report describes its principles. The program has been developed from TRAWA-program. It models the interior of the pressure vessel and related subsystems of BWR viz. reactor core, recirculation loop including the upper part of the vessel, recirculation pumps, incoming and outgoing flow systems, and control and protection systems. Concerning core phenomena and all flow channel hydraulics the submodels are one-dimensional of main features. The geometry is very flexible. The program has been made particularly to simulate various reactivity transients, but it is applicable more generally to reactor incidents and accidents in which no flow reversal or no emptying of the circuit must occur below the water level. The program is extensively supplied by input and output capabilities. The user can act upon the simulation of a transient by defining external disturbances, scheduled timevariations for any system variable, by modeling new subsystems, which are representable with ordinary linear differential equations, and by defining relations of functional form between system variables. The run of the program can be saved and restarted. (author)

  10. Standard Technical Specifications, General Electric Plants, BWR/6

    International Nuclear Information System (INIS)

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/4, and documents the positions of the Nuclear Regulatory Commission based on the BWR Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. 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

  11. Low power level safety management of Finnish BWR

    International Nuclear Information System (INIS)

    Good practices in work coordination and safety management have contributed to short refueling outage duration in Finnish BWR plants. Human and organizational factors are considered especially important in the low paper states, which consist of start-up, shut-down and the outage period itself. This originates from the use of external labour during the outage, the number of both contemporary and sequentially linked human actions and the variety of potential ways the personnel can affect the plant state. While the containment barrier does not exist, more organizational and administrative means have to be used in risk management. To promote the safety further, special studies have been carried out. This paper discusses both the low power mode PSA and the studies of work orientation and competence among the operating staff in Olkiluoto BWR plant. An advanced outage control requires also open-minded consideration of potential risks and the means for their reduction. Good results in low power risk management can be reached only by the involvement of both the plant operating and the maintenance staff. A profound safety management is a prerequisite for safe low power states. (author)

  12. BWR stability analysis with the BNL Engineering Plant Analyzer

    International Nuclear Information System (INIS)

    March 9, 1989 instability at the LaSalle-2 Power Plant and more than ninety related BWR transients have been simulated on the BNL Engineering Plant Analyzer (EPA). Power peaks were found to be potentially seventeen times greater than the rated power, flow reversal occurs momentarily during large power oscillations, the fuel centerline temperature oscillates between 1,030 and 2,090 K, while the cladding temperature oscillates between 560 and 570 K. The Suppression Pool reaches its specified temperature limit either never or in as little as 4.3 minutes, depending on operator actions and transient scenario. Thermohydraulic oscillations occur at low core coolant flow (both Recirculation Pumps tripped), with sharp axial or redial fission power peaking and with partial loss of feedwater preheating while the feedwater is flow kept high to maintain coolant inventory in the vessel. Effects from BOP system were shown to influence reactor stability strongly through dosed-loop resonance feedback. High feedwater flow and low temperature destabilize the reactor. Low feedwater flow restabilizes the reactor, because of steam condensation and feedwater preheating in the downcomer, which reduces effectively the destabilizing core inlet subcooling. The EPA has been found to be capable of analyzing BWR stability '' shown to be effective for scoping calculations and for supporting accident management

  13. BWR lower plenum debris bed models for MELCOR

    International Nuclear Information System (INIS)

    Work is underway at Oak Ridge National Laboratory (ORNL) to incorporate certain models of the Boiling Water Reactor Severe Accident Response (BWRSAR) code into a local version of MELCOR. Specifically, the BWR lower plenum debris bed and bottom head response models taken from BWRSAR are being tested within the local MELCOR code structure. Upon successful completion of testing, recommendations for formal adoption of these models will be made to the Nuclear Regulatory Commission (NRC) and to the MELCOR code development staff at Sandia National Laboratories (SNL). The SNL code development staff retain exclusive responsibility for maintaining the configuration control for the official version of MELCOR. The BWR lower plenum debris bed and bottom head response models permit the calculation of heatup, melting, and relocation of the debris after dryout. They predict the response of the lower plenum internal structures and the bottom head as well as the composition and timing of material release from the vessel. They have been previously applied in severe accident analyses for the Containment Performance Improvement (CPI) Program and the Mark I shell survivability study (NUREG/CR-5423), and in recent assessments of candidate accident management strategies. This paper provides a brief description of the purpose and operation of these models. 11 refs., 15 figs., 5 tabs

  14. Aging assessment of BWR control rod drive systems

    Energy Technology Data Exchange (ETDEWEB)

    Greene, R.H.

    1991-01-01

    This study examines the aging phenomena associated with boiling water reactor (BWR) control rod drive mechanisms (CRDMs) and assess the merits of various methods of managing this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of Nuclear Plant Reliability Data System (NPRDS) failure cases attributed to the CRD system, and (4) personal information exchange with industry experts. As part of this study, nearly 3500 NPRDS failure reports have been analyzed to examine the prevailing failure trends for CRD system components. An investigation was conducted to summarize the occurrence frequency of these component failures, discovery methods, reported failure causes, their respective symptoms, and actions taken by utilities to restore component and system service. The results of this research have identified the predominant CRDM failure modes and causes. In addition, recommendations are presented that identify specific actions utilities can implement to mitigate CRDM aging. An evaluation has also been made of certain maintenance practices and tooling which have enabled some utilities to reduce ALARA exposures received from routine CRDM replacement and rebuilding activities. 5 refs., 8 figs., 2 tabs.

  15. Aging assessment of BWR control rod drive systems

    International Nuclear Information System (INIS)

    This study examines the aging phenomena associated with boiling water reactor (BWR) control rod drive mechanisms (CRDMs) and assess the merits of various methods of managing this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of Nuclear Plant Reliability Data System (NPRDS) failure cases attributed to the CRD system, and (4) personal information exchange with industry experts. As part of this study, nearly 3500 NPRDS failure reports have been analyzed to examine the prevailing failure trends for CRD system components. An investigation was conducted to summarize the occurrence frequency of these component failures, discovery methods, reported failure causes, their respective symptoms, and actions taken by utilities to restore component and system service. The results of this research have identified the predominant CRDM failure modes and causes. In addition, recommendations are presented that identify specific actions utilities can implement to mitigate CRDM aging. An evaluation has also been made of certain maintenance practices and tooling which have enabled some utilities to reduce ALARA exposures received from routine CRDM replacement and rebuilding activities. 5 refs., 8 figs., 2 tabs

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

    International Nuclear Information System (INIS)

    A denatured (U-233/Th)O2 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 UO2 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)O2-fueled BWR will likely have reduced operating flexibility. A (U-235/Th)O2-fueled BWR should perform similar to a UO2-fueled BWR under all operating conditions. A (Pu/Th)O2-fueled BWR may have reduced thermal margins and similar accident response and be less stable than a UO2-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

  17. Study of behavior on bonding and failure mode of pressurized and doped BWR fuel rod

    International Nuclear Information System (INIS)

    The study of transient behavior on the bonding and the failure mode was made using the pressurized/doped 8 x 8 BWR type fuel rod. The dopant was mullite minerals consisted mainly of silicon and aluminum up to 1.5 w/o. Pressurization of the fuel rod with pure helium was made to the magnitude about 0.6 MPa. As a reference, the non-pressurized/non-doped 8 x 8 BWR fuel rod and the pressurized/7 x 7 BWR fuel rod up to 0.6 MPa were prepared. Magnitude of energy deposition given to the tested fuel rods was 248, 253, and 269 cal/g·fuel, respectively. Obtained results from the pulse irradiation in NSRR are as follows. (1) It was found from the experiment that alternation of the fuel design by the adoption of pressurization up to 0.6 MPa and the use of wider gap up to 0.38 mm could avoid the dopant BWR fuel from the overall bonding. The failure mode of the present dopant fuel was revealed to be the melt combined with rupture. (2) The time of fuel failure of the pressurized/doped 8 x 8 BWR fuel defected by the melt/rupture mode is of order of two times shorter than that of the pressurized/ 7 x 7 BWR defected by the rupture mode. Failure threshold of the pressurized/doped 8 x 8 BWR BWR tended to be lower than that of non-pressurized/non-doped 8 x 8 BWR one. Cracked area of the pressurized/doped 8 x 8 BWR was more wider and magnitude of oxidation at the place is relatively larger than the other tested fuels. (3) Failure mode of the non-pressurized/ 8 x 8 BWR fuel rod was the melt/brittle accompanied with a significant bonding at failed location. While, failure mode of the pressurized/ 7 x 7 BWR fuel rod was the cladding rupture accompanied with a large ballooning. No bonding at failed location of the latter was observed. (author)

  18. Water chemistry control practices and data of the European BWR fleet

    International Nuclear Information System (INIS)

    Nineteen BWR plants are in operation in Europe, nine built by ASEA Atom, six by Siemens KWU and four by General Electric. This paper gives an overview of water chemistry operation practices and parameters of the European BWR plants. General design characteristics of the plants are described. Chemistry control strategies and underlying water chemistry guidelines are summarized. Chemistry data are presented and discussed with regard to plant design characteristics. The paper is based on a contract of the European BWR Forum with AREVA on a chemistry sourcebook for member plants. The survey of chemistry data was conducted for the years 2002 to 2008. (author)

  19. 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)

  20. System control model of a turbine for a BWR; Modelo del sistema de control de una turbina para 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.; Delfin L, A. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)], e-mail: rodolfo.amador@inin.gob.mx

    2009-10-15

    In this work is presented a design of a control system of a turbine for a nuclear power plant with a BWR like energy source. The model seeks to implement later on at thermal hydraulics code of better estimate RELAP/SCDAPSIM. The model is developed for control and protection of turbine, and the consequent protection to the BWR, considering that the turbine control could be employed for one or several turbines in series. The quality of present designs of control pattern of turbine it is that it considers the parameters more important in the operation of a turbine besides that is has incorporated at control the secondary parameters that will be activated originally as true when the turbine model is substituted by a model more detailed. The development of control model of a turbine will be good in short and medium term to realize analysis about the operation of turbine with different operation conditions, of vapor extraction specific steps of turbine to feed other equipment s, besides analyzing the separate effect and integrated effect. (Author)

  1. Irradiated behavior for BWR advanced Zr alloy (HiFi alloy)

    International Nuclear Information System (INIS)

    Irradiation tests of BWR advanced Zr alloys (HiFi alloy) were carried out in a Japanese commercial reactor and the irradiation properties of the material were investigated. HiFi alloy and Zry-2 showed excellent resistance to corrosion up to 70 GWd/t, and furthermore, HiFi kept lower hydrogen pickup compared with Zry-2. As a result of TEM observation, the Fe/(Fe+Cr) ratio of Zr(Fe,Cr)2 type second phase particles (SPPs) for HiFi alloy and Zry-2 tended to decrease as fast neutron fluence increased and to saturate in high fluence. Zr-Fe-Cr SPPs did not completely disappear for even 6 cycles for the irradiated HiFi alloy and Zry-2. In order to clarify the mechanism of hydrogen absorption, the electrochemical technique was applied for oxide film of both materials as part of the out-of-pile test. The relation between the oxide surface potential and the hydrogen fraction was estimated. From the relation, it was thought that the potential difference over the oxide film suppressed hydrogen (proton) diffusion in the oxide film. In addition, a DHC test was carried out to estimate the extent of hydrogen embrittlement using Zry-2 tube. The hydrogen redistribution and concentration of hydride at the crack tip were observed under a tensile load of 100-200 N in elastic deformation region, which contributed to propagation of the crack length. (author)

  2. 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.

  3. IASCC susceptibility under BWR conditions of welded 304 and 347 stainless steels

    Energy Technology Data Exchange (ETDEWEB)

    Castano, M.L. [CIEMAT, Complutense 22, 28040 Madrid (Spain); Schaaf, B. van der [NRG, Petten (Netherlands); Roth, A. [Framatome ANP, Erlangen (Germany); Ohms, C. [JRC-IE, Petten (Netherlands); Gavillet, D. [PSI, Villigen (Switzerland); Dyck, S. van [SCK - CEN, Mol (Belgium)

    2004-07-01

    In-service cracking of Boiling Water Reactors (BWR) and Pressurized Water Reactors (PWR) internal components has been attributed to Irradiation Assisted Stress Corrosion Cracking (IASCC), a high temperature degradation process that austenitic stainless steels exhibit, when subjected to stress and exposed to relatively high fast neutron flux. Most of the cracking incidents in BWRs were associated to the heat-affected zone (HAZ) of welds. Although the maximum end-of- life dose for this structure is about 3 x 10{sup 20} n/cm{sup 2}, below the threshold fluence of 5 x 10{sup 20} n/cm{sup 2} (equivalent to {approx} 1 dpa) for IASCC in BWR of annealed materials, the influence of neutron irradiation in the weld and HAZ is still an open question. As a consequence of the welding process, residual stresses, microstructural and microchemical modifications are expected. In addition, exposure to neutron irradiation can induce variations in the material's characteristics that can modify the stress corrosion resistance of the welded components. While the IASCC susceptibility of base materials is being widely studied in many international projects, the specific conditions of irradiated weldments are rarely assessed. The INTERWELD project, partially financed by the 5. Framework program of the European Commission, was defined to elucidate neutron radiation induced changes in the HAZ of austenitic stainless steel welds that may promote intergranular cracking. To achieve this goal the evolution of residual stresses, microstructure, micro-chemistry, mechanical properties and the stress corrosion behaviour of irradiated materials are being evaluated. Fabrication of appropriate welds of 304 and 347 stainless steels, representative of core components, was performed. These weld materials were irradiated in the High Flux Reactor (HFR) in Petten to two neutron dose levels, i.e. 0.3 and 1 dpa. Complete characterization of the HAZ of both materials, before and after irradiation is

  4. Plant operation performance improvements of the General Electric (GE) boiling water reactors (BWR'S)

    International Nuclear Information System (INIS)

    This paper summarizes some of the plant operation performance improvement techniques developed by the General Electric Company Nuclear Energy Business Operation for the General Electric Boiling Water Reactors (GE BWR's). Through the use of both thermal and plant hardware operating margins, substantial additional flexibility in plant operation can be achieved resulting in significant improvements in plant capacity and availability factor and potential fuel cycle economics for the currently operating or requisition GE BWR plants. This list of techniques includes expanding the BWR thermal power/moderator flow operating domain to the maximum achievable region, operation with a single recirculation loop out of service and operation at rated thermal power with reduced feedwater temperatures. These plant improvements and operating techniques can potentially increase plant capacity factor by 1% to 2% and provide additional fuel cycle economics savings to the GE BWR's owners

  5. Rapid dewatering of Powdex resins at a BWR

    International Nuclear Information System (INIS)

    For most BWR's a large portion of their radioactive waste produced is water demineralization resins, both powdered and bead. In order to minimize the quantities of resins produced, proper demineralizer operation, volume reduction and minimization techniques are relevant to spent resin dewatering and packaging. To meet burial requirements spent resin needs to be dewatered and packaged properly. Methods of dewatering spent resins have included centrifuge separation and pulling water out of the resin with a diaphragm pump. Various vendors are offering systems that provide rapid dewatering and volume reduction using filter/liners in combination with vacuum pumps and air blowers. The various systems required a standardized test program for proper comparison and evaluation. The program is described in this paper

  6. Radial nodalization effects on BWR [boiling water reactor] stability calculations

    International Nuclear Information System (INIS)

    Computer simulations have shown that stability calculations in boiling water reactors (BWRs) are very sensitive to a number of input parameters and modeling assumptions. In particular, the number of thermohydraulic regions (i.e., channels) used in the calculation can affect the results of decay ratio calculations by as much as 30%. This paper presents the background theory behind the observed effects of radial nodalization in BWR stability calculations. The theory of how a radial power distribution can be simulated in time or frequency domain codes by using ''representative'' regions is developed. The approximations involved in this method of solution are reviewed, and some examples of the effect of radial nodalization are presented based on LAPUR code solutions. 2 refs., 4 figs., 2 tabs

  7. Cobra-TF simulation of BWR bundle dry out experiments

    International Nuclear Information System (INIS)

    The COBRA-TF computer code uses a two-fluid, three-field and three-dimensional formulation to model a two-phase flow field in a specific geometry. The liquid phase is divided in a continuous liquid field and a separate dispersed field, which is used to describe the entrained liquid drops. For each space dimension, the code solves three momentum equations, three mass conservation equations and two energy conservation equations. Entrainment and depositions models are implemented into the code to model the mass transfer between the two liquid fields. This study presents the results obtained with COBRA-TF for the simulation of the Siemens 9-9Q BWR Bundle Dryout experiments. The model includes 20 channels and 34 axial nodes in the heated section. The predicted critical power and dryout location is compared with the measured values. An assessment of the code entrainment and de-entrainment models is presented. (authors)

  8. Study of transient turbine shot without bypass in a BWR

    International Nuclear Information System (INIS)

    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)

  9. Siemens Nuclear Power Corporation experience with BWR and PWR fuels

    International Nuclear Information System (INIS)

    The large data base of fuel performance parameters available to Siemens Nuclear Power Corporation (SNP), and the excellent track record of innovation and fuel reliability accumulated over the last twenty-three years, allows SNP to have a clear insight on the characteristics of future developments in the area of fuel design. Following is a description of some of SNP's recent design innovations to prevent failures and to extend burnup capabilities. A goal paramount to the design and manufacture of BWR and PWR fuel is that of zero defects from any case during its operation in the reactor. Progress has already been made in achieving this goal. This paper summarized the cumulative failure rate of SNP fuel rod through January 1992

  10. An ecological interface design for BWR nuclear power plants

    International Nuclear Information System (INIS)

    An ecological interface design was applied to realize the support function for the operator's direct perception and analytical reasoning in the development of an intelligent man-machine system for BWR nuclear power plants. The abstraction-aggregation functional hierarchy representation of the work domain is a base of the ecological interface design. Another base is the concept of the level of cognitive control. The former was mapped into the interface to externalize the operator's normative mental model of the plants, which will reduce his/her cognitive work load and support knowledge-based problem solving. In addition, the same framework can be used for the analytical evaluation of man-machine interfaces. The information content and structure of a prototype interface were evaluated. This approach seems promising from these experiences. (author)

  11. Recent training technology of BWR operators using full scope simulators

    International Nuclear Information System (INIS)

    Nuclear power plants are being operated at high standards now in Japan. There are far few opportunities for operators to perform in challenging situations. To maintain skill and refinement, the simulator training is indispensable for them. BWR Operator Training Center (BTC) provides training courses according to the grade and duty of the operators. The training force constitutes of personnel from utilities', manufacturers' and also BTC-hired personnel. One of the big features of BTC training is composite team type. In this form of training, men from different plants make a team and help each other study. On the human factor viewpoint, error experience on simulators is one of the important items. Training on recognizing subtle symptom is an example of a recent development. Team training for actual crew is effective from various viewpoints. (author)

  12. Influence of local regulations on TN dual purpose BWR casks

    International Nuclear Information System (INIS)

    Transnucleaire (Paris, France) and Transnuclear, Inc. (Hawthorne, New York, United Sates) have both developed Pressurized Water Reactor (PWR) spent fuel casks for storage and transport purposes. The products are supplied in Europe by Transnucleaire and in the United States by Transnuclear, Inc. Now the TN Group is working on a design for Boiling Water Reactor (BWR) spent fuel assemblies: the TN 52 L cask is designed for transport and storage is Switzerland, the TN 68 cask is designed for transport and storage in the United States. For storage purpose, national regulatory requirements have to be met: each country has specific demands and criteria. As a consequence, differences between the TN 52 L design and the TN 68 design for rather similar contents appear in several fields: the design work, the licensing process, the manufacturing and the operational life. (author)

  13. Development of jet pump inspection equipments in BWR

    International Nuclear Information System (INIS)

    This paper describes development of the remotely operated equipments for jet pump ultrasonic testing (UT) in boiling water reactors (BWRs) to enhance the availability of operating nuclear power plants. Stress corrosion cracking (SCC) in the reactor internals has been a major concern in the BWR in recent years. The developed equipments can accomplish the appropriate positioning precision as an application of the Toshiba phased array immersion UT technique and enhance the jet pump inspection performance with a shorter duration and reducing the load for the installation of them. Three types of inspection equipments are developed to cover the outside and inside of the jet pump inlet mixer and the diffuser without disassembling the inlet mixer and the outside of the jet pump riser elbow. Their configurations and specifications are shown in the paper respectively. (author)

  14. TRAB - A transient analysis program for BWR. Part 2

    International Nuclear Information System (INIS)

    TRAB is a transient analysis code for BWRs developed at the Technical Research Centre of Finland. It models the phenomena in the interior of the BWR pressure vessel and in related subsystems. The core model of TRAB can be used separately for LWR modelling. For PWR modelling the core model of TRAB is connected to circuit model SMABRE to form the SMATRA code. This report is a user's manual and documents the structure, contents and preparation of input for TRAB. The structure of TRAB input is very flexible, featuring input groups and subgroups identified with keywords and given in any order as well as data items in free format, freely mixed with explanatory texts. Users interface of the code can be used for modelling within input: through normal input it is possible to create new submodels. These may be functional or tabulated dependencies of the code variables, different types of delays, or ordinary linear differential equations

  15. Development of methodology for early detection of BWR instabilities

    International Nuclear Information System (INIS)

    Full text of publication follows: The objective of the work presented in this paper research, which is supported by the US Department of Energy under the NEER program, is to develop an early anomaly detection methodology in order 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 model of coupled (two-phase) thermal-hydraulic and neutron flux dynamics, based on the US NRC coupled code TRACE/PARCS, is being utilized as a generator of time series data for anomaly detection at an early stage. The concept of the methodology is based on the fact that nonlinear systems show bifurcation, which is a change in the qualitative behavior as the system parameters vary. Some of these parameters may change on their own accord and account for the anomaly, while certain parameters can be altered in a controlled fashion. The non-linear, non-autonomous BWR system model considered in this research exhibits phenomena at two time scales. Anomalies occur at the slow time scale while the observation of the dynamical behavior, based on which inferences are made, takes place at the fast time scale. It is assumed that: (i) the system behavior is stationary at the fast time scale; and (ii) any observable non-stationary behavior is associated with parametric changes evolving at the slow time scale. The goal is to make inferences about evolving anomalies based on the asymptotic behavior derived from the computer simulation. However, only sufficient changes in the slowly varying parameter may lead to detectable difference in the asymptotic behavior. The need to detect such small changes in parameters and hence early detection of an anomaly motivate the utilized stimulus-response approach. In this approach, the model

  16. Core concept for long operating cycle simplified BWR (LSBWR)

    International Nuclear Information System (INIS)

    An innovative core concept for a long operating cycle simplified BWR (LSBWR) is currently being developed under a Toshiba Corporation and Tokyo Institute of Technology joint study. In this core concept, the combination of enriched uranium oxide fuels and loose-pitched lattice is adopted for an easy application of natural circulation. A combination of enriched gadolinium and 0.7-times sized small bundle with peripheral-positioned gadolinium rod is also adopted as a key design concept for 15-year cycle operation. Based on three-dimensional nuclear and thermal hydraulic calculation, a nuclear design for fuel bundle has been determined. Core performance has been evaluated based on this bundle design and shows that thermal performance and reactivity characteristics meet core design criteria. Additionally, a control rod operation plan for an extension of control rod life has been successfully determined. (author)

  17. Dynamic safety systems in BWR plant safety systems

    International Nuclear Information System (INIS)

    Dynamic Safety Systems (DSSs) are reactor safety function systems that are functionally controlled using dynamic rather than static processes. All components including software, whose failure could result in a critical safety system failure, are operationally verified by hard-wired components. Dynamic Safety Systems have been enveloped in the United Kingdom by AEA Technology for use in gas cooled reactors. One such system, known as ISAT trademark, is described in this paper. Through use of scenario testing of a DDS emulator on a Boiling Water Reactor plant training simulator described in this paper. Through use of scenario testing of a DSS emulator on a Boiling Water Reactor plant training simulator, it is shown that a DSS can provide a cost effective safety system in BWR power plants

  18. Control method for water quality of BWR type reactor

    International Nuclear Information System (INIS)

    The present invention provides a method of suppressing radiation exposure upon periodical inspection of a BWR type reactor, suppressing leaching of radioactive materials deposited and activated on fuels, and reducing radioactive deposition on pipelines and equipments made of a carbon steel and austenite stainless steel. Namely, control of water quality described below is conducted under the conditions that the Ni metal ion concentration is from 2 to 10ppb and the Zn metal ion concentration of from 3 to 15ppb in reactor water. (1) controlling the water quality based on neutral/purified water during normal operation and upon injection of hydrogen, (2) using fuels having spring members made of a Ni based alloy processed by aging hardening in atmospheric air, (3) using reactor water recycling pipelines made of an electrolyzed and polished austenite stainless steel, and (4) using carbon steel or low alloy steel for pipelines and equipments of a reactor system. (I.S.)

  19. Fault tree analysis on BWR core spray system

    International Nuclear Information System (INIS)

    Fault Trees which describe the failure modes for the Core Spray System function in the Browns Ferry Nuclear Plant (BWR 1065MWe) were developed qualitatively and quantitatively. The unavailability for the Core Spray System was estimated to be 1.2 x 10-3/demand. It was found that the miscalibration of four reactor pressure sensors or the failure to open of the two inboard valves (FCV 75-25 and 75-53) could reduce system reliability significantly. It was recommended that the pressure sensors would be calibrated independently. The introduction of the redundant inboard valves could improve the system reliability. Thus this analysis method was verified useful for system analysis. The detailed test and maintenance manual and the informations on the control logic circuits of each active component are necessary for further analysis. (author)

  20. BWR Full Integral Simulation Test (FIST). Phase I test results

    International Nuclear Information System (INIS)

    A new full height BWR system simulator has been built under the Full-Integral-Simulation-Test (FIST) program to investigate the system responses to various transients. The test program consists of two test phases. This report provides a summary, discussions, highlights and conclusions of the FIST Phase I tests. Eight matrix tests were conducted in the FIST Phase I. These tests have investigated the large break, small break and steamline break LOCA's, as well as natural circulation and power transients. Results and governing phenomena of each test have been evaluated and discussed in detail in this report. One of the FIST program objectives is to assess the TRAC code by comparisons with test data. Two pretest predictions made with TRACB02 are presented and compared with test data in this report

  1. BWR Full Integral Simulation Test (FIST). Phase I test results

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, W S; Alamgir, M; Sutherland, W A

    1984-09-01

    A new full height BWR system simulator has been built under the Full-Integral-Simulation-Test (FIST) program to investigate the system responses to various transients. The test program consists of two test phases. This report provides a summary, discussions, highlights and conclusions of the FIST Phase I tests. Eight matrix tests were conducted in the FIST Phase I. These tests have investigated the large break, small break and steamline break LOCA's, as well as natural circulation and power transients. Results and governing phenomena of each test have been evaluated and discussed in detail in this report. One of the FIST program objectives is to assess the TRAC code by comparisons with test data. Two pretest predictions made with TRACB02 are presented and compared with test data in this report.

  2. BWR plant analyzer development at BNL [Brookhaven National Laboratory

    International Nuclear Information System (INIS)

    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

  3. Obtention control bars patterns for a BWR using Tabo search

    International Nuclear Information System (INIS)

    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 tiempotabu 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)

  4. Experimental investigation of the enthalpy- and mass flow-distribution in 16-rod clusters with BWR-PWR-geometries and conditions

    International Nuclear Information System (INIS)

    The enthalpy- and mass-flow-distribution at the outlet of two different 16-rod cluster test sections with uniform heating in axial and radial direction under steady state conditions has been measured for the first time by simultaneous sampling of 5 from 6 present characteristic subchannels in the bundle using the isokinetic technique and analysing the outlet quantities by a calorimetic method. The test-sections are provided with typical geometrical configurations for BWR s (70 bars; test section PELCO-S) and PWR s (160 bars; test-section EUROP). The latter has also been tested under BWR conditions (70 bars) to study the influence of geometry and pressure. The results showed the abnormal behaviour of the corner subchannel under BWR typical conditions (70 bars) which could not be found for PWR conditions (160 bars) and which is only an effect of pressure and not of geometry. The analysis of the experimental data confirms the usefullness of the subchannel sampling technique for the better understanding of the complex thermohydraulic phenomena under two-phase flow conditions in multirod bundles. Calculations of subchannel resistance coefficients for both types of spacers under one-phase flow conditions have been made with a special sub-structure method which showed a rather high local value of the corner subchannel. With the local drag coefficents the total resistance of the spacer has been evaluated and agreed well with measured values under adiabatic conditions. The measured subchannel data permit a direct valuation and examination of respective computer codes in a fundamental manner which are, however, not subject of this report

  5. Summary report of seismic PSA of BWR model plant

    International Nuclear Information System (INIS)

    This report presents a seismic PSA (Probabilistic Safety Assessment) methodology developed at the Japan Atomic Energy Research Institute (JAERI) for evaluating risks of nuclear power plants (NPPs) and the results from an application of the methodology to a BWR plant in Japan, which is termed Model Plant'. The seismic PSA procedures developed at JAERI are to evaluate core damage frequency (CDF) and have the following four steps: (1) evaluation of seismic hazard, (2) evaluation of realistic response, (3) evaluation of component capacities and failure probabilities, and (4) evaluation of conditional probability of system failure and CDF. Although these procedures are based on the methodologies established and used in the United States, they include several unique features: (1) seismic hazard analysis is performed with use of available knowledge and database on seismological conditions in Japan; (2) response evaluation is performed with a response factor method which is cost effective and associated uncertainties can be reduced with use of modern methods of design calculations; (3) capacity evaluation is performed with use of test results available in Japan in combination with design information and generic capacity data in the U.S.A.; (4) systems reliability analysis, performed with use of the computer code SECOM-2 developed at JAERI, includes identification of dominant accident sequences, importance analysis of components and systems as well as the CDF evaluation with consideration of the effect of correlation of failures by a newly developed method based on the Monte Carlo method. The effect of correlation has been recognized as an important issue in seismic PSAs. The procedures was used to perform a seismic PSA of a 1100 MWe BWR plant. Results are shown as well as the insights derived and future research needs identified in this seismic PSA. (J.P.N.)

  6. Spectral effects in cavitation of BWR jet pumps

    International Nuclear Information System (INIS)

    BWR jet-pumps vibrate violently when the recirculation drive flow becomes sufficiently high that cavitation occurs in the throat section at the local temperature of the suction flow. This process occurs rapidly as the fluid flows through the nozzle throat and while mixing of the drive and suction flows is in process. Continued mixing during transit through the throat eventually results in a pressure recovery to the saturation pressure and subsequent collapse of the voids near the entrance to the diffuser. The liquid surrounding a vapor volume, or void, performs work during void collapse. The result is that energy is deposited in the liquid and converted locally to an over-pressure, or sonic wave. The power contained in this wave may be very large, since it is deposited in the relatively short void collapse time. The frequency spectrum of the sonic excitation depends on the temporal nature of the collapse, including multiple events when the void is composed of a swarm of bubbles of various sizes. The frequency spectrum of a single spherical void collapse is calculated from the bubble collapse dynamics as it propagates in an increasing ambient pressure field that is a function of drive flow rate, Qd. It is shown that the void collapse is a tandem process in the time frame of the bubble swarm. These considerations lead to comparisons with strain-gauge data taken in BWR tests with the head off. Averaged micro-strain data is compared with the calculated values at several pump speeds, then fit to the data set. The normalized result is in excellent agreement, indicating that the void collapse mechanism is the most likely source of vibration excitation in the jet pump structure

  7. Spectral effects in cavitation of BWR jet pumps

    Energy Technology Data Exchange (ETDEWEB)

    Terhune, J.H.; Karim-Panahi, K. [GE Nuclear Energy, San Jose, CA (United States)

    1996-12-01

    BWR jet-pumps vibrate violently when the recirculation drive flow becomes sufficiently high that cavitation occurs in the throat section at the local temperature of the suction flow. This process occurs rapidly as the fluid flows through the nozzle throat and while mixing of the drive and suction flows is in process. Continued mixing during transit through the throat eventually results in a pressure recovery to the saturation pressure and subsequent collapse of the voids near the entrance to the diffuser. The liquid surrounding a vapor volume, or void, performs work during void collapse. The result is that energy is deposited in the liquid and converted locally to an over-pressure, or sonic wave. The power contained in this wave may be very large, since it is deposited in the relatively short void collapse time. The frequency spectrum of the sonic excitation depends on the temporal nature of the collapse, including multiple events when the void is composed of a swarm of bubbles of various sizes. The frequency spectrum of a single spherical void collapse is calculated from the bubble collapse dynamics as it propagates in an increasing ambient pressure field that is a function of drive flow rate, Q{sub d}. It is shown that the void collapse is a tandem process in the time frame of the bubble swarm. These considerations lead to comparisons with strain-gauge data taken in BWR tests with the head off. Averaged micro-strain data is compared with the calculated values at several pump speeds, then fit to the data set. The normalized result is in excellent agreement, indicating that the void collapse mechanism is the most likely source of vibration excitation in the jet pump structure.

  8. Development of alternative materials for BWR fuel springs

    International Nuclear Information System (INIS)

    Major sources of radioactivity introduced into reactor water of BWR were estimated fuel crud and in-core materials (especially, fuel springs). Fuel springs are used for fixation of fuel cladding tubes with spacer grid. Those are small parts (total length is only within 25 mm) and so many numbers are loaded simultaneously and then total surfaces area are calculated up to about 200 m2. Fuel springs are located under high radiation field and high oxidative environment. Conventional fuel spring is made of alloy-X750 which is one of nickel-based alloy and is reported to show relatively higher corrosion release rate. 58Co and 60Co will be released directly into reactor water from intensely radio-activated fuel springs surface and increase radioactivity concentrations in primary coolant. Corrosion release control from fuel springs is an important technical item and a development of alternative material instead of alloy-X750 for fuel spring is a key subject to achieve ultra low man-rem exposure BWR plant. In present work, alloy-X718 which started usage for PWR fuel springs and stainless steel type 316L which has many mechanical property data are picked up for alternative materials and compared their corrosion behaviors with conventional material. Corrosion experiment was conducted under vapor-water two phases flow which is simulated fuel cladding surface boiling condition. After exposure, corrosion film formed under corrosion test was analyzed in detail and corrosion film amount and corrosion release amount are estimated among three materials. (authors)

  9. BWR core shroud replacement. A cost-effective alternative

    International Nuclear Information System (INIS)

    Intergranular stress corrosion cracking of stainless steel internals of BWRs is a longstanding and recognized problem. Well-engineered approaches have been developed by utilities and engineering contractors to address many of the known problems. One of these is cracking of the horizontal and vertical welds in the core shrouds. The approach being used world wide to address cracking in the horizontal welds is to install vertical tie rods and supplemental lateral restraints which provide redundant vertical and lateral support of the shroud and thereby structurally replace the function of the shroud horizontal welds. This approach has been thoroughly analyzed, approved by regulatory bodies and implemented in the United States, Europe, and Taiwan in over 16 BWR units. Importantly, many of these tie rod installations have been installed on a preemptive basis as a prudent preventive maintenance action. As BWR plants age, additional cracking of core shrouds, core supports, core spray piping and jet pumps has been experienced at some plants, indicating the need for a cost-effective, technically-acceptable approach for dealing with these problems. One such approach is complete replacement of the core shroud assembly, including core supports and attached core spray piping. This operation also makes the jet pumps accessible for replacement if this is determined to be necessary. A complete replacement by welding has been accomplished in Japan, and provides one acceptable replacement method. An alternative replacement concept, described in this paper, takes advantage of the proven tie rod modification to provide an approach which can be accomplished without in-reactor welding and without draining the reactor vessel. As described in this paper, this approach can be accomplished with significantly reduced radiation exposure to workers, reduced costs, and in a fraction of the time required for the fully-welded approach. An additional benefit is that as-welded joints in the reactor

  10. Tritium in liquid phase in a BWR-5 like Laguna Verde; Tritio en fase liquida en un BWR-5 como Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Tijerina S, F.; Vargas A, A.; Cardenas J, J., E-mail: francisco.tijerina@cfe.gob.mx [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Subgerencia de Ingenieria, Carretera Veracruz-Medellin Km 7.5, Veracruz (Mexico)

    2011-11-15

    In boiling water reactors (BWR), the tritium (H{sub 3}) takes place mainly as a result of ternary fissions in the nuclear reactors, of those which 75% are in gaseous form and 25% in liquid form. In the liquid phase, the tritium is transported to the pipes of the primary coolant toward condensed tanks or tanks of drainage excesses of radioactive equipment, located in external areas of a BWR, as well as to the processes of radioactive wastes to be able to be directed to the liquid effluents. For that reason, is necessary to know the possible routes of the transport and processes of the tritium in a BWR to control this radioisotope in the site of the event of leaks in equipment s and buried pipes, avoiding that emigrates toward underground flowing and an impact to the environment and to the people in general. (Author)

  11. PPARα: Mechanism of species differences and hepatocarcinogenesis of peroxisome proliferators

    International Nuclear Information System (INIS)

    Peroxisome proliferator chemicals are classic non-genotoxic carcinogens. These agents cause liver cancers when chronically administered to rats and mice. Peroxisome proliferators include the widely prescribed lipid and cholesterol lowering fibrate drugs. In contrast to the results in rodents, there is no evidence that fibrates are associated with elevated risk of liver cancer or any other neoplasms in humans thus indicating a species difference in the hepatocarcinogenic response. The biological effects of peroxisome proliferators are mediated by the peroxisome proliferator-activated receptor (PPAR)α. Pparα-null mice are resistant to all of the pleiotropic effects of peroxisome proliferators, including cell proliferation and hepatocarcinogenesis. The mechanism of hepatocellular proliferation involves downregulation of the microRNA let-7c gene by PPARα. Let-7c controls levels of proliferative c-myc by destabilizing its mRNA. Thus, upon suppression of let-7c, c-myc mRNA and protein are elevated resulting in enhanced hepatocellular proliferation. In contrast, PPARα-humanized mice, that respond to Wy-14,643 by lower serum triglycerides and induction of genes encoding fatty acid metabolizing enzymes, are resistant to peroxisome proliferator-induced cell proliferation and cancer. These mice do not exhibit downregulation of let-7c gene expression thus forming the basis for the resistance to hepatocellular carcinogenesis

  12. ROSA-III experimental program for BWR LOCA/ECCS integral simulation tests

    International Nuclear Information System (INIS)

    This is the final report of the ROSA-III experimental program, in which the summary of integral simulation test results is described on thermal-hydraulic behavior during a loss-of-coolant accident (LOCA) of a boiling water reactor (BWR) and on the effectiveness of the emergency core cooling system (ECCS). Also presented in the report are the assessment results of computer codes for the BWR LOCA analysis and of the similarity between ROSA-III test results and thermal-hydraulic phenomena during a BWR LOCA by using ROSA-III test data and code analysis results. The ROSA-III facility is a volumetrically scaled (1/424) BWR system with an electrically heated core consisting of four half-length bundles. Many test series were conducted between April 1978 and March 1983. The similarity between a ROSA-III test and a BWR LOCA concerning the fundamental thermal-hydraulic phenomena has been confirmed for major ROSA-III tests. The accident scenario has been well understood and defined for various break locations and break sizes. The effectiveness of the current BWR ECCS design has been well demonstrated. (author)

  13. Investigation of BWR [boiling water reactor] instability phenomena using RETRAN-03

    International Nuclear Information System (INIS)

    In 1988, LaSalle, a boiling water reactor (BWR)/5, experienced severe flux oscillations following a trip of both recirculation pumps. The flux oscillations were terminated by an automatic scram at 118% of rated neutron flux. As a result of this event, the U.S. Nuclear Regulatory Commission has asked the BWR utilities to develop procedural or hardware changes that will assure protection of all safety limits. The rapid growth of the oscillations at LaSalle, and the fact that previous stability analyses had predicted the plant to be very stable, emphasizes that a better understanding of this phenomenon is needed before the success of the long-term fixes can be assured. The intent of the Electric Power Research Institute's work was to use BWR transient methods to model reactor instabilities and investigate the factors that dominate this phenomenon. The one-dimensional transient code RETRAN-03 (Ref. 1) was used. The following conclusions are drawn: (1) RETRAN has demonstrated the ability to model BWR instability (nonlinear oscillations). (2) The general system behavior predicted by RETRAN in BWR stability analyses matches theoretical prediction and plant data. (3) These one-dimensional, time-domain results have increased the understanding of BWR stability phenomena and have helped optimize the long-term solutions being developed by the utilities

  14. Development of proliferation resistant isotope separation technology

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Doyoung; Ko, Kwanghoon; Kim, Taeksoo; Park, Hyunmin; Lim, Gwon; Cha, Yongho; Han, Jaemin; Baik, Sunghoon; Cha, Hyungki

    2012-02-15

    This project was accomplished with an aim of establishing the industrial facilities for isotope separation in Korea. The experiment for the measurement of neutrino mass that has been an issue in physics, needs very much of enriched calcium-48 isotope. However, calcium-48 isotope can be produced only by the electro-magnetic method and, thus, its price is very expensive. Therefore, we expect that ALSIS can replace the electro-magnetic method for calcium-48 isotope production. In this research stage, the research was advanced systematically with core technologies, such as atomic vapor production, the measurement of vapor characteristics and stable and powerful laser development. These researches will be the basis of the next research stages. In addition, the international research trends and cooperation results are reported in this report.

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

    International Nuclear Information System (INIS)

    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)

  16. Proliferation: myth or reality?

    International Nuclear Information System (INIS)

    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.)

  17. Accumulation of operator workload data by using A-BWR training simulator

    International Nuclear Information System (INIS)

    Human-machine interface (HMI) of A-BWR has been developed in order to improve operational safety, reliability and to reduce workload. A-BWR HMI is fully computerized. JNES (Japan Nuclear Energy Safety Organization) and BTC (BWR Operator Training Center Corporation) have accumulated the operator workload quantitative data, related to the observation and operation at typical transient conditions, in order to evaluate the difference of operational workloads between A-BWR HMI and conventional type HMI. The workload evaluation shows the following results: - The workload density (observation and operation frequencies per unit time) of ABWR just after the plant trip is less than that of BWR-5. - At stable conditions after the transient, the workload density of ABWR becomes higher comparing that of BWR-5. A-BWR alarm system may increase the workload density caused by alarm multi-layer structure, because an operator has to use the flat and/or the CRT display to pursue every alarm. The analysis of shift team training at BTC shows that total workload is reduced at ABWR but alarm confirmation work still remains as burden. These results show some modifications might be needed for future HMI. To grasp the tendency of operator workload difference by the control panel type difference, the operator workload quantitative data have accumulated using ABWR type simulator and conventional type simulator at the same typical transient condition. These data were arranged operation frequency data, number of alarm generating data and number of switching CRT pictures data according to plant behaviour. The ABWR type HMI's characteristic have become clear by these operator workload data and the team characteristic evaluation data which BTC evaluated comparing the team performance difference of HMI type

  18. Director's series on proliferation

    International Nuclear Information System (INIS)

    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

  19. Utility of Social Modeling for Proliferation Assessment - Preliminary Findings

    International Nuclear Information System (INIS)

    Often the methodologies for assessing proliferation risk are focused around the inherent vulnerability of nuclear energy systems and associated safeguards. For example an accepted approach involves ways to measure the intrinsic and extrinsic barriers to potential proliferation. This paper describes preliminary investigation into non-traditional use of social and cultural information to improve proliferation assessment and advance the approach to assessing nuclear material diversion. Proliferation resistance assessment, safeguard assessments and related studies typically create technical information about the vulnerability of a nuclear energy system to diversion of nuclear material. The purpose of this research project is to find ways to integrate social information with technical information by explicitly considering the role of culture, groups and/or individuals to factors that impact the possibility of proliferation. When final, this work is expected to describe and demonstrate the utility of social science modeling in proliferation and proliferation risk assessments.

  20. Proliferation Networks and Financing

    International Nuclear Information System (INIS)

    The objective of this study is to propose practical solutions aimed at completing and strengthening the existing arrangement for the control of nuclear proliferation through a control of financial as well as material or immaterial flows. In a first part, the author proposes a systemic analysis of networks of suppliers and demanders. He notably evokes the Khan's network and the Iraqi acquisition network during the 1993-2001 period. He also proposes a modelling of proliferation networks (supplier networks and acquisition networks) and of their interactions. In a second part, the author examines possible means and policies aimed at neutralising proliferation networks: organisation, adaptation and improvement of intelligence tools in front of proliferation networks, and means, limitations and perspectives of network neutralisation. He also briefly addresses the possibility of military action to contain proliferation flows

  1. The Japanese utilities' requirements for a next century BWR

    International Nuclear Information System (INIS)

    This paper reports on the progress of studies to establish a plant concept for a Boiling Water Reactor (BWR) of the next century. The studies were initiated in 1990 by the Japanese utilities, jointly with NSSS vendors, to investigate evolutionary and long term nuclear power plants. The plant concept is based on the evolution of the ABWR taking advantage of new technology. Fundamental plant philosophies are expressed by the following four desired characteristics: Economical, Benign to human, Simple, Flexible. According to these philosophies, concrete objectives of the plant design are reduction of operating burden and maintenance, increase of safety margin and flexibility to adjust to possible changes in economic circumstances in the years to come. The basic utilities' requirements for the new generation BWR were discussed based on the future social needs and the current operational experiences. Start of operation is to be in the 2010's when the early generation LWRs may need to be replaced. Plant power generation capacity will be about 1500 MWe since this level rating will be achievable by extrapolation of current technology. One important requirement is to achieve power generation costs competitive with other generation methods. An outline of the utilities' requirements follows: Operability; prevent inadvertent reactor scram and engineering safety system actuation due to single failure of normal duty systems or single operator error, achieve same load following capability as ABWR, design for plant availability of up to 90%, achieve plant design life of 60 years, maintain annual inspection period at less than 40 days, reduce maintenance activities in harsh environments, reduce employees' dose to less than that of ABWR, consider 'N+2' design to reduce peak loads during annual inspection. Safety margin; increase grace period for transient and accident events, adopt severe accident countermeasures, keep core damage frequency lower than that of ABWR and conditional

  2. Advanced BWR core component designs and the implications for SFD analysis

    International Nuclear Information System (INIS)

    Prior to the DF-4 boiling water reactor (BWR) severe fuel damage (SFD) experiment conducted at the Sandia National Laboratories in 1986, no experimental data base existed for guidance in modeling core component behavior under postulated severe accident conditions in commercial BWRs. This paper will present the lessons learned from the DF-4 experiment (and subsequent German CORA BWR SFD tests) and the impact on core models in the current generation of SFD codes. The DF-4 and CORA BWR test assemblies were modeled on the core component designs circa 1985; that is, the 8 x 8 fuel assembly with two water rods and a cruciform control blade constructed of B4C-filled tubelets. Within the past ten years, the state-of-the-art with respect to BWR core component development has out-distanced the current SFD experimental data base and SFD code capabilities. For example, modern BWR control blade design includes hafnium at the tips and top of each control blade wing for longer blade operating lifetimes; also water rods have been replaced by larger water channels for better neutronics economy; and fuel assemblies now contain partial-length fuel rods, again for better neutronics economy. This paper will also discuss the implications of these advanced fuel assembly and core component designs on severe accident progression and on the current SFD code capabilities

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

    International Nuclear Information System (INIS)

    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)

  4. Space–time convergence analysis on BWR stability using TRACE/PARCS

    International Nuclear Information System (INIS)

    Highlights: ► Quantify TRACE/PARCS space–time discretization error for BWR stability prediction. ► Establish space and time discretization necessary for space–time converged models. ► Show that the space–time converged model gives more reliable results for both stable and unstable reactor. ► Use of the space–time converged model increases confidence in the prediction of BWR stability. -- Abstract: Unstable behavior of Boiling Water Reactors (BWRs) is known to occur during operation at certain power and flow conditions. Even though BWR instability is not a severe safety concern, it could cause reactor scram and significantly decrease the economic performance of the plant. This paper aims to (a) quantify TRACE/PARCS space–time discretization error for simulation of BWR stability, (b) establish space (nodalization) and time discretization necessary for space–time converged model and (c) show that the space–time converged model gives more reliable results for both stable and unstable reactor. The space–time converged model is obtained when further refinement of numerical discretization parameters (nodalization and time step) has negligible effect on the solution. The study is significant because performing a space–time convergence analysis is a necessary step of qualification of the TRACE/PARCS model, and use of the space–time converged model increases confidence in the prediction of BWR stability.

  5. 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)

  6. Upgrading BWR training simulators for annual outage operation training

    International Nuclear Information System (INIS)

    Based upon the recently developed quality assurance program by the Japanese electric companies, BWR Operator Training Center (BTC) identified the needs to enhance operators' knowledge and skills for operations tasks during annual outage, and started to develop a dedicated operator training course specialized for them. In this paper, we present the total framework of the training course for annual outage operations and the associated typical three functions of our full-scope simulators specially developed and upgraded to conduct the training; namely, (1) Simulation model upgrade for the flow and temperature behavior concerning residual heat removal (RHR) system with shutdown cooling mode, (2) Addition of malfunctions for DC power supply equipment, (3) Simulation model upgrade for water filling operation for reactor pressurization (future development). We have implemented a trial of the training course by using the upgraded 800MW full-scope training simulator with functions (1) and (2) above. As the result of this trial, we are confident that the developed training course is effective for enhancing operators' knowledge and skills for operations tasks during annual outage. (author)

  7. Water injection system for turbine driven BWR type reactor

    International Nuclear Information System (INIS)

    The present invention provides a water injection system of a turbine driven nuclear reactor for maintaining the function thereof even upon occurrence of a severe accident in a BWR type nuclear reactor. That is, the system comprises a differential pressure detection means for measuring a pressure difference between the downstream of a the turbine and a reactor container and an interrupting means for stopping the supply of steams to the turbine when the differential pressure exceeds a predetermined value. With such a constitution, when the pressure in the turbine driven water injection system is locally increased, the differential pressure detection means detects the differential pressure, to interrupt the supply of the steams to the turbine. Further, upon occurrence of a severe accident that a pressure in the reactor container is abnormally elevated, differential pressure is not caused between the downstream of the turbine and the reactor container. Accordingly, a protection function is not operated by the differential pressure detection means. Accordingly, injection of coolants to the reactor can be continued even upon loss of AC power source. (I.S.)

  8. Standard Technical Specifications, General Electric plants, BWR/6

    International Nuclear Information System (INIS)

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the B ampersand W Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. 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. This document 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

  9. Development of BWR operator training simulator and training support systems

    International Nuclear Information System (INIS)

    This paper describes a BWR operator training simulator and training support systems that have been developed with the aim of providing support throughout operator training. The operator training simulator is needed in order to improve simulation fidelity and enlarge simulation scope. A 3-dimensional reactor core model has been developed in order to improve the understanding of operators respecting neutronics through realistic training. A severe accident model has been developed for training operators and technical support center teams respecting plant operation and for studying various phenomena. The severe accident is simulated by connecting the physical parameters continuously from the conventional model to the severe accident model. An emergency procedure guideline support system is adopted in order to improve efficiency of operation training for emergencies, since the emergency operation procedures are complicated and based on multiple parameter conditions. The operator training support system is also introduced so as to help training instructors to evaluate the operation and to give instructions to operators to improve operational accuracy. An instructor's burden is eased by automatically evaluating the operation errors based on signals of a simulator. The effects of these systems are evaluated and found to be effective in an actual training center and in engineers' examinations. (author)

  10. Development and implementation of a BWR digital feedwater control system

    International Nuclear Information System (INIS)

    EPRI and Northern States Power Company (NSP) realized that fault-tolerant digital technology could improve Feedwater Control System reliability and operations. The Digital Feedwater Control System (DFCS) is the first major fault-tolerant digital control application in a nuclear power plant in the US. The microprocessor-based controller replaced the analog controller in the feedwater control loop to improve performance and reliability of control including ease of maintenance and spare parts supply. The DFCS at Monticello plant has been in operation since July 1986 without any failure of the control system. This system replaced and upgraded the main and start-up analog controllers at Monticello BWR. It features automatic control, on-line signal validation, controller self-diagnosis, and fault tolerance. The dual-redundant hardware configuration minimizes spare parts availability problems. At a control room panel, operators select each feedwater valve's operating mode (one or three element control, manual, and so on) or set bias inputs for individual feed-water-valve demand. These and other features permit more exact feedwater control system tuning, improving feedwater control in all modes of plant operation. Signal validation using parity-space techniques isolated failed sensors and permit system switching to accurate sensors, thus avoiding outages. To ensure successful operation of the system, extensive verification and validation effect were conducted. These included design reviews, factory acceptance testing using simulation code, site acceptance testing using full-scale plant simulator, and pre-operational and operational testing at Monticello plant

  11. BWR [boiling water reactor] shutdown margin model in SIMULATE-3

    International Nuclear Information System (INIS)

    Boiling water reactor (BWR) technical specifications require that the reactor be kept subcritical (by some prescribed margin) when at room temperature rodded conditions with any one control rod fully withdrawn. The design of an acceptable core loading pattern may require hundreds or thousands of neutronic calculations in order to predict the shutdown margin for each control rod. Direct, full-core, three-dimensional calculations with the SIMULATE-3 two-group advanced nodal code require 3 to 6 CPU min (on a SUN-4 workstation) for each statepoint/control rod that is computed. Such computing and manpower requirements may be burdensome, particularly during the early core design process. These requirements have been significantly reduced by the development of a fast, accurate shutdown margin model in SIMULATE-3. The SIMULATE-3 shutdown margin model achieves a high degree of accuracy and speed without using axial collapsing approximations inherent in many models. The mean difference between SIMULATE-3 one-group and two-group calculations is approximately - 12 pcm with a standard deviation of 35 pcm. The SIMULATE-3 shutdown margin model requires a factor of ∼15 less CPU time than is required for stacked independent two-group SIMULATE-3 calculations

  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. A New Method for Early Anomaly Detection of BWR Instabilities

    International Nuclear Information System (INIS)

    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.

  14. Standard Technical Specifications, General Electric plants, BWR/4

    International Nuclear Information System (INIS)

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the B ampersand W Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. 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. This document Volume 3, contains the Bases for Sections 3.4--3.10 of the improved STS

  15. Standard Technical Specifications, General Electric Plants, BWR/6

    International Nuclear Information System (INIS)

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the B ampersand W Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. 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. This document 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

  16. BWR Fuel Lattice Design Using an Ant Colony Model

    Energy Technology Data Exchange (ETDEWEB)

    Montes, Jose L.; Ortiz, Juan J. [Instituto Nacional de Investigaciones Nucleares, Depto. de Sistemas Nucleares, Carretera Mexico Toluca S/N. La Marquesa Ocoyoacac. 52750, Estado de Mexico (Mexico); Francois, Juan L.; Martin-del-Campo, Cecilia [Depto. de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico Paseo Cuauhnahuac 8532. Jiutepec, Mor. 62550 (Mexico)

    2008-07-01

    This paper deals with one of the steps of the nuclear fuel design: the radial fuel lattice design. It can be seen as a combinatorial optimization problem for determining the optimal 2D fuel rods enrichment and gadolinia distribution. In order to solve this optimization problem, the ant colony system technique is proposed. The main idea of the ant colony approach consists of emulating the real ant colony behaviour in their searching for minimum paths between two given points, usually between the nest and a food source. In this case, the environment where the artificial ants move is the space defined by the discrete possible values of Gd{sub 2}O{sub 3} contents, the U{sup 235} enrichment, and the valid locations inside the 10x10 BWR fuel lattice array. In order to assess any candidate fuel lattice in the optimization process, the HELIOS neutronic transport code is used. The results obtained in the application of the implemented model show that the proposed technique is a powerful tool to tackle this step of the fuel design. (authors)

  17. Validation study of core analysis methods for full MOX BWR

    International Nuclear Information System (INIS)

    JNES has been developing a technical data base used in reviewing validation of core analysis methods of LWRs in the occasions: (1) confirming core safety parameters of the initial core (one-third MOX core) through a full MOX core in Oma Nuclear Power Plant, which is under the construction, (2) licensing high-burnup MOX cores in the future and (3) reviewing topical reports on core analysis codes for safety design and evaluation. Based on the technical data base, JNES will issue a guide of reviewing the core analysis methods used for safety design and evaluation of LWRs. The data base will be also used for validation and improving of core analysis codes developed by JNES. JNES has progressed with the projects (1) measurements of Doppler reactivity in experimental MOX core simulating LWR cores, (2) measurement of isotopic compositions of fission product nuclides on high-burn up BWR UO2 fuels and (3) neutronics analysis of the experimental data that has been obtained in the international joint programs such as FUBILA and REBUS. (author)

  18. Validation study of core analysis methods for full MOX BWR

    International Nuclear Information System (INIS)

    JNES has been developing a technical data base used in reviewing validation of core analysis methods of LWRs in the coming occasions: (1) confirming the core safety parameters of the initial core (one-third MOX core) through a full MOX core in Oma Nuclear Power Plant, which is under the construction, (2) licensing high-burnup MOX cores in the future and (3) reviewing topical reports on core analysis codes for safety design and evaluation. Based on the technical data base, JNES will issue a guide of reviewing the core analysis methods used for safety design and evaluation of LWRs. The data base will be also used for validation and improving of core analysis codes developed by JNES. JNES has progressed with the projects (1) analysis of the measurement data of Doppler reactivity in experimental MOX core simulating LWR cores, (2) measurements of isotopic compositions of fission product nuclides on high-burnup BWR UO2 fuels and the analysis of the measurement data, and (3) neutronics analysis of the experimental data that has been obtained in the international joint programs such as FUBILA and REBUS. (author)

  19. Natural heat transfer augmentation in passive advanced BWR plants

    International Nuclear Information System (INIS)

    In the European Simplified Boiling Water Reactor (ESBWR), the long-term post-accident containment pressure is determined by the combination of non condensable gas pressure and steam pressure in the wet well gas space. Since there are no active systems for heat removal in the wet well, energy transmitted to the wet well gas space, by a variety of means, must be removed by passive heat transfer to the walls and suppression pool (SP). The cold suppression pool located below the hotter gas space provides a stable configuration in which convection currents are suppressed thus limiting heat and mass transfer between the gas space and pool. However, heat transfer to the walls results in natural circulation currents that can augment the heat and mass transfer to the pool surface. Using a simplified model, parametric studies are carried out to show that augmentation of the order of magnitude expected can significantly impact the heat and mass transfer to the pool. Additionally a review of available literature in the area of augmentation and mixed convection of this type is presented and indicates the need for additional experimental work in order to develop adequate models for heat and mass transfer augmentation in the configuration of a BWR suppression pool. (author)

  20. Quantitative evaluations of BWR regional oscillations using higher harmonics subcriticalities

    International Nuclear Information System (INIS)

    A quantitative study of a mechanism for BWR regional stability has been carried out from the higher harmonics viewpoint using a three-dimensional higher harmonics analysis code and is reported in this paper. The results show that the first azimuthal harmonics subcriticality takes a relatively small value under the regionally unstable condition. The calculated patterns of the first azimuthal harmonics agree quite well with the measured regional oscillation mode patterns. Comparing the subcriticality and the steady state power distribution, it is shown that the distribution exists whose first azimuthal harmonics subcriticality takes a small value. This implies that regionally unstable conditions can be estimated from fundamental mode distributions. The decomposition method of the oscillated power responses into the harmonics modes is presented. The results show that core-wide oscillation power responses almost entirely consist of the fundamental mode and regional oscillation power responses almost entirely consist of the first azimuthal harmonics mode. This indicates that the regional oscillation is the phenomenon in which the first azimuthal harmonics mode oscillates on the basis of the fundamental mode. (author)

  1. Impurity hideout/hideout return at the Susquehanna 2 BWR

    International Nuclear Information System (INIS)

    An impurity hideout return study was performed at the Susquehanna 2 BWR to provide an understanding of impurity hideout processes during normal operation and their impact on high temperature solution chemistry in corrosion product deposits on the fuel. Limited hideout return data obtained during shutdowns at 10 BWRs previously had indicated reasonable consistency with expectations based on MULTEQ high temperature solution chemistry modeling of hideout processes. Observations at Susquehanna 2 were consistent with expectations. Cumulative returns of species forming precipitates at low concentration factors above the bulk water concentration, e.g., calcium, magnesium, sulfate and silica were much greater than those of species having a minimal tendency to precipitate, e.g., sodium and chloride. Solutions present in the fuel cladding surface during normal operation were predicted to contain high concentrations (0.1 to 2 molal) of sodium, potassium, chloride, sulfate, silica and nitrate. The predicted solution pH at 300 degrees C was 9.4 (neutral pH = 5.5). The increase in conductivity observed during and after shutdown was shown to be due to solubilization of precipitates with retrograde solubilities rather than chemical/resin intrusion. Variations in reactor water concentrations during reactor water cleanup system isolation and power reductions were consistent with predictions developed from a mass balance around the reactor coolant system

  2. Calibration of the TVO spent BWR reference fuel assembly

    International Nuclear Information System (INIS)

    In 1989 the Support Programmes of Finland (FSP) and Sweden (SSP) initiated a joint task to cross calibrate the burnup of the IAEA spent BWR reference fuel assembly at the TVO AFR storage facility (TVO KPA-STORE) in Finland. The reference assembly, kept separately under the IAEA seal, is used for verification measurements of spent fuel by GBUV method (SG-NDA-38). The cross calibration was performed by establishing a calibration curve, 244Cm neutron rate versus burnup, using passive neutron assay (PNA) measurements. The declared burnup of the reference assembly was compared with the burnup value deduced from the calibration curve. A calibration line was also established by using the GBUV method with the aid of high resolution gamma ray spectrometry (HRGS). Normalization between the two different facilities was performed using sealed neutron and gamma calibration sources. The results of the passive neutron assay show consistency, better than 1 %, between the declared mean burnup of the reference assembly and the burnup deduced from the calibration curve. The corresponding consistency is within +-2 % for the HRGS measurements

  3. Artificial intelligence applied to fuel management in BWR type reactors

    International Nuclear Information System (INIS)

    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)

  4. Derivation of general scaling criteria for BWR containment tests

    International Nuclear Information System (INIS)

    General top-down scaling criteria for facilities used to study Boiling Water Reactor (BWR) containments including a pressure suppression system are derived, with particular attention to the recent passive BWRS. The criteria are derived by considering the generic processes in classes of containment subsystems (e.g., containment volumes, pools, pipes, etc.). In reactor containments, the thermodynamic behavior of the system (essentially, its pressure history) is linked to its thermal-hydraulic behavior (the flows of mass and energy between volumes). The case of prototypical fluids under prototypical thermodynamic conditions is treated. The study confirms the validity of the (familiar) scaling of power, volumes, horizontal areas in volumes, mass flow rates, and heat transfer areas with a system scale. Important pressure drops and the corresponding flows are controlled by the submergence depth of vents or by hydrostatic pressure differences in connected vessels. The analysis of these processes justify the choice of 1:1 scaling for the pressure drops, vertical heights, submergence depths and level differences. The importance of certain distortions regarding inertial response and transit times is minor

  5. Dry well cooling systems in BWR type nuclear power plants

    International Nuclear Information System (INIS)

    Purpose: To prevent the damages of pipeways due to salt damages at the surface of control rod drives in BWR type reactors. Constitution: In control rod drives and the lowermost area in the dry well in which surface corrosion and pitching have been resulted by the salt contents in air due to the increase in the humidity accompanying the lowering of the temperature, a blower is disposed to the upstream of the cooling coils and a portion of high temperature air returned to the lower cooler is replaced with a low temperature feed air to increase the feed temperature in the area. Further, by upwardly turning the downwarded feed air drawing port in which cold feed air has so far been descended as it is, the descendance of the cold air is suppressed. As a result, temperature lowering in the driving mechanisms and the lower area can be prevented to obtain a predetermined temperature, whereby the dewing on the surface can be prevented and thereby preventing the occurrence of corrosion and pitching. (Horiuchi, T.)

  6. A damage tolerance analysis of an old BWR pressure vessel

    International Nuclear Information System (INIS)

    A program has been completed in Sweden where the oldest of the Swedish BWR-units has been subjected to an extensive investigation regarding signs of aging and degradation. The core and all the internal parts were removed and the stripped pressure vessel was decontaminated. Almost every weld in the vessel including all nozzles in the bottom head have then been inspected by aid of UT-inspection. An important part of the programs was the damage tolerance analysis. It has involved postulated surface cracks and embedded cracks (positioned along and across the welds) in all the inspected welds. By using the R6-method the maximum acceptable and critical crack size have then been determined for the most limiting load case and accounting for the individual material properties of each weld, cladding and base material. Of special interest is the core region. The base material is made of A 302 Grade B with rather high Cu- and Ni-contents, which have caused irradiation embrittlement in the beltline region. This implies that during certain cold loading cases, the critical crack size for a postulated surface crack in the core region, can be quite small. However, for load cases during normal operation, the material is on the upper shelf region and the critical crack sizes are large

  7. Design of a redundant meteorological station for a BWR reactor

    International Nuclear Information System (INIS)

    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)

  8. Nonlinear behavior under regional neutron flux oscillations in BWR cores

    International Nuclear Information System (INIS)

    A three-dimensional time-domain core analysis code was applied to numerical simulations for an actual regional neutron flux oscillation observed in a commercial BWR core, in order to investigate potential nonlinear behavior in its coupled neutronic and thermohydraulic system. The present study shows existence of the nonlinear reactivity interaction between the fundamental and first azimuthal spatial harmonics modes of neutron flux distribution under the regional event. The spectrum analysis of the simulated data provides a unique result, that is, temporal harmonics peaks are excited at the even- and odd-order multiples of the characteristic resonance frequency in the fundamental and first spatial harmonics responses, respectively. The numerical simulation also shows that the strong nonlinearity of the coupled neutronic and thermohydraulic dynamics locally appears where the power unstably oscillates with large amplitudes, inducing the power shift and reactivity bias which are shown in the core-wide situation under the global oscillations. This contributes to suppression of the divergence of the local power oscillation, and also to development of the saturated self-limited cycles under the regional oscillations. (author)

  9. BWR Fuel Lattice Design Using an Ant Colony Model

    International Nuclear Information System (INIS)

    This paper deals with one of the steps of the nuclear fuel design: the radial fuel lattice design. It can be seen as a combinatorial optimization problem for determining the optimal 2D fuel rods enrichment and gadolinia distribution. In order to solve this optimization problem, the ant colony system technique is proposed. The main idea of the ant colony approach consists of emulating the real ant colony behaviour in their searching for minimum paths between two given points, usually between the nest and a food source. In this case, the environment where the artificial ants move is the space defined by the discrete possible values of Gd2O3 contents, the U235 enrichment, and the valid locations inside the 10x10 BWR fuel lattice array. In order to assess any candidate fuel lattice in the optimization process, the HELIOS neutronic transport code is used. The results obtained in the application of the implemented model show that the proposed technique is a powerful tool to tackle this step of the fuel design. (authors)

  10. Identification of the reduced order models of a BWR reactor

    International Nuclear Information System (INIS)

    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)

  11. 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.

  12. Effects of Cr and Nb contents on the susceptibility of Alloy 600 type Ni-base alloys to stress-corrosion cracking in a simulated BWR environment

    International Nuclear Information System (INIS)

    In order to discuss the effects of chromium and niobium contents on the susceptibility of Alloy 600 type nickel-base alloys to stress-corrosion cracking in the BWR primary coolant environment, a series of creviced bent-beam (CBB) tests were conducted in a high-temperature, high-purity water environment. Chromium, niobium, and titanium as alloying elements improved the resistivity to stress-corrosion cracking, whereas carbon enhanced the susceptibility to it. Alloy-chemistry-based correlations have been defined to predict the relative resistances of alloys to stress-corrosion cracking. A strong correlation was found, for several heats of alloys, between grain-boundary chromium depletion and the susceptibility to stress-corrosion cracking

  13. Evaluation of aluminum brass coupons in BWR condensate environment in presence of metal ions

    International Nuclear Information System (INIS)

    Effect of cobalt and cesium ions in the simulated BWR condensate environment (two phase water at 150 °C) on the oxide formed on the aluminium brass has been studied by exposing active and prepassivated coupons in respective environments. Surface changes in the exposed coupons were evaluated by SEM, EDAX and electrochemical studies. The SEM and EDAX data of the exposed coupons indicated marked difference in the surface morphology with varying water chemistry. Presence of nodular grains were seen on SEM images of the pre-passivated Al brass coupons in the Co based media while more granular oxide formation could be seen in presence of Cs. With the mixture of Co and Cs, oxides with larger particle size were seen in the SEM images. The weight change measurement also indicated that Co affects the outer oxide layer to a higher extent as compared to Cs. EDAX measurements indicated incorporation of Co in the oxide layer for the coupons exposed in the Co based media whereas higher aluminum composition was seen in the oxide layer for the coupons exposed in the Cs based media. Cathodic reduction of the oxide layer in sodium perchlorate medium indicated that the oxide grown in only water based media are primarily Cu2O with a minor amount of ZnO but there is a significant amount of Co in the oxide layer for the coupons exposed in the Co based medium. Impedance measurement of the coupons indicated similar protective nature of the passive layers formed under various conditions based on the values of charge transfer resistance obtained by fitting the experimental data to the Randles circuit. However, the higher capacitance values for oxides formed in the Co based medium indicated its porous nature. Thus, there is significant sorption of Co in the passive layer of the aluminium brass while there is no evidence of Cs sorption over aluminium brass could be obtained in the present study. (author)

  14. Getting serious about proliferation

    International Nuclear Information System (INIS)

    The US needs to give a higher priority to nuclear non-proliferation, but Reagan's policies assume that proliferation is inevitable and that it is more important to be a reliable supplier than to cause trade frictions by trading only with those nations which sign the non-proliferation treaty (NPT). This undercuts US leadership and the intent of the agreement. Several bills now before Congress could help to restore US leadership by tightening export restrictions and the use of plutonium from the US

  15. 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

  16. Strengthening the nuclear non-proliferation regime

    International Nuclear Information System (INIS)

    Although the nuclear non-proliferation regime has enjoyed considerable success, today the regime has never been under greater threat. Three states have challenged the objectives of the NPT, and there is a technology challenge - the spread of centrifuge enrichment technology and know-how. A major issue confronting the international community is, how to deal with a determined proliferator? Despite this gloomy scenario, however, the non-proliferation regime has considerable strengths - many of which can be developed further. The regime comprises complex interacting and mutually reinforcing elements. At its centre is the NPT - with IAEA safeguards as the Treaty's verification mechanism. Important complementary elements include: restraint in the supply and the acquisition of sensitive technologies; multilateral regimes such as the CTBT and proposed FMCT; various regional and bilateral regimes; the range of security and arms control arrangements outside the nuclear area (including other WMD regimes); and the development of proliferation-resistant technologies. Especially important are political incentives and sanctions in support of non-proliferation objectives. This paper outlines some of the key issues facing the non-proliferation regime

  17. Supporting non proliferation and global security efforts

    International Nuclear Information System (INIS)

    CEA contributes as a major actor of France's action against nuclear proliferation and to the strengthening of nuclear security at national level as European and International levels, in particular through the support of the IAEA activities in nuclear non proliferation with the French Support Programme for the IAEA safeguards system and security with the contribution to the IAEA Nuclear Security Plan and cooperation projects with the European Commission. The CEA is a French government funded technological research organization, organized around 5 branches: Nuclear Energy, Technological Researches, Defence (DAM), Material Sciences and Life Sciences. Within the scope of its activities, CEA covers most of the research areas and techniques in nuclear non-proliferation and security. The CEA is also the advisor of the French Government on nuclear policy. Treaty monitoring and the development and implementation of non proliferation and global security programs is an important mission of DAM which rely on nuclear weapons manufacture and past testing experience. The programmes on non proliferation and global security carried out to fulfil DAM's mission cover the following areas: development of monitoring and detection methods and equipments, country profiles and nuclear stockpiles assessment, arms control instruments, proliferation resistance of nuclear fuel cycle, monitoring of nuclear tests, operation and maintenance of national detection capabilities and contribution to CTBT verification systems. (A.C.)

  18. 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)

  19. 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)

  20. BWR Water Chemistry Guidelines: 1993 Revision, Normal and hydrogen water chemistry

    International Nuclear Information System (INIS)

    The goal of water chemistry control is to extend the operating life of the reactor and rector coolant system, balance-of-plant components, and turbines while simultaneously controlling costs to safeguard the continued economic viability of the nuclear power generation investment. To further this goal an industry committee of chemistry personnel prepared guidelines to identify the benefits, risks, and costs associated with water chemistry in BWRs and to provide a template for an optimized water chemistry program. This document replaces the BWR Normal Water Chemistry Guidelines - 1986 Revision and the BWR Hydrogen Water Chemistry Guidelines -- 1987 Revision. It expands on the previous guidelines documents by covering the economic implications of BWR water chemistry control

  1. Analysis of a BWR direct cycle forced circulation power plants operation

    International Nuclear Information System (INIS)

    First, it is established a general view over the operational problems of the BWR direct cycle forced circulation power plants, and then it is analysed the possibility of the utilization of the energy purged from the turbine as an additional energy for the electrical generation. To simulate the BWR power plant and to obtain the solution of the mathematical model it was developed a computer code named ATOR which shows the feasibility of the proposed method. In this way it is shown the possibility to get a better maneuvering allowance for the BWR power plant whenever it is permitted a convenient use of the vapor extracted from the turbine for the feedwater pre-heaters of the reactor. (author)

  2. 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.

  3. Design guideline to prevent the pipe rupture by combustion of radiolysis gases in BWR steam piping

    International Nuclear Information System (INIS)

    In late 2001, pipe rupture accidents due to fast combustion of radiolysis gas occurred in Japan and elsewhere's BWR power plants. TENPES began to set up the guideline as action to such a new problem to prevent accumulation and combustion of radiolysis gas in BWR steam piping. And then, the first edition of guideline was published in October 2005, and the 2nd edition in March 2007. Afterwards, the experimental study about combustion/detonation of radiolysis gas have been continued. And in March 2010, JANTI published the 3rd edition of the guideline. This is the report of the final edition of that guideline. According to this guideline, it became possible to design BWR's steam piping to prevent pipe rupture accident due to combustion of radiolysis gas. (author)

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

    International Nuclear Information System (INIS)

    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 Classclose 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

  5. Validation of SCALE (SAS2H) Isotopic Predictions for BWR Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hermann, O.W.

    1998-01-01

    Thirty spent fuel samples obtained from boiling-water-reactor (BWR) fuel pins have been modeled at Oak Ridge National Laboratory using the SAS2H sequence of the SCALE code system. The SAS2H sequence uses transport methods combined with the depletion and decay capabilities of the ORIGEN-S code to estimate the isotopic composition of fuel as a function of its burnup history. Results of these calculations are compared with chemical assay measurements of spent fuel inventories for each sample. Results show reasonable agreement between measured and predicted isotopic concentrations for important actinides; however, little data are available for most fission products considered to be important for spent fuel concerns (e.g., burnup credit, shielding, source-term calculations, etc.). This work is a follow-up to earlier works that studied the ability to predict spent fuel compositions in pressurized-water-reactor (PWR) fuel pins. Biases and uncertainties associated with BWR isotopic predictions are found to be larger than those of PWR calculations. Such behavior is expected, as the operation of a BWR is significantly more complex than that of a PWR plant, and in general the design of a BWR has a more heterogeneous configuration than that of a PWR. Nevertheless, this work shows that the simple models employed using SAS2H to represent such complexities result in agreement to within 5% (and often less than 1%) or less for most nuclides important for spent fuel applications. On the other hand, however, the set of fuel samples analyzed represent a small subset of the BWR fuel population, and results reported herein may not be representative of the full population of BWR spent fuel.

  6. Prediction of BWR performance under the influence of Isolation Condenser-using RAMONA-4 code

    International Nuclear Information System (INIS)

    The purpose of the Boiling Water Reactor (BWR) Isolation Condenser (IC) is to passively control the reactor pressure by removing heat from the system. This type of control is expected to reduce the frequency of opening and closing of the Safety Relief Valves (SRV). A comparative analysis is done for a BWR operating with and without the influence of an IC under Main Steam Isolation Valve (MSIV) closure. A regular BWR, with forced flow and high thermal power, has been considered for analysis. In addition, the effect of ICs on the BWR performance is studied for natural convection flow at lower power and modified riser geometry. The IC is coupled to the steam dome for the steam inlet flow and the Reactor Pressure Vessel (RPV) near the feed water entrance for the condensate return flow. Transient calculations are performed using prescribed pressure set points for the SRVs and given time settings for MSIV closure. The effect of the IC on the forced flow is to reduce the rate of pressure rise and thereby decrease the cycling frequency ofthe SRVS. This is the primary objective of any operating IC in a BWR (e.g. Oyster Creek). The response of the reactor thermal and fission power, steam flow rate, collapsed liquid level, and core average void fraction are found to agree with the trend of pressure. The variations in the case of an active IC can be closely related to the creation of a time lag and changes in the cycling frequency of the SRVS. An analysis for natural convection flow in a BWR indicates that the effect of an IC on its transient performance is similar to that for the forced convection system. In this case, the MSIV closure, has resulted in a lower peak pressure due to the magnitude of reduced power. However, the effect of reduced cycling frequency of the SRV due to the IC, and the time lag between the events, are comparable to that for forced convection

  7. BWR SFAT, gross-defect verification of spent BWR fuel. Final report on Task FIN A563 on the Finnish Support Programme to IAEA Safeguards including BWR SFAT User Manual

    International Nuclear Information System (INIS)

    A measurement instrument called Spent Fuel Attribute Tester, SFAT, has been designed, fabricated and taken into use by the IAEA in gross defect verification of spent BWR fuel assemblies. The equipment consists of an underwater measurement head connected with cables to a control unit on the bridge of the fuel handling machine as well as to a PMCA for measurement of the gamma spectra. The BWR SFAT is optimized for the AFR interim storage, TVO KPA-STORE, of the TVO Power Company in Olkiluoto, Finland. It has a shape and it is moved like a fuel assembly using the fuel handling machine. No fuel movements are needed. Spent fuel specific radiation from the fission product 137Cs at the gamma-ray energy of 662 keV is detected above the assemblies in the storage rack using a NaI(Tl) detector. In the design and in licensing the requirements of the IAEA, operator and the safety authority have been taken into account. The BWR SFAT allows modifications for other LWR fuel types with minor changes. The work has been carried out under the task FIN A 563 of the Finnish Support Programme to IAEA Safeguards. (orig.) (9 refs., 22 figs.)

  8. Role of BWR MK I secondary containments in severe accident mitigation

    International Nuclear Information System (INIS)

    The recent advent of detailed containment analysis codes such as CONTAIN and MELCOR has facilitated the development of the first large-scale, architectural-based BWR secondary containment models. During the past year ORNL has developed detailed, plant-specific models of the Browns Ferry and Peach Bottom secondary containments, and applied these models in a variety of studies designed to evaluate the role and effectiveness of BWR secondary containments in severe accident mitigation. The topology and basis for these models is discussed, together with some of the emerging insights from these studies

  9. Control Rod Pattern Planning of a BWR using Enhanced Nelder-Mead Method

    International Nuclear Information System (INIS)

    We propose a new optimization algorithm for the short-term planning of control rod patterns in an operating BWR. This algorithm is based on the enhanced Nelder-Mead simplex method in which convergence ability is improved for constrained problems in several ways. The main characteristic of this approach is it uses continuous values for the axial positions of control rods. Through calculations in an actual BWR plant, we showed that the new algorithm is effective for automation of short-term planning and reduction of the engineer's workload. (authors)

  10. Validation and application of TRACE for transient analysis of a German BWR plant

    International Nuclear Information System (INIS)

    The Karlsruhe Institute of Technology (KIT) is participating on CAMP to validate TRACE for LWR transient analysis. The validation of safety-relevant heat transfer models of TRACE, data BFBT bundle tests are available for institutions taking part on international OECD/NEA Benchmarks. In the first step, the BWR-relevant models of TRACE were validated using experimental data. Then, an integral plant model of a German BWR of Type-72 was developed using the multidimensional VESSEL-component of TRACE to simulate a plant events such as the TUSA. In this paper, the validation work, the plant modeling and selected results will be presented and discussed. (author)

  11. AREVA solutions to licensing challenges in PWR and BWR reload and safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Curca-Tivig, Florin [AREVA GmbH, Erlangen (Germany)

    2016-05-15

    Regulatory requirements for reload and safety analyses are evolving: new safety criteria, request for enlarged qualification databases, statistical applications, uncertainty propagation.. In order to address these challenges and access more predictable licensing processes, AVERA is implementing consistent code and methodology suites for PWR and BWR core design and safety analysis, based on first principles modeling and extremely broad verification and validation data base. Thanks to the high computational power increase in the last decades methods' development and application now include new capabilities. An overview of the main AREVA codes and methods developments is given covering PWR and BWR applications in different licensing environments.

  12. Interpretation of cross-correlation transit time from LPRM signals in a BWR

    International Nuclear Information System (INIS)

    In a commercial BWR, the cross correlation between two noise signals from Local Power Range Monitors is performed. As a consequence, a transit time is determined. There is not yet a wide acceptation of what exactly this transit time means. Four different techniques for the noise analysis are used; the results were alike, but the Rehocence method stood out as the most robust. Noise measurements were performed in a commercial BWR. Understanding what the transit time means, an attempt is made. Comparison with the output of a detailed thermohydraulic code shows agreement with the transit time measurements. (authors)

  13. The Delft desire facility for studies on (natural circulation) BWR primary system statics and dynamics

    International Nuclear Information System (INIS)

    A test facility for research on BWR core statics and dynamics was designed and built in Delft. The loop, DESIRE, consists of a BWR fuel assembly, a riser, condenser and a downcorner section. Freon-12 is used as a coolant. Presently, research on this facility is focused on investigations of the physical aspects of natural-circulation cooling and reactor kinetic stability. To this end, an artificial feedback from in-core void fraction to heating power is being established. The void fraction is determined on a sub-channel level by measuring the transmission of a collimated gamma beam

  14. Generic safety evaluation report regarding integrity of BWR scram system piping

    International Nuclear Information System (INIS)

    Safety concerns associated with postulated pipe breaks in the boiling water reactor (BWR) scram system were identified during the staff's continuing investigation of the Browns Ferry Unit 3 control rod partial insertion failure on June 28, 1980. This report includes an evaluation of the licensing basis for the BWR scram discharge volume (SDV) piping and an assessment of the potential for the SDV piping to fail while in service. A discussion of the means available for mitigation an unlikely SDV system failure is provided. Generic recommendations are made to improve mitigation capability and ensure that system integrity is maintained in service

  15. Improving nuclear economics without increasing proliferation risk

    International Nuclear Information System (INIS)

    The postponement of recycle systems is an insurance against the premature plunge into widespread commercial use of plutonium, with its potential for increasing proliferation risks. We can afford to postpone the deployment process with little penalty by improving light water reactors in once-through mode (LWROT). This would make nuclear power cheaper without increasing proliferation risk. The deferral is desirable because very valuable information about the feasibility, competitiveness, and resistivity to proliferation of technologies such as advance converters, the once-through fast reactor, and the fusion-fission hybrid reactor will be learned in the next 10 years. There is a distinct possibility that, as far as nuclear energy is concerned, we can rely on LWR(OT)s with evolutionary improvements far into the future. 29 references, 2 tables

  16. Timing criteria for supplemental BWR emergency response equipment

    International Nuclear Information System (INIS)

    The Great Tohuku Earthquake and subsequent Tsunami represented a double failure event which destroyed offsite power connections to Fukushima-Daiichi site and then destroyed on-site electrical systems needed to run decay heat removal systems. The accident could have been mitigated had there been supplemental portable battery chargers, supplemental pumps, and in-place piping connections to provide alternate decay heat removal. In response to this event in the USA, two national response centers, one in Memphis, Tennessee, and another in Phoenix, Arizona, will begin operation. They will be able to dispatch supplemental emergency response equipment to any nuclear plant in the U.S. within 24 hours. In order to define requirements for supplemental nuclear power plant emergency response equipment maintained onsite vs. in a regional support center it is necessary to confirm: (a) the earliest time such equipment might be needed depending on the specific scenario, (b) the nominal time to move the equipment from a storage location either on-site or within the region of a nuclear power plant, and (c) the time required to connect in the supplemental equipment to use it. This paper describes an evaluation process for a BWR-4 with a Mark I Containment starting with: (a) severe accident simulation to define best estimate times available for recovery based on the specific scenario, (b) identify the key supplemental response equipment needed at specific times to accomplish recovery of key safety functions, and (c) evaluate what types of equipment should be warehoused on-site vs. in regional response centers. (authors)

  17. Description and status of BWR water radiolysis model

    International Nuclear Information System (INIS)

    A water radiolysis model for boiling water reactor primary coolant circuits has been developed jointly with Harwell Laboratories. Hydrogen water chemistry plant tests, on which the model is based, started in 1982. Since then, ten plant tests have been completed. Test results indicate that: (1) oxygen concentrations in the steam from plant to plant decreases fairly uniformly with increasing hydrogen concentration; (2) oxygen concentrations in the recirculating water varies widely with increasing hydrogen concentration in the feedwater; and (3) copper significantly suppresses the reduction of oxygen. The radiolysis model was developed to provide predictive conditions in the BWR primary circuit including the core. The objectives of the work are: (1) to develop and calibrate the model against experimental data; (2) to predict the concentration of various radiolytic species and oxygen suppression in various parts of the primary system under hydrogen water chemistry conditions; and (3) to define the role of copper on the reactions. The model considers the thermodynamics and kinetics of homogeneous chemical reactions in eight regions of the reactor and its external circuit. From initial concentrations, the model constructs a set of simultaneous equations to calculate hydrogen, hydrogen peroxide, and oxygen concentrations in various regions. Pilgrim data were used for the initial calibration. The model includes neutron and gamma dose rates in the core region and gamma dose rates in the other various regions, as well as temperature, power levels, heat transfer, steam quality, and gas/liquid transport in the core region, etc. The results of calculations for the oxygen concentrations in the recirculation water as a function of hydrogen concentration in the feedwater show good agreement with the experimental data obtained in a number of reactor tests

  18. Liquid films and droplet deposition in a BWR fuel element

    International Nuclear Information System (INIS)

    In the upper part of boiling water reactors (BWR) the flow regime is dominated by a steam-water droplet flow with liquid films on the nuclear fuel rod, the so called (wispy) annular flow regime. The film thickness and liquid flow rate distribution around the fuel rod play an important role especially in regard to so called dryout, which is the main phenomenon limiting the thermal power of a fuel assembly. The deposition of droplets in the liquid film is important, because this process sustains the liquid film and delays dryout. Functional spacers with different vane shapes have been used in recent decades to enhance droplet deposition and thus create more favorable conditions for heat removal. In this thesis the behavior of liquid films and droplet deposition in the annular flow regime in BWR bundles is addressed by experiments in an adiabatic flow at nearly ambient pressure. The experimental setup consists of a vertical channel with the cross-section resembling a pair of neighboring subchannels of a fuel rod bundle. Within this double subchannel an annular flow is established with a gas-water mixture. The impact of functional spacers on the annular flow behavior is studied closely. Parameter variations comprise gas and liquid flow rates, gas density and spacer shape. The setup is instrumented with a newly developed liquid film sensor that measures the electrical conductance between electrodes flush to the wall with high temporal and spatial resolution. Advanced post-processing methods are used to investigate the dynamic behavior of liquid films and droplet deposition. The topic is also assessed numerically by means of single-phase Reynolds-Averaged-Navier-Stokes CFD simulations of the flow in the gas core. For this the commercial code STAR-CCM+ is used coupled with additional models for the liquid film distribution and droplet motion. The results of the experiments show that the liquid film is quite evenly distributed around the circumference of the fuel rods. The

  19. 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)

  20. Strengthening the foundations of proliferation assessment tools.

    Energy Technology Data Exchange (ETDEWEB)

    Rexroth, Paul E.; Saltiel, David H.; Rochau, Gary Eugene; Cleary, Virginia D.; Ng, Selena (AREVA NC, Paris, France); Greneche, Dominique (AREVA NC, Paris, France); Giannangeli, Don (Texas A& M University, College Station, TX); Charlton, William S. (Texas A& M University, College Station, TX); Ford, David (Texas A& M University, College Station, TX)

    2007-09-01

    Robust and reliable quantitative proliferation assessment tools have the potential to contribute significantly to a strengthened nonproliferation regime and to the future deployment of nuclear fuel cycle technologies. Efforts to quantify proliferation resistance have thus far met with limited success due to the inherent subjectivity of the problem and interdependencies between attributes that lead to proliferation resistance. We suggest that these limitations flow substantially from weaknesses in the foundations of existing methodologies--the initial data inputs. In most existing methodologies, little consideration has been given to the utilization of varying types of inputs--particularly the mixing of subjective and objective data--or to identifying, understanding, and untangling relationships and dependencies between inputs. To address these concerns, a model set of inputs is suggested that could potentially be employed in multiple approaches. We present an input classification scheme and the initial results of testing for relationships between these inputs. We will discuss how classifying and testing the relationship between these inputs can help strengthen tools to assess the proliferation risk of nuclear fuel cycle processes, systems, and facilities.

  1. Limiting Future Proliferation and Security Risks

    International Nuclear Information System (INIS)

    A major new technical tool for evaluation of proliferation and security risks has emerged over the past decade as part the activities of the Generation IV International Forum. The tool has been developed by a consensus group from participating countries and organizations and is termed the Proliferation Resistance and Physical Protection (PR and PP) Evaluation Methodology. The methodology defines a set of challenges, analyzes system response to these challenges, and assesses outcomes. The challenges are the threats posed by potential actors (proliferant states or sub-national adversaries). It is of paramount importance in an evaluation to establish the objectives, capabilities, resources, and strategies of the adversary as well as the design and protection contexts. Technical and institutional characteristics are both used to evaluate the response of the system and to determine its resistance against proliferation threats and robustness against sabotage and terrorism threats. The outcomes of the system response are expressed in terms of a set of measures, which thereby define the PR and PP characteristics of the system. This paper summarizes results of applications of the methodology to nuclear energy systems including reprocessing facilities and large and small modular reactors. The use of the methodology in the design phase a facility will be discussed as it applies to future safeguards concepts.

  2. 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)

  3. 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)

  4. The nuclear proliferation; La proliferation nucleaire

    Energy Technology Data Exchange (ETDEWEB)

    Gere, F. [Ecole Polytechnique, 91 - Palaiseau (France)

    1995-04-01

    In this book is detailed the beginning of nuclear military power, with the first bomb of Hiroshima, the different ways of getting uranium 235 and plutonium 239, and how the first countries (Usa, Ussr, China, United kingdom, France) got nuclear weapons. Then the most important part is reviewed with the details of non-proliferation treaty and the creation of IAEA to promote civilian nuclear power in the world and to control the use of plutonium and uranium in nuclear power plants. The cases of countries who reached the atom mastery, such Israel, South Africa, Pakistan, Iraq, North Korea, Argentina, Brazil, Iran, Algeria, Taiwan and the reasons which they wanted nuclear weapon for or why they gave up, are exposed.

  5. Plutonium and Minor Actinides Recycling in Standard BWR using Equilibrium Burnup Model

    Directory of Open Access Journals (Sweden)

    Abdul Waris

    2008-03-01

    Full Text Available Plutonium (Pu and minor actinides (MA recycling in standard BWR with equilibrium burnup model has been studied. We considered the equilibrium burnup model as a simple time independent burnup method, which can manage all possible produced nuclides in any nuclear system. The equilibrium burnup code was bundled with a SRAC cell-calculation code to become a coupled cell-burnup calculation code system. The results show that the uranium enrichment for the criticality of the reactor, the amount of loaded fuel and the required natural uranium supply per year decrease for the Pu recycling and even much lower for the Pu & MA recycling case compared to those of the standard once-through BWR case. The neutron spectra become harder with the increasing number of recycled heavy nuclides in the reactor core. The total fissile rises from 4.77% of the total nuclides number density in the reactor core for the standard once-through BWR case to 6.64% and 6.72% for the Plutonium recycling case and the Pu & MA recycling case, respectively. The two later data may become the main basis why the required uranium enrichment declines and consequently diminishes the annual loaded fuel and the required natural uranium supply. All these facts demonstrate the advantage of plutonium and minor actinides recycling in BWR.

  6. 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

  7. Analysis of relationship between stability and flow parameters in a BWR

    International Nuclear Information System (INIS)

    Results of quantitative analysis of mutual relationship between the BWR stability and channel steam velocity are presented. The stability parameter, defined by the damping ratio, and the steam velocity are estimated by analysis of neutron noise data from local power range monitor (LPRM) detector signals. These parameters are treated as varying randomly as a function of time

  8. Coupled field effects in BWR stability simulations using SIMULATE-3K

    International Nuclear Information System (INIS)

    The SIMULATE-3K code is the transient analysis version of the Studsvik advanced nodal reactor analysis code, SIMULATE-3. Recent developments have focused on further broadening the range of transient applications by refinement of core thermal-hydraulic models and on comparison with boiling water reactor (BWR) stability measurements performed at Ringhals unit 1, during the startups of cycles 14 through 17

  9. Predictions by the proper orthogonal decomposition reduced order methodology regarding non-linear BWR stability

    International Nuclear Information System (INIS)

    Unexpected non-linear boiling water reactor (BWR) instability events in various plants, e.g. LaSalle II in 1988 and Oskarshamn II in 1990 amongst others, emphasize the major safety relevance and the existence of parameter regions with unstable behavior. A detailed description of the complete dynamical non-linear behavior is of paramount importance for BWR operation. An extension of state-of-the-art methodology towards a more general stability description, also applicable in the non-linear region, could lead to a deeper understanding of non-linear BWR stability phenomena. With the intention of a full non-linear stability analysis of the two-phase BWR system, the present paper aims at a general non-linear methodology capable to achieve reliable and numerical stable reduced order models (ROMs), representing the dynamical behavior of an original system based on a small number of transients. Model-specific options and aspects of the proposed methodology are focused on and illustrated by means of a strongly non-linear dynamical system showing complex oscillating behavior. Prediction capability of the proposed methodology is also addressed. (orig.)

  10. Reducing BWR O and M costs through on-line performance monitoring

    International Nuclear Information System (INIS)

    Competition in the electric power industry has placed significant emphasis on reducing operating and maintenance (O and M) costs at nuclear facilities. Therefore, on-line performance monitoring to locate power losses for boiling water reactor (BWR) plants is creating tremendous interest. In addition, the ability to automate activities such as data collection, analysis, and reporting increases the efficiency of plant engineers and gives them more time to concentrate on solving plant efficiency problems. This capability is now available with a unique software product called GEBOPS. GE Nuclear Energy, in conjunction with Joint Venture partner Black and Veatch, has undertaken development of the General Electric/Black and Veatch On-line Performance System (GEBOPS), an on-line performance monitoring system for BWR plants. The experience and expertise of GE Nuclear Energy with BWR plants, coupled with the proven on-line monitoring software development experience and capability of Black and Veatch, provide the foundation for a unique product which addresses the needs of today's BWR plants

  11. Boron concentration evolution in the temporary curtains of a BWR reactor. Burcur code

    International Nuclear Information System (INIS)

    The theoretical model and the user's guide of the code Burcur is included. This code analyzes the burnable poison concentration of the temporary curtains as a function of time, for BWR reactors of the 7 x 7 design. The computing time being reasonably short, the number of burnup steps is as high as necessary.(author)

  12. Bivariate empirical mode decomposition applied to the estimation of out-of-phase oscillations in BWR

    International Nuclear Information System (INIS)

    Highlights: • Bivariate empirical mode decomposition (BEMD) in BWR’s instabilities is studied. • The phase determines out-of-phase oscillations in the BWR instability. • The method based in BEMD does not represent a high computational complexity. • The methodology was validated with Nuclear Power Plants stability benchmarks. • The results show that the method contributes to detect out-of-phase oscillations. - Abstract: In this paper a new method based on the bivariate empirical mode decomposition to estimate the phase of regional (out-of-phase) or global (in-phase) modes associated with instabilities in boiling water reactors (BWR), is explored. The proposed method allows decomposing the analyzed signal (constructed from two different Local Power Range Monitors, LPRMs) in different levels or intrinsic mode functions (IMF). The estimation of the phase between these LPRM signals can be achieved by tracking the modes associated to the instability of the BWR and obtaining the cross-correlation function of their corresponding IMF. This phase determines possible out-of-phase oscillations, which play an important role in the BWR instability. The method is relatively simple to implement and it does not represent a high computational complexity. The methodology was tested with simulated signals and validated with two events reported in the Forsmark and Ringhals stability benchmarks. The results of the cases studied show that the proposed method clearly contributes on the fact to detect possible cases of out-of-phase oscillations

  13. Technical report on material selection and processing guidelines for BWR coolant pressure boundary piping

    International Nuclear Information System (INIS)

    This report sets forth the NRC staff's revised acceptable methods to reduce the intergranular stress corrosion cracking susceptibility of BWR ASME Code Class 1 and 2 pressure boundary piping and safe end. For plants that cannot fully comply with the material selection, testing, and processing guidelines of this document, varying degrees of augmented inservice inspection and leak detection requirements are presented

  14. Identification of dose-reduction techniques for BWR and PWR repetitive high-dose jobs

    International Nuclear Information System (INIS)

    As a result of concern about the apparent increase in collective radiation dose to workers at nuclear power plants, this project will provide information to industry in preplanning for radiation protection during maintenance operations. This study identifies Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) repetitive jobs, and respective collective dose trends and dose reduction techniques. 3 references, 2 tables

  15. Water chemistry management of nuclear power plant. Water chemistry management of BWR plant

    International Nuclear Information System (INIS)

    There are two kinds of nuclear power plants such as Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) in Japan. In this paper, a water chemistry management of BWR plant is explained. BWR plant makes steam produced in the reactor send to the turbine and produce power, then condensate in the main condenser and use again as feed water. The objects of water chemistry management of BWR are security of good conditions of fuel and structure materials and reduction of the dose equivalent and the radioactive waste. The volume of coolant depends on the temperature change, the concentration of boric acid for neutron absorber, lithium hydroxide for pH control and hydrogen gas for corrosion are controlled. Impurity metals in water of reactor are removed by the condensate demineralizer. The concentration of boron and lithium is controlled from 0 to 4000 ppm and from 0.2 to 2.2 ppm, respectively. On water chemistry technologies for dose reduction, oxygen injection into feed water and control operation of rate of Ni/Fe are explained. On the technologies for preventive maintenance, degassing operation of reactor and hydrogen injection into feed water are described. (S.Y.)

  16. Implications of crevice chemistry for cracking of BWR recirculation inlet safe-ends

    International Nuclear Information System (INIS)

    A mathematical model has been utilized to predict the electrochemistry within crevices exposed to BWR environments (0.3 μS/cm and 300oC). For both Normal Water Chemistry (NWC) and Hydrogen Water Chemistry (HWC), the modeled chemistry within the crevice was assessed to see if initiation of stress corrosion cracking was probable. (author)

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

    International Nuclear Information System (INIS)

    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

  18. 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

  19. Design and axial optimization of nuclear fuel for BWR reactors

    International Nuclear Information System (INIS)

    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

  20. Radial optimization of a BWR fuel cell using genetic algorithms

    International Nuclear Information System (INIS)

    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 U235 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 the placement of

  1. The development of emergency core cooling systems in the PWR, BWR, and HWR Candu type of nuclear power plants

    International Nuclear Information System (INIS)

    Emergency core cooling systems in the PWR, BWR, and HWR-Candu type of nuclear power plant are reviewed. In PWR and BWR the emergency cooling can be catagorized as active high pressure, active low pressure, and a passive one. The PWR uses components of the shutdown cooling system: whereas the BWR uses components of pressure suppression contaiment. HWR Candu also uses the shutdown cooling system similar to the PWR except some details coming out from moderator coolant separation and expensive cost of heavy water. (author)

  2. 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)

  3. Cell Proliferation in Neuroblastoma

    Directory of Open Access Journals (Sweden)

    Laura L. Stafman

    2016-01-01

    Full Text Available Neuroblastoma, the most common extracranial solid tumor of childhood, continues to carry a dismal prognosis for children diagnosed with advanced stage or relapsed disease. This review focuses upon factors responsible for cell proliferation in neuroblastoma including transcription factors, kinases, and regulators of the cell cycle. Novel therapeutic strategies directed toward these targets in neuroblastoma are discussed.

  4. Battling Nuclear Proliferation

    Institute of Scientific and Technical Information of China (English)

    2007-01-01

    As the North Korean and Iranian nuclear issues develop and efforts to resolve them continue, global attention to anti-nuclear proliferation and the work of the International Atomic Energy Agency (IAEA) has become even more intense. Pang Sen, Chairman of

  5. Behavior of small-sized BWR fuel under reactivity initiated accident conditions

    International Nuclear Information System (INIS)

    The present work was performed on this small-sized BWR fuel, where Zr liner and rod prepressurization were taken as experimental parameters. Experiment was done under simulated reactivity initiated accident (RIA) conditions at Nuclear Safety Research Reactor (NSRR) belonged to Japan Atomic Energy Research Institute (JAERI). Major remarks obtained are as follows: (1) Three different types of the fuel rods consisted of (a) Zr lined/pressurized (0.65MPa), (b) Zr lined/non-pressurized and (c) non-Zr lined/pressurized (o.65MPa) were used, respectively. Failure thresholds of these were not less than that (260 cal/g·fuel) described in Japanese RIA Licensing Guideline. Small-sized BWR and conventional 8 x 8 BWR fuels were considered to be in almost the same level in failure threshold. Failure modes of the three were (a) cladding melt/brittle, (b) cladding melt/brittle and (c) rupture by large ballooning, respectively. (2) The magnitude of pressure pulse at fuel fragmentation was also studied by lined/pressurized and non-lined/pressurized fuels. Above the energy deposition of 370 cal/g·fuel, mechanical energy (or pressure) was found to be released from these fragmented fuels. No measurable difference was, however, observed between the tested fuels and NSRR standard (and conventional 8 x 8 BWR) fuels. (3) It is worthy of mentioning that Zr liner tended to prevent the cladding from large ballooning. Non-lined/pressurized fuel tended to cause wrinkle deformation at cladding. Hence, cladding external was notched much by the wrinkles. (4) Time to fuel failure measured from the tested BWR fuels (pressurization < 0.6MPA) was longer than that measured from PWR fuels (pressurization < 3.2MPa). The magnitude of the former was of the order of 3 ∼ 6s, while that of the latter was < 1s. (J.P.N.)

  6. Study of the Utilization BWR Type Nuclear Power Reactor for Desalination Process

    International Nuclear Information System (INIS)

    The needs of fresh water increased by rapid population growth and industrials expansion, but these demands can not be prepared naturally. Following this case, seawater desalination becomes the primer option which can fulfill the need through the nuclear desalination technology. The coupled nuclear power reactor enables to supply thermal energy for auxiliary equipment and pumps operation. The utilization study of power reactor type BWR coupled with desalination process has been performed. The goal of study is to obtain characteristic data of desalted water specification which desalination system coupling with nuclear power plant produced energy for desalination process. The study is carried out by browsing data and information, and comprehensive review of thermal energy correlation between NPP with desalination process installation. According to reviewing are found that the thermal energy and electric power utilization from the nuclear power reactor are enable to remove the seawater to produce desalted water and also to operate auxiliary equipments. The assessment results is VK-300 reactor prototype, BWR type 250 MW(e) power are cogeneration unit can supplied hot steam temperature 285 °C to the extraction turbine to empower 150 MW electric power, and a part of hot steam 130 °C is use to operate desalination process and remind heat is distribute to the municipal and offices at that region. The coupled of VK-300 reactor power type BWR with desalination installation of MED type enable to produce desalted water with high quality distillate. Based on the economic calculation that the VK-300 reactor power of BWR type produced water distillate capacity is 300.000 m3/hour with cost US$ 0.58/m3. The coupling VK-300 reactor power type BWR with MED desalination plant is competitive economically. (author)

  7. 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

  8. 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

  9. Proliferation after the Iraq war

    International Nuclear Information System (INIS)

    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.)

  10. 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

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

    International Nuclear Information System (INIS)

    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

  12. Preliminary design study of small long life boiling water reactor (BWR) with tight lattice thorium nitride fuel

    Science.gov (United States)

    Trianti, Nuri; Su'ud, Zaki; Arif, Idam; Riyana, EkaSapta

    2014-09-01

    Neutronic performance of small long-life boiling water reactors (BWR) with thorium nitride based fuel has been performed. A recent study conducted on BWR in tight lattice environments (with a lower moderator percentage) produces small power reactor which has some specifications, i.e. 10 years operation time, power density of 19.1 watt/cc and maximum excess reactivity of about 4%. This excess reactivity value is smaller than standard reactivity of conventional BWR. The use of hexagonal geometry on the fuel cell of BWR provides a substantial effect on the criticality of the reactor to obtain a longer operating time. Supported by a tight concept lattice where the volume fraction of the fuel is greater than the moderator and fuel, Thorium Nitride give good results for fuel cell design on small long life BWR. The excess reactivity of the reactor can be reduced with the addition of gadolinium as burnable poisons. Therefore the hexagonal tight lattice fuel cell design of small long life BWR that has a criticality more than 20 years of operating time has been obtained.

  13. Preliminary design study of small long life boiling water reactor (BWR) with tight lattice thorium nitride fuel

    International Nuclear Information System (INIS)

    Neutronic performance of small long-life boiling water reactors (BWR) with thorium nitride based fuel has been performed. A recent study conducted on BWR in tight lattice environments (with a lower moderator percentage) produces small power reactor which has some specifications, i.e. 10 years operation time, power density of 19.1 watt/cc and maximum excess reactivity of about 4%. This excess reactivity value is smaller than standard reactivity of conventional BWR. The use of hexagonal geometry on the fuel cell of BWR provides a substantial effect on the criticality of the reactor to obtain a longer operating time. Supported by a tight concept lattice where the volume fraction of the fuel is greater than the moderator and fuel, Thorium Nitride give good results for fuel cell design on small long life BWR. The excess reactivity of the reactor can be reduced with the addition of gadolinium as burnable poisons. Therefore the hexagonal tight lattice fuel cell design of small long life BWR that has a criticality more than 20 years of operating time has been obtained

  14. 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.

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

    OpenAIRE

    2008-01-01

    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 presen...

  16. Can we predict nuclear proliferation

    International Nuclear Information System (INIS)

    The author aims at improving nuclear proliferation prediction capacities, i.e. the capacities to identify countries susceptible to acquire nuclear weapons, to interpret sensitive activities, and to assess nuclear program modalities. He first proposes a retrospective assessment of counter-proliferation actions since 1945. Then, based on academic studies, he analyzes what causes and motivates proliferation, with notably the possibility of existence of a chain phenomenon (mechanisms driving from one program to another). He makes recommendations for a global approach to proliferation prediction, and proposes proliferation indices and indicators

  17. Application of eddy current inspection to the Inconel weld of BWR internals

    International Nuclear Information System (INIS)

    In order to definite the basic specifications of application of ECT (Eddy Current Test) to Inconel weld of BWR internals, the inspection and numerical analysis were carried out. The characteristics of the existing ECT probe were studied by making sample as same as CRD stud tube, measuring the relative permeability and electric conductivity of Inconel and alloy and evaluating ECT probe. On the basis of the results obtained, the basic specifications were determined and a new eddy current probe for inspection was designed and produced. The new ECT probe was able to detect small notch in Inconel weld, to classify the defects by eddy current inspection signal and sizing the length and depth. It is concluded that the new ECT probe is able to apply the Inconel weld of BWR internals. (S.Y.)

  18. Investigation of distorted geometry simulation of pool dynamics in horizontal-vent BWR containments

    International Nuclear Information System (INIS)

    Experiments were conducted to investigate the accuracy of distorted geometry testing of pool dynamics in horizontal-vent BWR containments. Distorted-geometry testing implies testing in systems where the flow-wise dimensions are full scale, but all dimensions transverse to the flow are reduced in the same proportion. The assumption is that flow velocities, pressures and other thermodynamic properties will be the same in the distorted-geometry system as in its correctly proportioned counterpart. The experiments, which were done at small scale using the established scaling laws, showed that the geometric distortions can have a significant effect on the pool swell under conditions which are roughly representative of horizontal-vent BWR containment systems during a LOCA. Breakthrough occurred later, the water ligament was thicker, and pool velocity lower in a system where the cross-sectional areas were reduced by a factor of three. Some reasons for the differences are discussed

  19. Aggressive chemical decontamination tests on small valves from the Garigliano BWR

    International Nuclear Information System (INIS)

    In order to check the effectiveness of direct chemical decontamination on small and complex components, usually considered for storage without decontamination because of the small amount, some tests were performed on the DECO experimental loop. Four small stainless steel valves from the primary system of the Garigliano BWR were decontaminated using mainly aggressive chemicals such as HC1, HF, HNO3 and their mixtures. On two valves, before the treatment with aggressive chemicals, a step with soft chemical (oxalic and citric acid mixture) was performed in order to see whether a softening action enhances the following aggressive decontamination. Moreover, in order to increase as much as possible the decontamination effectiveness, a decontamination process using ultrasounds jointly with aggressive chemicals was investigated. After an intensive laboratory testing programme, two smaller stainless steel valves from the primary system of the Garigliano BWR were decontaminated using ultrasounds in aggressive chemical solutions

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

    International Nuclear Information System (INIS)

    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)