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Sample records for bwr proliferation resistance

  1. Enhancing BWR proliferation resistance fuel with minor actinides

    Science.gov (United States)

    Chang, Gray S.

    2009-03-01

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

  2. Proliferation resistance modeling

    International Nuclear Information System (INIS)

    Bari, R.; Peterson, P.; Roglans, J.; Mladineo, S.; Nuclear Engineering Division; BNL; Univ. of California at Berkely; PNNL

    2004-01-01

    The National Nuclear Security Administration is developing methods for nonproliferation assessments. A working group on Nonproliferation Assessment Methodology (NPAM) assembled a toolbox of methods for various applications in the nonproliferation arena. One application of this methodology is to the evaluation of the proliferation resistance of Generation IV nuclear energy systems. This paper first summarizes the key results of the NPAM program and then provides results obtained thus far in the ongoing application, which is co-sponsored by the DOE Office of Nuclear Energy Science and Technology. In NPAM, a top-level measure of proliferation resistance for a fuel cycle system is developed from a hierarchy of metrics. The problem is decomposed into: metrics to be computed, barriers to proliferation, and a finite set of threats. The analyst models the process undertaken by the proliferant to overcome barriers to proliferation and evaluates the outcomes. In addition to proliferation resistance (PR) evaluation, the application also addresses physical protection (PP) evaluation against sabotage and theft. The Generation IV goal for future nuclear energy systems is to assure that they are very unattractive and the least desirable route for diversion or theft of weapons-usable materials, and provide increased physical protection against terrorism. An Expert Group, addressing this application, has identified six high-level measures for the PR goals (six measures have also been identified for the PP goals). Combined together, the complete set of measures provides information for program policy makers and system designers to compare specific system design features and integral system characteristics and to make choices among alternative options. The Group has developed a framework for a phased evaluation approach to analyzing PR and PP of system characteristics and to quantifying metrics and measures. This approach allows evaluations to become more detailed and representative

  3. Corrosion resistance improvement of ferritic steels through hydrogen additions to the BWR coolant

    International Nuclear Information System (INIS)

    Gordon, B.M.; Jewett, C.W.; Pickett, A.E.; Indig, M.E.

    1984-01-01

    Motivated by the success of oxygen suppression for mitigation of intergranular stress corrosion cracking (IGSCC) in weld sensitized austenitic materials used in Boiling Water Reactors (BWRs), oxygen suppression, through hydrogen additions to the feedwater was investigated to determine its affect on the corrosion resistance of ferritic and martensitic BWR structural materials. The results of these investigations are presented in this paper, where particular emphasis is placed on the corrosion performance of BWR pressure vessel low alloy steels, carbon steel piping materials and martensitic pump materials. It is important to note that the corrosion resistance of these materials in the BWR environment is excellent. Consequently this investigation was also motivated to determine whether there were any detrimental effects of hydrogen additions, as well as to identify any additional margin in ferritic/martensitic materials corrosion performance

  4. 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)

  5. Framework of Comprehensive Proliferation Resistance Evaluation Methodology

    International Nuclear Information System (INIS)

    Kim, Min Su; Jo, Seong Youn; Kim, Min Soo; Kim, Jae San; Lee, Hyun Kyung

    2007-01-01

    Civilian nuclear programs can be used as a pretext to acquire technologies, materials, equipment for military weapon programs. Consequently, international society has a strong incentive to develop a nuclear system more proliferation resistant to assure that the civilian nuclear energy system is an unattractive and least desirable route for diversion of weapon usable material. The First step developing a more proliferation resistant nuclear energy system is to develop a systematic and standardized evaluation methodology to ensure that any future nuclear energy system satisfies the proliferation resistance goals. Many attempts to develop systematic evaluation methodology have been proposed and many systems for assessing proliferation resistance have been previously studied. However, a comprehensive proliferation resistance evaluation can not be achieved by simply applying one method since complicated proliferation resistance characteristics, including inherent features and extrinsic features, should be completely evaluated. Therefore, it is necessary to develop one incorporated evaluation methodology to make up for weak points of each evaluation method. The objective of this study is to provide a framework of comprehensive proliferation resistance evaluation methodology by incorporating two generally used evaluation methods, attribute and scenario analysis

  6. Comparative analysis of proliferation resistance assessment methodologies

    International Nuclear Information System (INIS)

    Takaki, Naoyuki; Kikuchi, Masahiro; Inoue, Naoko; Osabe, Takeshi

    2005-01-01

    Comparative analysis of the methodologies was performed based on the discussions in the international workshop on 'Assessment Methodology of Proliferation Resistance for Future Nuclear Energy Systems' held in Tokyo, on March 2005. Through the workshop and succeeding considerations, it is clarified that the proliferation resistance assessment methodologies are affected by the broader nuclear options being pursued and also by the political situations of the state. Even the definition of proliferation resistance, despite the commonality of fundamental issues, derives from perceived threat and implementation circumstances inherent to the larger programs. Deep recognitions of the 'difference' among communities would help us to make further essential and progressed discussion with harmonization. (author)

  7. Alternative Zr alloys with irradiation resistant precipitates for high burnup BWR application

    International Nuclear Information System (INIS)

    Garzarolli, F.; Ruhmann, H.; Van Swan, L.

    2002-01-01

    In the core of BWRs, the second-phase particles (SPP) of Zircaloy-2 and Zircaloy-4, the Zr(FeCr) 2 and the Zr 2 (FeNi) phase, release Fe and dissolve. The degree of dissolution depends on initial size and fluence. These SPP, however, are important for the corrosion behavior of Zircaloy. Zircaloy shows an increase of corrosion at a certain burnup, depending on the initial SPP size and fast neutron fluence. Only Zr alloys with irradiation resistant SPP avoid this type of increased corrosion completely. Two types of irradiation resistant materials were considered. One is a Zr-Sn-Fe alloy containing the Zr 3 Fe phase, which is irradiation resistant under BWR conditions. The other material is a Zr-Sn-Nb alloy containing the irradiation resistant β-Nb phase. In-BWR tests have shown that a Sn content of >0.8% is mandatory to minimize the nodular corrosion. Two prototypes of irradiation resistant alloys, Zr1.3Sn0.25-0.3 Fe and Zr1Sn2-3Nb, were irradiated in a BWR for 1372 days to a fast fluence of 9 x 10 21 n/cm 2 (E > 1 MeV). These irradiation tests showed that Zr1.3Sn0.25-0.3 Fe has a little lower resistance against nodular corrosion than optimized LTP (Low Temperature Process) Zircaloy-2/4 and revealed that Zr1Sn2-3Nb is superior to LTP Zircaloy-2/4 with respect to nodular and shadow corrosion resistance. The BWR corrosion resistance of Zr1Sn2-3Nb depends on heat treatment. The lowest corrosion was observed with material fabricated completely in the α-range, but also material manufactured in the lower (α+β)-range exhibits low corrosion. Material fabricated in the upper (α+β)-range showed a somewhat higher corrosion, a corrosion behavior similar to LTP Zircaloy-2/4. As far as final annealing is concerned, a long time annealing at 540 deg C is superior to a standard recrystallization treatment (e.g., at 580 deg C), which still leads to a corrosion behavior that is better than stress relieved Zr1Sn2-3Nb. Zr1Sn2-3Nb is resistant to shadow corrosion, when fabricated

  8. Proliferation resistance of small modular reactors fuels

    Energy Technology Data Exchange (ETDEWEB)

    Polidoro, F.; Parozzi, F. [RSE - Ricerca sul Sistema Energetico,Via Rubattino 54, 20134, Milano (Italy); Fassnacht, F.; Kuett, M.; Englert, M. [IANUS, Darmstadt University of Technology, Alexanderstr. 35, D-64283 Darmstadt (Germany)

    2013-07-01

    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 UO{sub 2} 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 UO{sub 2} 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.

  9. Proliferation resistance assessment of thermal recycle systems

    International Nuclear Information System (INIS)

    1979-02-01

    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

  10. Proliferation resistance assessment of pyro processing

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, E. H.; Ko, W. I.; Kim, H. D. [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    In 2002, world experts gathered and defined the term proliferation resistance as 'the characteristic of a nuclear energy system that impedes the diversion or undeclared production of nuclear material, or misuse of technology, by State in order to acquire nuclear weapons or other nuclear explosive devices.' The same report also defines the following terms: Intrinsic barriers (technical features) of proliferation resistance are features that result from the technical design of nuclear energy systems, including those that facilitate the implementation of extrinsic measures. Extrinsic barriers (institutional measures) of proliferation resistance are features that result from the decisions and undertakings of states related to nuclear energy system. Intrinsic barriers are further divided into material barriers.the 'intrinsic, or inherent, qualities of materials that reduce the inherent desirability or attractiveness of the material as an explosive' and technical barriers. The 'intrinsic technical lements of the fuel cycle, its facilities, processes, and equipment that serve to make it difficult to gain access to materials and/or to use or misuse facilities to obtain weapons usable materials.' Material barriers include isotopic, chemical, radiological, mass and bulk, and detectability, whereas technical barriers include facility unattractiveness, accessibility, available fissile mass, detectability of and time required for diversion, and skills, expertise, and knowledge. Assessing the proliferation resistance of pyro processing is meaningful only when compared with other processes. This paper attempts to discuss the features of pyro processing by comparing it with direct disposal and aqueous separation processes from a proliferation resistance viewpoint.

  11. Proliferation resistance criteria for fissile material disposition

    International Nuclear Information System (INIS)

    Close, D.A.; Fearey, B.L.; Markin, J.T.; Rutherford, D.A.; Duggan, R.A.; Jaeger, C.D.; Mangan, D.L.; Moya, R.W.; Moore, L.R.; Strait, R.S.

    1995-04-01

    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

  12. Proliferation resistance assessment of nuclear systems

    International Nuclear Information System (INIS)

    1978-09-01

    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

  13. 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.

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

    International Nuclear Information System (INIS)

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

    1979-01-01

    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

  15. Proliferation resistance assessment of nuclear systems

    International Nuclear Information System (INIS)

    1978-09-01

    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

  16. Proliferation resistance of the lithium reduction process

    International Nuclear Information System (INIS)

    Ko, W. I.; Ha, J. H.; Lee, S. Y.; Song, D. Y.; Kim, H. D.; Park, S. W.

    2002-01-01

    This paper addresses the characteristics of proliferation resistance of the lithium reduction process and international domestic safeguarding methods. In addition to dealing with qualitative features of the proliferation resistance, this study is emphasizing on the quantitative analysis of radiation barrier, which could be a significant accessibility barrier if the field is high enough to force a theft to shield the object during a theft. From the radiation barrier analysis, it is indicated that whole-body radiation dose is about 20 rem/hr at one meter of smelt and ingot metal of 40 kgHM, which could be considered to be a significant reduction in risk of theft. For safeguarding of this process, we propose a NDA concept for nuclear material accounting which is to measure the amount of curium in the reduction metal and associated process samples using a neutron coincidence counter and then to convert the curium mass into special nuclear material with predetermined curium ratios. For this, a well-type neutron coincidence counter with substantial shielding to protect the system from high gamma radiation is conceptually designed

  17. Criteria for proliferation resistance of nuclear fuel cycle options

    International Nuclear Information System (INIS)

    Kiriyama, Eriko; Pickett, Susan; Suzuki, Tatsujiro

    2000-01-01

    In order to understand the concept of nuclear proliferation resistance, this paper examines the technical definitions of proliferation resistance. Although nuclear proliferation resistance is often included as one of the major goals of advanced reactor research and development, the criteria for nuclear proliferation resistance of nuclear fuel cycles is not defined clearly. The implied meaning of proliferation resistance was compared in proposals regarding the nuclear fuel cycle. Discrepancies amongst the proposals regarding the technical definition of proliferation resistance is found. While all these proposals indicate proliferation resistance, few clearly spell out exactly what criteria they are measuring themselves against. However we found there are also common feature in many proposals. They are; (1) Reduction of Pu, (2) Less separated Weapon Usable Materials, (3) Fewer steps, (4) Barrier for Weapon Usable Materials. Recognizing that there are numerous political and infrastructure measures that may also be taken to guard against proliferation risks, we have focused here on the definition of proliferation resistance in terms of technical characteristics. Another important conclusion is that in many proposals proliferation resistance is only one of the important criteria such as energy security, economical efficiency, and safety. (author)

  18. Model for evaluating nuclear strategies with proliferation resistance

    International Nuclear Information System (INIS)

    Shay, M.R.; Hardie, R.W.; Omberg, R.P.

    1979-03-01

    A model was developed at HEDL to specifically analyze proliferation resistant strategies. The model was not designed to predict the future, but rather to provide a method for estimating the consequences of decisions affecting proliferation resistance in a rational and plausible manner. The characteristics of the model are described

  19. Evaluation of proliferation resistance using the INPRO methodology

    International Nuclear Information System (INIS)

    Yang, Myung Seung; Park, Joo Hwan; Ko, Won Il; Song, Kee Chan; Choi, Kun Mo; Kim, Jin Kyoung

    2007-01-01

    The IAEA launched the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) and developed the INPRO Methodology to provide guidelines and to assess the characteristics of a future innovative nuclear energy system in areas such as safety, economics, waste management, and proliferation resistance. The proliferation resistance area of the INPRO Methodology is reviewed here, and modifications for further improvements are proposed. The evaluation metrics including the evaluation parameters, evaluation scales and acceptance limits are developed for a practical application of the methodology to assess the proliferation resistance. The proliferation resistant characteristics of the DUPIC fuel cycle are assessed by applying the modified INPRO Methodology based on the developed evaluation metrics and acceptance criteria. The evaluation procedure and the metrics can be utilized as a reference for an evaluation of the proliferation resistance of a future innovative nuclear energy system

  20. Effect of chemical composition on corrosion resistance of Zircaloy fuel cladding tube for BWR

    International Nuclear Information System (INIS)

    Inagaki, Masahisa; Akahori, Kimihiko; Kuniya, Jirou; Masaoka, Isao; Suwa, Masateru; Maru, Akira; Yasuda, Teturou; Maki, Hideo.

    1990-01-01

    Effects of Fe and Ni contents on nodular corrosion susceptibility and hydrogen pick-up of Zircaloy were investigated. Total number of 31 Zr alloys having different chemical compositions; five Zr-Sn-Fe-Cr alloys, eight Zr-Sn-Fe-Ni alloys and eighteen Zr-Sn-Fe-Ni-Cr alloys, were melted and processed to thin plates for the corrosion tests in the environments of a high temperature (510degC) steam and a high temperature (288degC) water. In addition, four 450 kg ingots of Zr-Sn-Fe-Ni-Cr alloys were industrially melted and BWR fuel cladding tubes were manufactured through a current material processing sequence to study their producibility, tensile properties and corrosion resistance. Nodular corrosion susceptibility decreased with increasing Fe and Ni contents of Zircaloys. It was seen that the improved Zircaloys having Fe and Ni contents in the range of 0.30 [Ni]+0.15[Fe]≥0.045 (w%) showed no susceptibility to nodular corrosion. An increase of Fe content resulted in a decrease of hydrogen pick-up fraction in both steam and water environments. An increase of Fe and Ni content of Zircaloys in the range of Fe≤0.25 w% and Ni≤0.1 w% did not cause the changes in tensile properties and fabricabilities of fuel cladding tube. The fuel cladding tube of improved Zircaloy, containing more amount of Fe and Ni than the upper limit of Zircaloy-2 specification showed no susceptibility to nodular corrosion even in the 530degC steam test. (author)

  1. Methodologies for evaluating the proliferation resistance of nuclear fuel cycles

    International Nuclear Information System (INIS)

    Shiotani, Hiroki; Hori, Kei-ichiro; Takeda, Hiroshi

    2001-01-01

    The Japan Nuclear Cycle Development Institute (JNC) believes that the development of future nuclear fuel cycle technology should be conducted with careful consideration given to non-proliferation. JNC is studying methodologies for evaluating proliferation resistance of nuclear fuel cycle technologies. However, it is difficult to establish the methodology for evaluating proliferation resistance since the results greatly depend on the assumption for the evaluation and the surrounding conditions. This study grouped factors of proliferation resistance into categories through reviewing past studies and studied the relationships between the factors. Then, this study tried to find vulnerable nuclear material (plutonium) in some FBR fuel cycles from the proliferation perspective, and calculated the time it takes to convert the materials from various nuclear fuel cycles into pure plutonium metal under some assumptions. The result showed that it would take a long time to convert the nuclear materials from the FBR fuel cycles without plutonium separation. While it is a preliminary attempt to evaluate a technical factor of proliferation resistance as the basis of the institutional proliferation resistance, the JNC hopes that it will contribute to future discussions in this area. (author)

  2. Status of Methodology Development for the Evaluation of Proliferation Resistance

    International Nuclear Information System (INIS)

    Lee, Yong Deok; Ko, Won Il; Lee, Jung Won

    2010-01-01

    Concerning the increasing energy demand and green house effect, nuclear energy is now the most feasible option. Therefore, recently, oil countries even have a plan to build the nuclear power plant for energy production. If nuclear systems are to make a major and sustainable contribution to the worlds energy supply, future nuclear energy systems must meet specific requirements. One of the requirements is to satisfy the proliferation resistance condition in an entire nuclear system. Therefore, from the beginning of future nuclear energy system development, it is important to consider a proliferation resistance to prevent the diversion of nuclear materials. The misuse of a nuclear system must be considered as well. Moreover, in the import and export of nuclear system, the evaluation of the proliferation resistance on the nuclear system becomes a key factor The INPRO (International Project on Innovative Nuclear Reactors and Fuel Cycles) program initiated by the IAEA proposed proliferation resistance (PR) as a key component of a future innovative nuclear system (INS) with a sustainability, economics, safety of nuclear installation and waste management. The technical goal for Generation IV (Gen IV) nuclear energy systems (NESs) highlights a Proliferation Resistance and Physical Protection (PR and PP), sustainability, safety, reliability and economics as well. Based on INPRO and Gen IV study, the methodology development for the evaluation of proliferation resistance has been carried out in KAERI. Finally, the systematic procedure for methodology was setup and the indicators for the procedure were decided. The methodology involves the evaluation from total nuclear system to individual process. Therefore, in this study, the detailed procedure for the evaluation of proliferation resistance and the newly proposed additional indicators are described and several conditions are proposed to increase the proliferation resistance in the future nuclear system. The assessment of PR

  3. Effect of zinc injection on BWR fuel cladding corrosion. Pt. 1. Study on an accelerated corrosion condition to evaluate corrosion resistance of zircaloy-2 fuel cladding

    International Nuclear Information System (INIS)

    Kawamura, Hirotaka; Kanbe, Hiromu; Furuya, Masahiro

    2002-01-01

    Japanese BWR utilities have a plan to apply zinc injection to the primary coolant in order to reduce radioactivity accumulation on the structure. Prior to applying the zinc injection to BWR plants, it is necessary to evaluate the effect of zinc injection on corrosion resistance of fuel cladding. The objective of this report was to examine the accelerated corrosion condition for evaluation of BWR fuel cladding corrosion resistance under non-irradiated conditions, as the first step of a zinc injection evaluation study. A heat transfer corrosion test facility, in which a two phase flow condition could be achieved, was designed and constructed. The effects of heat flux, void fraction and solution temperature on BWR fuel cladding corrosion resistance were quantitatively investigated. The main findings were as follows. (1) In situ measurements using high speed camera and a void sensor together with one dimensional two phase flow analysis results showed that a two phase flow simulated BWR core condition can be obtained in the corrosion test facility. (2) The heat transfer corrosion test results showed that the thickness of the zirconium oxide layer increased with increasing solution temperature and was independent of heat flux and void fraction. The corrosion accelerating factor was about 2.5 times in the case of a temperature increase from 288degC to 350degC. (author)

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

    International Nuclear Information System (INIS)

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

    2003-06-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-06-01

    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.

  6. ASSESSING THE PROLIFERATION RESISTANCE OF INNOVATIVE NUCLEAR FUEL CYCLES

    International Nuclear Information System (INIS)

    BARI, R.; ROGLANS, J.; DENNING, R.; MLADINEO, S.

    2003-01-01

    The National Nuclear Security Administration is developing methods for nonproliferation assessments to support the development and implementation of U.S. nonproliferation policy. This paper summarizes the key results of that effort. Proliferation resistance is the degree of difficulty that a nuclear material, facility, process, or activity poses to the acquisition of one or more nuclear weapons. A top-level measure of proliferation resistance for a fuel cycle system is developed here from a hierarchy of metrics. At the lowest level, intrinsic and extrinsic barriers to proliferation are defined. These barriers are recommended as a means to characterize the proliferation characteristics of a fuel cycle. Because of the complexity of nonproliferation assessments, the problem is decomposed into: metrics to be computed, barriers to proliferation, and a finite set of threats. The spectrum of potential threats of nuclear proliferation is complex and ranges from small terrorist cells to industrialized countries with advanced nuclear fuel cycles. Two general categories of methods have historically been used for nonproliferation assessments: attribute analysis and scenario analysis. In the former, attributes of the systems being evaluated (often fuel cycle systems) are identified that affect their proliferation potential. For a particular system under consideration, the attributes are weighted subjectively. In scenario analysis, hypothesized scenarios of pathways to proliferation are examined. The analyst models the process undertaken by the proliferant to overcome barriers to proliferation and estimates the likelihood of success in achieving a proliferation objective. An attribute analysis approach should be used at the conceptual design level in the selection of fuel cycles that will receive significant investment for development. In the development of a detailed facility design, a scenario approach should be undertaken to reduce the potential for design vulnerabilities

  7. User requirements and criteria for proliferation resistance in INPRO

    International Nuclear Information System (INIS)

    Choi, K.; Shea, T.E.; Hurt, R.D.; Nishimura, R.

    2004-01-01

    In designing future nuclear energy systems, it is important to consider the potential that such systems could be misused for the purpose of producing nuclear weapons. INPRO set out to provide guidance on incorporating proliferation resistance into innovative nuclear energy systems (INS). Generally two types of proliferation resistance measures are distinguished: intrinsic and extrinsic. Intrinsic features consist of technical design features that reduce the attractiveness of nuclear material for nuclear weapon program, or prevent the diversion of nuclear material or production of undeclared nuclear material for nuclear weapons. Extrinsic measures include commitments, obligations and policies of states such as the Treaty on the Non-Proliferation of Nuclear Weapons (NPT) and IAEA safeguards agreements. INPRO has produced five basic principles and five user requirements for INS. It emphasizes that INS must continue to be an unattractive means to acquire fissile material for a nuclear weapon program. It also addresses as user requirements: 1) a balanced and optimised combination of intrinsic features and extrinsic measures, 2) the development and implementation of intrinsic features, 3) an early consideration of proliferation resistance in the development of INS and 4) the utilization of intrinsic features to increase the efficiency of extrinsic measures. INPRO has also developed criteria, consisting of indicators and acceptance limits, which would be used by a state to assess how an INS satisfies those user requirements. For the first user requirement, the most important but complex one, INPRO provides a 3-layer hierarchy of indicators to assess how unattractive a specific INS would be as part of a nuclear weapon program. Attributes of nuclear material and facilities are used as indicators to assess intrinsic features. Extrinsic measures imposed on the system are also assessed. Indicators to assess defence in depth for proliferation resistance include the number and

  8. Revised INPRO Methodology in the Area of Proliferation Resistance

    International Nuclear Information System (INIS)

    Park, J.H.; Lee, Y.D.; Yang, M.S.; Kim, J.K.; Haas, E.; Depisch, F.

    2008-01-01

    The official INPRO User Manual in the area of proliferation resistance is being processed for the evaluation of innovative nuclear energy systems. Proliferation resistance is one of the goals to be satisfied for future nuclear energy systems in INPRO. The features of currently updated and released INPRO methodology were introduced on basic principles, user requirements and indicators. The criteria for an acceptance limit were specified. The DUPIC fuel cycle was evaluated based on the updated INPRO methodology for the applicability of the INPRO User Manual. However, the INPRO methodology has some difficulty in quantifying the multiplicity and robustness as well as the total cost to improve proliferation resistance. Moreover, the integration method for the evaluation results still needs to be improved.

  9. Complexities in gauging time-dependency of proliferation resistance

    International Nuclear Information System (INIS)

    Avens, L.R.; Eller, P.G.; Stanbro, W.D.

    2004-01-01

    To a considerable extent, policy decisions on nuclear fuel cycle issues depend upon how decision makers recognize and weigh 'long-term' and 'short-term' nuclear proliferation risk factors. Priorities and structures of advanced fuel cycle and safeguards research and development programs are affected similarly. Unfortunately, there is a diversity of understanding of the precise meanings of these proliferation risk terms, leading to lack of precision in their usage. In addition, proliferation risk evaluation fundamentally involves value judgments on the relative importance of time-dependent risks. Poor communication and diverse conclusions often result. This paper explores some complexities in gauging 'long-term' and 'short-term' proliferation risk in the context of advanced nuclear fuel cycles. A convenient vehicle for this purpose is a commonly used notional plot of some proliferation resistance attribute of spent fuel or separated plutonium versus years from reactor discharge, often overlain with similar notional curves denoting multiple fuel irradiation and recycle. A common basis for misuse of such plots is failure to clearly define the range of proliferation threats being evaluated, as illustrated by several common examples of such omissions. Partial arguments of this type can be misleading and provide a disservice to policy makers who must have a clear picture of the tradeoffs being made. This paper concludes with a call for much greater care to avoid overly simplistic interpretations of notional proliferation-related concepts and greater precision in general in use of proliferation-related terminology.

  10. Analysis of nuclear proliferation resistance reprocessing and recycling technologies

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-05-01

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

  11. Analysis of nuclear proliferation resistance reprocessing and recycling technologies

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  12. Development of seamless forged pipe and fitting for BWR recirculation loop piping with improved resistance to intergranular stress corrosion cracking

    International Nuclear Information System (INIS)

    Ohnishi, Keizo; Tsukada, Hisashi; Kobayashi, Masayoshi; Iwadate, Tadao; Ono, Shinichi

    1981-01-01

    As a primary remedy for IGSCC of primary loop piping, especially Recirculation Loop Piping of BWR, extra low carbon stainless steel with high nitrogen content has become to be used. While, in order to make In-service Inspection easier and complete, new design of piping which decrease both number and total length of weld line has been considered. Japan Steel Works has developed the research on large size seamless forged pipe and fitting made from high nitrogen extra low carbon 316 stainless steel. This paper describes the key points of manufacturing technology as well as the material properties, especially strength and intergranular-corrosion and intergranular- stress-corrosion-cracking-resistivities of these forged pipe and fitting. (author)

  13. Technical features to enhance proliferation resistance of nuclear energy systems

    International Nuclear Information System (INIS)

    2010-01-01

    It is generally accepted that proliferation resistance is an essential issue for the continued development and sustainability of nuclear energy. Several comprehensive assessment activities on the proliferation resistance of the nuclear fuel cycle have previously been completed, notably the International Nuclear Fuel Cycle Evaluation (INFCE) carried out under the auspices of the IAEA, and the Non-proliferation Alternative Systems Assessment Program (NASAP) review carried out by the USA. There have been, however, relatively few comprehensive treatments of the issue following these efforts in the 1970s. However, interest in and concern about this issue have increased recently, particularly because of greater interest in innovative nuclear fuel cycles and systems. In 2000, the IAEA initiated the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) and the US Department of Energy initiated the Generation IV International Forum (GIF). These projects are aimed at the selection and development of concepts of innovative nuclear energy systems and fuel cycles. Proliferation resistance is one of the fundamental considerations for both projects. In this context, the IAEA in 2001 initiated a study entitled 'Technical Aspects of Increasing Proliferation Resistance of the Nuclear Fuel Cycle'. This task is not intended as an effort to assess the merits of a particular fuel cycle system for the future, but to describe a qualitative framework for an examination of the proliferation resistance provided by the intrinsic features of an innovative nuclear energy system and fuel cycle. This task also seeks to provide a high level survey of a variety of innovative nuclear energy systems and fuel cycles with respect to that framework. The concept of proliferation resistance is considered in terms of intrinsic features and extrinsic measures. The intrinsic features, sometimes referred to as the physical/technical aspects, are those features that result from the

  14. INCREASED PROLIFERATION RESISTANCE FOR 21ST CENTURY NUCLEAR POWER

    International Nuclear Information System (INIS)

    Demuth, Scott F.; Thomas, Ken E.; Wallace, Richard K.

    2007-01-01

    World energy demand and greenhouse gases are expected to significantly increase in the near future. Key developing countries have identified nuclear power as a major contributor to their future energy sources. Consequently, the US and others are currently exploring the concept of a Global Nuclear Energy Partnership (GNEP) to address the concerns of nuclear proliferation. This effort is also being encouraged by the International Atomic Energy Agency (IAEA). While the IAEA currently provides the framework for monitoring of state sponsored nuclear proliferation by way of international treaties, a complimentary action is to promote more proliferation resistant fuel cycles and advanced safeguards technology. As such, it is the responsibility of current technology owners to increase their nuclear fuel cycle proliferation resistance. For those countries that have an active and well-developed fuel cycle, it will require future enhancements. For those countries with extensive nuclear energy experience, yet less active programs, it requires re-engagement for technology development and deployment. The following paper discusses potential fuel cycle and technology changes that affect proliferation resistance; and consequently, may form the basis of future technology development efforts.

  15. Proliferation resistance criteria for fissile material disposition issues

    International Nuclear Information System (INIS)

    Rutherford, D.A.; Fearey, B.L.; Markin, J.T.; Close, D.A.; Tolk, K.M.; Mangan, D.L.; Moore, L.

    1995-01-01

    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

  16. Enhancing VVER annular proliferation resistance fuel with minor actinides

    International Nuclear Information System (INIS)

    Chang, G. S.

    2007-01-01

    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 2 38Pu and 2 40Pu isotopes ratio to enhance the proliferation resistance, and (b) use of transuranic nuclides ( 2 37Np and 2 41Am) in the high burnup fuel, which can drastically increase the proliferation resistance isotope ratio of 2 38Pu /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

  17. MEHODOLOGY FOR PROLIFERATION RESISTANCE FOR ADVANCE NUCLEAR ENERGY SYSTEMS

    International Nuclear Information System (INIS)

    YUE, M.; CHANG, L.Y.; BARI, R.

    2006-01-01

    The Technology Goals for Generation IV nuclear energy systems highlight Proliferation Resistance and Physical Protection (PRandPP) as one of the four goal areas for Generation 1V nuclear technology. Accordingly, an evaluation methodology is being developed by a PRandPP Experts Group. This paper presents a possible approach, which is based on Markov modeling, to the evaluation methodology for Generation IV nuclear energy systems being developed for PRandPP. Using the Markov model, a variety of proliferation scenarios can be constructed and the proliferation resistance measures can be quantified, particularly the probability of detection. To model the system with increased fidelity, the Markov model is further developed to incorporate multiple safeguards approaches in this paper. The approach to the determination of the associated parameters is presented. Evaluations of diversion scenarios for an example sodium fast reactor (ESFR) energy system are used to illustrate the methodology. The Markov model is particularly useful because it can provide the probability density function of the time it takes for the effort to be detected at a specific stage of the proliferation effort

  18. Energy-efficiency and proliferation-resistance assessment factors

    International Nuclear Information System (INIS)

    1979-02-01

    Assessment factors suggested with regard to energy efficiency are: preservation of natural non-renewable resources: the degree of security of supply which can be achieved; the availability of necessary raw materials and technology; economic feasibility; and acceptability of a fuel cycle from environmental and safety views. In the area of proliferation resistance, it is suggested that the basic element is the political commitment by a Government not to use imported nuclear materials and equipment to manufacture nuclear explosives. 100% proliferation resistance is considered unattainable in practice. The role of international safeguards in detering possible diversion through the risk of early detection is described, and it is argued that efficient safeguards will force a Government willing to go nuclear to withdraw from its safeguards agreements. The second assessment factor, accordingly, is to consider different fuel cycles with regard to the efficient and rapid building up of a nuclear weapons capacity once the country has withdrawn from its safeguards commitments

  19. 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)

  20. Proliferation resistance assessment of high temperature gas reactors

    International Nuclear Information System (INIS)

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

    2014-10-01

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

  1. Development of proliferation resistance assessment methodology based on international standard

    International Nuclear Information System (INIS)

    Ko, W. I.; Chang, H. L.; Lee, Y. D.; Lee, J. W.; Park, J. H.; Kim, Y. I.; Ryu, J. S.; Ko, H. S.; Lee, K. W.

    2012-04-01

    Nonproliferation is one of the main requirements to be satisfied by the advanced future nuclear energy systems that have been developed in the Generation IV and INPRO studies. The methodologies to evaluate proliferation resistance has been developed since 1980s, however, the systematic evaluation approach has begun from around 2000. Domestically a study to develop national method to evaluate proliferation resistance (PR) of advanced future nuclear energy systems has started in 2007 as one of the long-term nuclear R and D subjects in order to promote export and international credibility and transparency of national nuclear energy systems and nuclear fuel cycle technology development program. In the first phase (2007-2010) development and improvement of intrinsic evaluation parameters for the evaluation of proliferation resistance, quantification of evaluation parameters, development of evaluation models, and development of permissible ranges of evaluation parameters have been carried out. In the second phase (2010-2012) generic principle of to evaluate PR was established, and techincal guidelines, nuclear material diversion pathway analysis method, and a method to integrate evaluation parameters have been developed. which were applied to 5 alternative nuclear fuel cycles to estimate their appicability and objectivity. In addition, measures to enhance PR of advanced future nuclear energy systems and technical guidelines of PR assessment using intrinsic PR evaluation parameters were developed. Lastly, requlatory requirements to secure nonproliferation requirements of nuclear energy systems from the early design stage, operation and to decommissioning which will support the export of newly developed advanced future nuclear energy system

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

    International Nuclear Information System (INIS)

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

    2013-01-01

    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

  3. Assessment of proliferation resistance of thermal recycle systems

    International Nuclear Information System (INIS)

    1979-02-01

    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

  4. Proliferation resistance considerations for remote small modular reactors

    Energy Technology Data Exchange (ETDEWEB)

    Whitlock, J., E-mail: whitlockj@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Sprinkle, J., E-mail: j.sprinkle@iaea.org [International Atomic Energy Agency, Vienna (Austria)

    2013-07-01

    Remotely located Small Modular Reactors at the low end of energy production (on the order of 10 MWe, referenced here as Very Small Modular Reactors or VSMRs) present unique proliferation resistance advantages and challenges. Addressing these challenges in the most efficient manner may not only be desirable, but necessary, for development of this technology. Incorporation of safeguards considerations early in the design process (Safeguards by Design) along with safety, security, economics and other key drivers, is of importance. Operational Transparency may become an essential aspect of the safeguards approach for such systems. (author)

  5. 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.

  6. Proliferation resistance and physical protection working group: methodology and applications

    International Nuclear Information System (INIS)

    Bari, Robert A.; Whitlock, Jeremy J.; Therios, Ike U.; Peterson, P.F.

    2012-01-01

    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.

  7. A strategic framework for proliferation resistance: a systematic approach for the identification and evaluation of technology opportunities to enhance the proliferation resistance of civilian nuclear energy systems

    International Nuclear Information System (INIS)

    Hassberger, J.A.; Isaac, T.; Schock, R.N.

    2001-01-01

    The United State Department of Energy Nuclear Energy Research Advisory Committee recently completed a study ''Technological Opportunities To Increase The Proliferation Resistance Of Global Civilian Nuclear Power Systems (TOPS)''. That effort included the development of a set of both intrinsic and extrinsic barriers to proliferation that technologies can directly impact. In this paper we will review these barriers as and framework for assisting in the evaluation of the relative proliferation resistance of various nuclear fuel cycles, technologies and alternatives. (author)

  8. Teleconference highlights-NE-NA proliferation resistance review

    International Nuclear Information System (INIS)

    Miller, Michael C.

    2009-01-01

    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

  9. Flow resistance of orifices and spacers of BWR thermal-hydraulic and neutronic coupling loop

    International Nuclear Information System (INIS)

    Iguchi, Tadashi; Asaka, Hideaki; Nakamura, Hideo

    2002-03-01

    Authors are performing THYNC experiments to study thermal-hydraulic instability under neutronic and thermal-hydraulic coupling. In THYNC experiments, the orifices are installed at the exit of the test section and the spacers are installed in the test section, in order to properly simulate in-core thermal-hydraulics in the reactor core. It is necessary to know the flow resistance of the orifices and spacers for the analysis of THYNC experimental results. Consequently, authors measured the flow resistance of orifice and spacer under single-phase and two-phase flows. Using the experimental results, authors investigated the dependency of the flow resistances on the parameters, such as pressure, mass flux, an geometries. Furthermore, authors investigated the applicability of the basic two-phase flow models, for example the separate flow model, to the two-phase flow multiplier. As the result of the investigation on the single-phase flow experiment, it was found (1) that the effects of pressure and mass flux flow resistance are described by a function of Reynolds number, and (2) that flow resistances of the orifice and the spacer are calculated with the previous prediction methods. However, it was necessary to introduce an empirical coefficient, since it was difficult to predict accurately the flow resistance only with the previous prediction method due to the complicated geometry dependency, for example a flow area blockage ratio. On the other hand, according to the investigation on two-phase flow experiment, the followings were found. (1) Relation between the two-phase flow multiplier and the quality is regarded to be linear under pressure of 2MPa - 7MPa. The relation is dependent on pressure and geometry, and is little dependent on mass flux. (2) Relation between the two-phase flow multiplier and void fraction is little dependent on pressure, mass flux, and geometry under pressure of 0.2MPa - 7MPa and void fraction less than 0.6. The relation is less dependent on

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

    International Nuclear Information System (INIS)

    Schenkel, R.; Glatz, J.-P.; Magill, J.; Mayer, K.

    2001-01-01

    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''

    International Nuclear Information System (INIS)

    Brown, N W; Hassberger, J A.

    1999-01-01

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-07-01

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

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

    International Nuclear Information System (INIS)

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

    2000-07-01

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

  15. Development of Proliferation Resistance Assessment Methodology Based on International Standards

    International Nuclear Information System (INIS)

    Lee, Yong Deok; Lee, Jung Won; Lee, Kwang Seok

    2009-03-01

    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 year, feasibility study of indicator, allowable limit of indicator, review of technical requirement of indicator 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

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

    International Nuclear Information System (INIS)

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

    2008-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-03-15

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

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

    International Nuclear Information System (INIS)

    Gouveia, Fernando

    2010-01-01

    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

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

    International Nuclear Information System (INIS)

    Englert, Matthias

    2009-01-01

    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.)

  20. BWR 90: The ABB advanced BWR design

    International Nuclear Information System (INIS)

    Haukeland, S.; Ivung, B.; Pedersen, T.

    1999-01-01

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    Ko, Won Il; Chang, Hong Lae

    2010-07-01

    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

  3. A study on the proliferation resistance evaluation methodology for nuclear energy system

    International Nuclear Information System (INIS)

    Kim, Min Su

    2007-02-01

    The framework of proliferation resistance evaluation methodology, based on attribute analysis and scenario analysis, for nuclear energy system is suggested in order to allow for the comprehensive assessment of proliferation resistance by addressing the intrinsic and extrinsic features of nuclear energy system. Proliferation resistance is viewed within the context of the success tree model of proliferator's diversion attempt and expressed by the value of top event probability of the success tree model. This study focused on the method that the value of top event is estimated. The methodology uses two different methods to quantify the likelihood of basic events constituting the top event. The likelihood of basic event success affected by intrinsic feature of nuclear energy system was assessed by using multi-attribute utility theory and likelihood of basic event related to the diversion detection measures was assessed by direct expert elicitation. The value of top event was calculated based on the intersection of probabilities of basic event success. Feasibility of the methodology was explored by applying it to selected reference nuclear energy systems. System-Integrated Modular Advanced Reactor (SMART) system and Light Water Reactor (LWR) were chosen as reference systems and the value proliferation resistance of SMART and LWR were evaluated. Characteristics of inherent features and hypothesized safeguards measures of both systems were identified and used as input data to evaluate proliferation resistance. The results and conclusions are applicable only within the context of subjectivity of this methodology

  4. Acquisition/Diversion Pathway Analysis of the DUPIC Fuel Cycle for the Assessment of Proliferation Resistance

    International Nuclear Information System (INIS)

    Chang, Hong Lae; Ko, Won Il

    2008-01-01

    Within the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) of the IAEA, a methodology for evaluating proliferation resistance (INPRO PR methodology) has been developed in order to provide guidance in using the INPRO methodology. However, it remains to develop the methodology to evaluate User Requirements (UR) 4 regarding multiplicity and robustness of barriers against proliferation (innovative nuclear energy systems should incorporate multiple proliferation resistance features and measures). To develop the assessment procedure and metrics for User Requirement 4 (UR4), the coarse acquisition/ diversion pathway analysis of the DUPIC Fuel Cycle has been performed. The most plausible pathways for the acquisition of weapons-usable nuclear material were identified and analyzed using a systematic approach herein, and future work to complete the assessment approach for the UR4 of the INPRO methodology regarding the multiplicity and robustness of barriers against proliferation are also proposed

  5. Acquisition/Diversion Pathway Analysis for the Assessment of Proliferation Resistance

    International Nuclear Information System (INIS)

    Chang, Hong Lae; Ko, Won Il

    2009-01-01

    The INPRO methodology in the area of proliferation resistance (PR) has one basic principle and five user requirements with relevant criteria, indicators, evaluation parameters, etc. The two Korean case studies on proliferation resistance of the DUPIC fuel cycle during 2004 and 2005 and various consultancy meetings have contributed to the establishment of the assessment metrics and procedures for three user requirements regarding States' commitment, attractiveness of nuclear material and technology, and difficulty and detectability of diversion. However, the assessment indicators and procedure for user requirement 4 regarding multiplicity and robustness of barriers against proliferation still need to be developed. In this paper, a systematic approach to identify and analyze the acquisition/diversion pathways in a nuclear energy system is described, including follow-up R and D plans to assess the multiplicity and robustness of barriers against proliferation

  6. Proliferation resistance of advanced sustainable nuclear fuel cycles

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-01

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

  7. Proliferation resistance of advanced sustainable nuclear fuel cycles

    International Nuclear Information System (INIS)

    Garcia, H.E.; Lineberry, M.J.; Aumeier, S.E.; McFarlane, H.F.

    2001-01-01

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

  8. The BWR Stability Issue

    International Nuclear Information System (INIS)

    D'Auria, F.

    2008-01-01

    The purpose of this paper is to supply general information about Boiling Water Reactor (BWR) stability. The main concerned topics are: phenomenological aspects, experimental database, modelling features and capabilities, numerical models, three-dimensional modelling, BWR system performance during stability, stability monitoring and licensing aspects.

  9. Proliferation Resistance: Acquisition/Diversion Pathway Analysis for the DUPIC Fuel Cycle

    International Nuclear Information System (INIS)

    Ko, Won Il; Chang, Hong Lae; Song, Dae Yong; Lee, Ho Hee; Kwon, Eun Ha; Jeong, Chang Joon; Kim, Ho Dong

    2009-07-01

    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, it remains to develop the methodology to evaluate User Requirements (UR) 4 regarding multiplicity and robustness of barriers against proliferation - innovative nuclear energy systems should incorporate multiple proliferation resistance features and measures. Since this requires an acquisition/diversion pathway analysis, this report describes a systematic approach developed for the identification and analysis of pathways for the acquisition of weapons-usable nuclear material using the DUPIC fuel cycle system. At the first step, the objectives of the proliferation were identified, including the quality and quantity of the material, the time required to acquire the material for the proliferation, thr capability of the potential proliferant country, etc. At the second step, the possible strategies, which the potential proliferant country could adopt, were identified: undeclared removal of nuclear material from the fuel cycle facilities; and further treatment of the diverted nuclear materials needed to acquire weapons-usable materials. At the final step, a systematic approach to select the plausible pathways for the acquisition/diversion of nuclear material during the whole fuel cycle has been developed. The coarse material diversion pathways for the DUPIC fuel cycle and the approach developed was reviewed and discussed at the experts meeting at the IAEA for its appropriateness and comprehensiveness

  10. Application of the Decision Tree Modeling Approach to Evaluation of Proliferation Resistance

    International Nuclear Information System (INIS)

    Coles, Garill A.; Zentner, Michael D.

    2007-01-01

    An experts group was created in 2002 by The Generation IV International Forum for the purpose of developing an internationally accepted methodology for assessing the proliferation resistance of a nuclear energy system (NES) and its individual elements. After three years of work, and some limited demonstration studies, a pilot study was initiated to exercise the methodologies being developed by assessing the proliferation resistance of a specific portion of a hypothetical NES called the Example Sodium Fast Reactor (ESFR). This paper summarizes the assessment approach, and describes the next steps to be taken in the development of the methodology.

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

    International Nuclear Information System (INIS)

    2012-05-01

    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.

  12. Compact modular BWR (CM-BWR)

    International Nuclear Information System (INIS)

    Fennern, Larry; Boardman, Charles; Carroll, Douglas G.; Hida, Takahiko

    2003-01-01

    A preliminary assessment has shown that a small 350 MWe BWR reactor can be placed within a close fitting steel containment vessel that is 7.1 meters inside diameter. This allows the technology and manufacturing capability currently used to fabricate large ABWR reactor vessels to be used to provide a factory fabricated containment vessel for a 350 MWe BWR. When a close fitted steel containment is combined with a passive closed loop isolation condenser system and a natural circulating reactor system that contains a large water inventory, primary system leaks cannot uncover the core. This eliminates many of the safety systems needed in response to a LOCA that are common to large, conventional plant designs including. Emergency Core Flooding, Automatic Depressurization System, Active Residual Heat Removal, Safety Related Auxiliary Cooling, Safety Related Diesel Generators, Hydrogen Re-Combiners, Ex-vessel Core Retention and Cooling. By fabricating the containment in a factory and eliminating most of the conventional safety systems, the construction schedule is shortened and the capital cost reduced to levels that would not otherwise be possible for a relatively small modular BWR. This makes the CM-BWR a candidate for applications where smaller incremental power additions are desired relative to a large ALWR or where the local infrastructure is not able to accommodate a conventional ALWR plant rated at 1350 MWe or more. This paper presents a preliminary design description of a Compact Modular BWR (CM-BWR) whose design features dramatically reduce the size and cost of the reactor building and associated safety systems. (author)

  13. Proliferation resistance characteristics of advanced nuclear energy systems: a safeguard ability point of view

    International Nuclear Information System (INIS)

    Sevini, F.; Cojazzi, G.G.M.; Renda, G.

    2008-01-01

    Among the international community there is a renewed interest in nuclear power systems as a major source for energy production in the near to mid future. This is mainly due to concerns connected with future availability of conventional energy resources, and with the environmental impact of fossil fuels. International initiatives have been set up like the Generation 4. International Forum (GIF), the International Project on Innovative Nuclear Reactors and Fuel Cycles (IAEA-INPRO), and, partially, the US driven Global Nuclear Energy Partnership (GNEP), aimed at defining and evaluating the characteristics, in which future innovative nuclear energy systems (INS) will have to excel. Among the identified characteristics, Proliferation Resistance plays an important role for being able to widely deploy nuclear technology worldwide in a secure way. Studies having the objective to assess Proliferation Resistance of nuclear fuel cycles have been carried out since the nineteen seventies, e.g., the International Nuclear Fuel Cycle Evaluation (INFCE) and the Non-proliferation Alternative Systems Assessment Program (NASAP) initiatives, and all agree in stating that absolute intrinsic proliferation resistance, although desirable, is not achievable in the foreseeable future. The above finding is still valid; as a consequence, every INS will have to comply with agreements related to the Non Proliferation Treaty (NPT) and will require safeguards measures, implemented through extrinsic measures. This consideration led to a renewed interest in the Safeguard ability concept which can be seen as a bridge between intrinsic features and extrinsic features and measures.

  14. Improving the proliferation resistance of research and test reactors

    International Nuclear Information System (INIS)

    Lewis, R.A.

    1978-01-01

    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 235 U loading in the reduced-enrichment fuel elements be the same as the 235 U 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

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

    International Nuclear Information System (INIS)

    Wang, Yuxia; Gao, Ying; Cheng, Hairong; Yang, Guichun; Tan, Wenhua

    2015-01-01

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

  16. Methodological considerations in evaluating a proliferation resistance of innovative nuclear energy systems

    International Nuclear Information System (INIS)

    Kikuchi, Masahiro; Takaki, Naoyuki; Murajiri, Masahiro; Nakagome, Yoshihiro; Tokiwai, Moriyasu

    2004-01-01

    Over 25 years ago, INFCE studied the evaluation methodology of proliferation resistance. Recently, INPRO and GEN-IV coordinated by the IAEA and the USDOE respectively seek an appropriate innovative fuel cycle system for next generation that is furnished safer, sustainable, economical and reliable features. The evaluation methodology of the proliferation resistance is also assigned as an essential part of both studies. The IAEA established and has been strictly implementing the verification measures with accurate material accountancy system from the early of the 1970s in order to detect diversion of plutonium that is individually separated from irradiated nuclear material and recycled as MOX fuel. This paper firstly identifies the impedibility of intrinsic features of innovative fuel cycles and the safeguardability of selected nonproliferation measures as two individual essential parameters for evaluation of a proliferation resistance capability. As a next step, this paper also shows methodological considerations in evaluating the proliferation resistance levels as a multiple model of several clusters that are identified the ability of each parameter. (author)

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

    International Nuclear Information System (INIS)

    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

  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)

    Haas, E.; Chang, H.-L.; Phillips, J.R.; Listner, C.

    2015-01-01

    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. Peroxisome Proliferator-Activated Receptors and Hepatitis C Virus-Induced Insulin Resistance

    Directory of Open Access Journals (Sweden)

    Francesco Negro

    2009-01-01

    Full Text Available Insulin resistance and type 2 diabetes are associated with hepatitis C virus infection. A wealth of clinical and experimental data suggests that the virus is directly interfering with the insulin signalling in hepatocytes. In the case of at least one viral genotype (the type 3a, insulin resistance seems to be directly mediated by the downregulation of the peroxisome proliferator-activated receptor γ. Whether and how this interaction may be manipulated pharmacologically, in order to improve the responsiveness to antivirals of insulin resistant chronic hepatitis C, patients remain to be fully explored.

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

    International Nuclear Information System (INIS)

    Renda, Guido; Alim, Fatih; Cojazzi, Giacomo GM.

    2015-01-01

    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

  1. Advanced core concepts with enhanced proliferation resistance by transmutation of minor actinides

    International Nuclear Information System (INIS)

    Saito, Masaki

    2005-01-01

    ''Protected Plutonium Production (P 3 )'' has been proposed to establish high burn-up cores and to produce protected with high proliferation resistance due to high decay heat and large number of spontaneous fission neutron of 238 Pu by the transmutation of Minor Actinides (MAs) which is presently treated as high-level waste. The burn-up calculations have shown that the advanced fuel with UO 2 (11-13% enrichment of 235 U) by doping 237 Np to produce 238 Pu in the commercialized large LWRs burn up to 100 GWd/t with 238 Pu to Pu ratio of about 20% which means the fuel is highly protected from proliferation. It was also predicted that medium or small size LWR cores with 15-17% enrichment, liquid metal cooled cores, and gas cooled cores added by 1-2% Np could achieve 100 GWd/t burning with bearing high proliferation resistance. The 237 Np mass balance calculations have revealed that more than 20 nuclear P 3 plants of 300 MWe could be supplied with enough 237 Np from the Japanese commercial plants in equilibrium fuel cycles. From the present studies, it is confirmed that MAs are treated as burnable and fertile materials not only to extend the core life but also to improve plutonium proliferation resistance of the future nuclear energy systems instead of their geological disposal or just their burning through fission. (author)

  2. Nuclear proliferation-resistance and safeguards for future nuclear fuel cycle

    International Nuclear Information System (INIS)

    Kuno, Y.; Inoue, N.; Senzaki, M.

    2009-01-01

    Corresponding to the world nuclear security concerns, future nuclear fuel cycle (NFC) should have high proliferation-resistance (PR) and physical protection (PP), while promotion of the peaceful use of the nuclear energy must not be inhibited. In order to accomplish nuclear non-proliferation from NFC, a few models of the well-PR systems should be developed so that international community can recognize them as worldwide norms. To find a good balance of 'safeguard-ability (so-called extrinsic measure or institutional barrier)' and 'impede-ability (intrinsic feature or technical barrier)' will come to be essential for NFC designers to optimize civilian nuclear technology with nuclear non-proliferation, although the advanced safeguards with high detectability can still play a dominant role for PR in the states complying with full institutional controls. Accomplishment of such goal in a good economic efficiency is a future key challenge

  3. Nuclear proliferation-resistance and safeguards for future nuclear fuel cycle

    Science.gov (United States)

    Kuno, Y.; Inoue, N.; Senzaki, M.

    2009-03-01

    Corresponding to the world nuclear security concerns, future nuclear fuel cycle (NFC) should have high proliferation-resistance (PR) and physical protection (PP), while promotion of the peaceful use of the nuclear energy must not be inhibited. In order to accomplish nuclear non-proliferation from NFC, a few models of the well-PR systems should be developed so that international community can recognize them as worldwide norms. To find a good balance of 'safeguard-ability (so-called extrinsic measure or institutional barrier)' and 'impede-ability (intrinsic feature or technical barrier)' will come to be essential for NFC designers to optimize civilian nuclear technology with nuclear non-proliferation, although the advanced safeguards with high detectability can still play a dominant role for PR in the states complying with full institutional controls. Accomplishment of such goal in a good economic efficiency is a future key challenge.

  4. Nuclear proliferation-resistance and safeguards for future nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Kuno, Y. [Japan Atomic Energy Agency (JAEA) Nuclear-Non-proliferation Science and Technology Centre (NPSTC), 2-4 Shirane Shirakata, Tokai-mura, Ibaraki, 319-1195 (Japan); University of Tokyo, Nuclear Engineering and Management, 7-3-1 Hongo, Bunkyo-ku, Tokyo 113-8656 (Japan)], E-mail: kuno.yusuke@jaea.go.jp; Inoue, N. [Japan Atomic Energy Agency (JAEA) Nuclear-Non-proliferation Science and Technology Centre (NPSTC), 2-4 Shirane Shirakata, Tokai-mura, Ibaraki, 319-1195 (Japan); University of Tokyo, Nuclear Engineering and Management, 7-3-1 Hongo, Bunkyo-ku, Tokyo 113-8656 (Japan); Senzaki, M. [Japan Atomic Energy Agency (JAEA) Nuclear-Non-proliferation Science and Technology Centre (NPSTC), 2-4 Shirane Shirakata, Tokai-mura, Ibaraki, 319-1195 (Japan)

    2009-03-15

    Corresponding to the world nuclear security concerns, future nuclear fuel cycle (NFC) should have high proliferation-resistance (PR) and physical protection (PP), while promotion of the peaceful use of the nuclear energy must not be inhibited. In order to accomplish nuclear non-proliferation from NFC, a few models of the well-PR systems should be developed so that international community can recognize them as worldwide norms. To find a good balance of 'safeguard-ability (so-called extrinsic measure or institutional barrier)' and 'impede-ability (intrinsic feature or technical barrier)' will come to be essential for NFC designers to optimize civilian nuclear technology with nuclear non-proliferation, although the advanced safeguards with high detectability can still play a dominant role for PR in the states complying with full institutional controls. Accomplishment of such goal in a good economic efficiency is a future key challenge.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-12-07

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

  6. Assessment of proliferation resistances of aqueous reprocessing techniques using the TOPS methodology

    International Nuclear Information System (INIS)

    Åberg Lindell, M.; Grape, S.; Håkansson, A.; Jacobsson Svärd, S.

    2013-01-01

    Highlights: • Proliferation resistances of three possible LFR fuel cycles are assessed. • The TOPS methodology has been chosen for the PR assessment. • Reactor operation, reprocessing and fuel fabrication are examined. • Purex, Ganex, and a combination of Purex, Diamex and Sanex, are compared. • The safeguards analysis speaks in favor of Ganex as opposed to the Purex process. - Abstract: The aim of this study is to assess and compare the proliferation resistances (PR) of three possible Generation IV lead-cooled fast reactor fuel cycles, involving the reprocessing techniques Purex, Ganex and a combination of Purex, Diamex and Sanex, respectively. The examined fuel cycle stages are reactor operation, reprocessing and fuel fabrication. The TOPS methodology has been chosen for the PR assessment, and the only threat studied is the case where a technically advanced state diverts nuclear material covertly. According to the TOPS methodology, the facilities have been divided into segments, here roughly representing the different forms of nuclear material occurring in each examined fuel cycle stage. For each segment, various proliferation barriers have been assessed. The results make it possible to pinpoint where the facilities can be improved. The results show that the proliferation resistance of a fuel cycle involving recycling of minor actinides is higher than for the traditional Purex reprocessing cycle. Furthermore, for the purpose of nuclear safeguards, group actinide extraction should be preferred over reprocessing options where pure plutonium streams occur. This is due to the fact that a solution containing minor actinides is less attractive to a proliferator than a pure Pu solution. Thus, the safeguards analysis speaks in favor of Ganex as opposed to the Purex process

  7. Low-decontamination approach to a proliferation-resistant fuel cycle

    International Nuclear Information System (INIS)

    Asquith, J.G.; Grantham, L.F.

    1978-01-01

    To prevent the diversion of nuclear material from power production to weapon production either by a nation or by clandestine groups within a nation, the nuclear fuel cycle must be proliferation-resistant and safeguarded. Potentially proliferation-resistant and safeguarded fuel cycles based on low-decontamination pyroreprocessing have been developed for the light water reactor (LWR), fast breeder reactor (FBR), and FBR-LWR combination. The major penalty for recycling fission products to the LWR is that fuel enrichment must be somewhat greater to overcome parasitic fission product absorption of neutrons. In the FBR, the major penalty is a slight reduction in breeding ratio due to the displacement of fertile material by fission products. Preliminary cost analysis indicates that these fuel cycles are economically competitive with fuel cycles using conventional reprocessing or those using virgin uranium if spent fuel storage costs are considered

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

    International Nuclear Information System (INIS)

    Sagara, Hiroshi; Saito, Masaki; Peryoga, Yoga; Ezoubtchenko, Alexey; Takivayev, Alan

    2005-01-01

    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 238 Pu 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)

  9. A Modified Nitride-Based Fuel for Long Core Life and Proliferation Resistance

    International Nuclear Information System (INIS)

    Ebbinghaus, B; Choi, J; Meier, T

    2003-01-01

    A modified nitride-based uranium fuel to support the small, secured, transportable, and autonomous reactor (SSTAR) concept is initiated at Lawrence Livermore National laboratory (LLNL). This project centers on the evaluation of modified uranium nitride fuels imbedded with other inert (e.g. ZrN), neutron-absorbing (e.g. HfN) , or breeding (e.g. ThN) nitrides to enhance the fuel properties to achieve long core life with a compact reactor design. A long-life fuel could minimize the need for on-site refueling and spent-fuel storage. As a result, it could significantly improve the proliferation resistance of the reactor/fuel systems. This paper discusses the potential benefits and detriments of modified nitride-based fuels using the criteria of compactness, long-life, proliferation resistance, fuel safety, and waste management. Benefits and detriments are then considered in recommending a select set of compositions for further study

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

    International Nuclear Information System (INIS)

    Bari, R.

    2008-01-01

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

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

    International Nuclear Information System (INIS)

    Moses, David Lewis

    2011-01-01

    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 and 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 and 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 and 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 and 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

  12. 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.)

  13. Methodology for proliferation resistance and physical protection of Generation IV nuclear energy systems

    International Nuclear Information System (INIS)

    Bari, R.; Peterson, P.; Nishimura, R.; Roglans-Ribas, J.

    2005-01-01

    Enhanced proliferation resistance and physical protection (PR and PP) is one of the technology goals for advanced nuclear concepts. Under the auspices of the Generation IV International Forum an international experts group has been chartered to develop an evaluation methodology for PR and PP. This methodology will permit an objective PR and PP comparison between alternative nuclear systems and support design optimization to enhance robustness against proliferation, theft and sabotage. The assessment framework consists of identifying the threats to be considered, defining the PR and PP measures required to evaluate the resistance of a nuclear system to proliferation, theft or sabotage, and establishing quantitative methods to evaluate the proposed measures. The defined PR and PP measures are based on the design of the system (e.g., materials, processes, facilities), and institutional measures (e.g., safeguards, access control). The assessment methodology uses analysis of pathways' with respect to specific threats to determine the PR and PP measures. Analysis requires definition of the threats (i.e. objective, capability, strategy), decomposition of the system into its relevant elements (e.g., reactor core, fuel recycle facility, fuel storage), and identification of targets. (author)

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

    International Nuclear Information System (INIS)

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

    2010-01-01

    We present an overview of the program on the evaluation methodology for proliferation resistance and physical protection (PR and 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 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.

  15. Multi-component Self-Consistent Nuclear Energy System: On proliferation resistance aspect

    International Nuclear Information System (INIS)

    Shmelev, A.; Saito, M; Artisyuk, V.

    2000-01-01

    Self-Consistent Nuclear Energy System (SCNES) that simultaneously meets four requirements: energy production, fuel production, burning of radionuclides and safety is targeted at harmonization of nuclear energy technology with human environment. The main bulk of SCNES studies focus on a potential of fast reactor (FR) in generating neutron excess to keep suitable neutron balance. Proliferation resistance was implicitly anticipated in a fuel cycle with co-processing of Pu, minor actinides (MA) and some relatively short-lived fission products (FP). In a contrast to such a mono-component system, the present paper advertises advantage of incorporating accelerator and fusion driven neutron sources which could drastically improve characteristics of nuclear waste incineration. What important is that they could help in creating advanced Np and Pa containing fuels with double protection against uncontrolled proliferation. The first level of protection deals with possibility to approach long life core (LLC) in fission reactors. Extending the core life-time to reactor-time is beneficial from the proliferation resistance viewpoint since LLC would not necessarily require fuel management at energy producing site, with potential advantage of being moved to vendor site for spent fuel refabrication. Second level is provided by the presence of substantial amounts of 238 Pu and 232 U in these fuels that makes fissile nuclides in them isotopically protected. All this reveals an important advantage of a multi-component SCNES that could draw in developing countries without elaborated technological infrastructure. (author)

  16. Development of advanced BWR

    International Nuclear Information System (INIS)

    Toyota, Masatoshi

    1982-01-01

    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.)

  17. DEVELOPMENT OF A METHODOLOGY TO ASSESS PROLIFERATION RESISTANCE AND PHYSICAL PROTECTION FOR GENERATION IV SYSTEMS

    International Nuclear Information System (INIS)

    Nishimura, R.; Bari, R.; Peterson, P.; Roglans-Ribas, J.; Kalenchuk, D.

    2004-01-01

    Enhanced proliferation resistance and physical protection (PR and PP) is one of the technology goals for advanced nuclear concepts, such as Generation IV systems. Under the auspices of the Generation IV International Forum, the Office of Nuclear Energy, Science and Technology of the U.S. DOE, the Office of Nonproliferation Policy of the National Nuclear Security Administration, and participating organizations from six other countries are sponsoring an international working group to develop an evaluation methodology for PR and PP. This methodology will permit an objective PR and PP comparison between alternative nuclear systems (e.g., different reactor types or fuel cycles) and support design optimization to enhance robustness against proliferation, theft and sabotage. The paper summarizes the proposed assessment methodology including the assessment framework, measures used to express the PR and PP characteristics of the system, threat definition, system element and target identification, pathway identification and analysis, and estimation of the measures

  18. Enzalutamide inhibits proliferation of gemcitabine-resistant bladder cancer cells with increased androgen receptor expression.

    Science.gov (United States)

    Kameyama, Koji; Horie, Kengo; Mizutani, Kosuke; Kato, Taku; Fujita, Yasunori; Kawakami, Kyojiro; Kojima, Toshio; Miyazaki, Tatsuhiko; Deguchi, Takashi; Ito, Masafumi

    2017-01-01

    Advanced bladder cancer is treated mainly with gemcitabine and cisplatin, but most patients eventually become resistance. Androgen receptor (AR) signaling has been implicated in bladder cancer as well as other types of cancer including prostate cancer. In this study, we investigated the expression and role of AR in gemcitabine-resistant bladder cancer cells and also the potential of enzalutamide, an AR inhibitor, as a therapeutic for the chemoresistance. First of all, we established gemcitabine-resistant T24 cells (T24GR) from T24 bladder cancer cells and performed gene expression profiling. Microarray analysis revealed upregulation of AR expression in T24GR cells compared with T24 cells. AR mRNA and protein expression was confirmed to be increased in T24GR cells, respectively, by quantitative RT-PCR and western blot analysis, which was associated with more potent AR transcriptional activity as measured by luciferase reporter assay. The copy number of AR gene in T24GR cells determined by PCR was twice as many as that of T24 cells. AR silencing by siRNA transfection resulted in inhibition of proliferation of T24GR cells. Cell culture in charcoal-stripped serum and treatment with enzalutamide inhibited growth of T24GR cells, which was accompanied by cell cycle arrest. AR transcriptional activity was found to be reduced in T24GR cells by enzalutamide treatment. Lastly, enzalutamide also inhibited cell proliferation of HTB5 bladder cancer cells that express AR and possess intrinsic resistance to gemcitabine. Our results suggest that enzalutamide may have the potential to treat patients with advanced gemcitabine-resistant bladder cancer with increased AR expression.

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

    Energy Technology Data Exchange (ETDEWEB)

    Whitlock, Jeremy [Non-Proliferation and Safeguards, AECL Chalk River Laboratories, Stn. 91, Chalk River, Ontario, K0J 1J0 (Canada); Inoue, Naoko; Senzaki, Masao [Japan Atomic Energy Agency - JAEA (Japan); Bley, Dennis [Buttonwood Consulting Inc., Oakton, VA (United States); Wonder, Ed [National Nuclear Security Administration, Department of Energy (United States)

    2009-06-15

    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

  20. Activated STAT5 Confers Resistance to Intestinal Injury by Increasing Intestinal Stem Cell Proliferation and Regeneration

    Directory of Open Access Journals (Sweden)

    Shila Gilbert

    2015-02-01

    Full Text Available Intestinal epithelial stem cells (IESCs control the intestinal homeostatic response to inflammation and regeneration. The underlying mechanisms are unclear. Cytokine-STAT5 signaling regulates intestinal epithelial homeostasis and responses to injury. We link STAT5 signaling to IESC replenishment upon injury by depletion or activation of Stat5 transcription factor. We found that depletion of Stat5 led to deregulation of IESC marker expression and decreased LGR5+ IESC proliferation. STAT5-deficient mice exhibited worse intestinal histology and impaired crypt regeneration after γ-irradiation. We generated a transgenic mouse model with inducible expression of constitutively active Stat5. In contrast to Stat5 depletion, activation of STAT5 increased IESC proliferation, accelerated crypt regeneration, and conferred resistance to intestinal injury. Furthermore, ectopic activation of STAT5 in mouse or human stem cells promoted LGR5+ IESC self-renewal. Accordingly, STAT5 promotes IESC proliferation and regeneration to mitigate intestinal inflammation. STAT5 is a functional therapeutic target to improve the IESC regenerative response to gut injury.

  1. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Chang, H. L.; Kwon, E. H.; Ahn, S. K.; Ko, W. I.; Kim, H. D. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    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.

  3. 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

  4. BWR stability analysis

    International Nuclear Information System (INIS)

    Valtonen, K.

    1990-01-01

    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

  5. TRAC-BWR development

    International Nuclear Information System (INIS)

    Weaver, W.L.; Rouhani, S.Z.

    1983-01-01

    The TRAC-BD1/MOD1 code containing many new or improved models has been assembled and is undergoing developmental assessment and testing and should be available shortly. The preparation of the manual for this code version is underway and should be available to the USNRC and their designated contractors by April of 1984. Finally work is currently underway on a fast running version of TRAC-BWR which will contain a one-dimensional neutron kinetics model

  6. BWR AXIAL PROFILE

    International Nuclear Information System (INIS)

    Huffer, J.

    2004-01-01

    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

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

    Science.gov (United States)

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

    2015-11-15

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

  8. Evaluating the attractiveness of nuclear material for proliferation-resistance and nuclear security

    International Nuclear Information System (INIS)

    Choi, Jor-Shan; Ikegame, Kou; Kuno, Yusuke

    2011-01-01

    The attractiveness of nuclear material, defined as a function of the isotopic composition of the nuclear material in formulas expressing the material's intrinsic properties, is of considerably debate in recent developments of proliferation-resistance measures of a nuclear energy system. A reason for such debate arises from the fact that the concept of nuclear material attractiveness can be confusing because the desirability of a material for nuclear explosive use depends on many tangible and intangible factors including the intent and capability of the adversary. In addition, a material that is unattractive to an advanced nation (in the case of proliferation) may be very attractive to a terrorist (in the case of physical protection and nuclear security). Hence, the concept of 'Nuclear Material Attractiveness' for different nuclear materials must be considered in the context of safeguards and security. The development of a ranking scheme on the attractiveness of nuclear materials could be a useful concept to start-off the strategies for safeguards and security on a new footing (i.e., why and how nuclear material is attractive, and what are the quantifiable basis). Japan may benefit from such concept regarding the attractiveness of nuclear materials when recovering nuclear materials from the damaged cores in Fukushima because safety, security, and safeguards (3S) would be a prominent consideration for the recovery operation, and it would be the first time such operation is performed in a non-nuclear weapons state. (author)

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

    International Nuclear Information System (INIS)

    Inoue, Naoko; Kawakubo, Yoko; Seya, Michio; Suzuki, Mitsutoshi; Kuno, Yusuke; Senzaki, Masao

    2010-01-01

    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)

  10. The encapsulated nuclear heat source reactor for proliferation-resistant nuclear energy

    International Nuclear Information System (INIS)

    Brown, N.W.; Hossain, Q.; Carelli, M.D.; Conway, L.; Dzodzo, M.; Greenspan, E.; Saphier, D.

    2001-01-01

    The encapsulated nuclear heat source (ENHS) is a modular reactor that was selected by the 1999 DOE NERI program as a candidate ''Generation-IV'' reactor concept. It is a fast neutron spectrum reactor cooled by Pb-Bi using natural circulation. It is designed for passive load following, for high level of passive safety, and for 15 years without refueling. One of the unique features of the ENHS is that the fission-generated heat is transferred from the primary coolant to the secondary coolant across the reactor vessel wall by conduction-providing for an essentially sealed module that is easy to install and replace. Because the fuel is encapsulated within a heavy steel container throughout its life it provides a unique improvement to the proliferation resistance of the nuclear fuel cycle. This paper presents the innovative technology of the ENHS. (author)

  11. Containment and Surveillance and Physical Protection Updates for Proliferation Resistance Analysis Using PRAETOR

    International Nuclear Information System (INIS)

    Chirayath, S.; Charlton, W.; Elmore, R.

    2015-01-01

    The Proliferation Resistance Analysis and Evaluation Tool for Observed Risk (PRAETOR) software code assesses the proliferation resistance (PR) of nuclear fuel cycle (NFC) systems. The Nuclear Security Science and Policy Institute (NSSPI) at Texas A&M University developed PRAETOR based on the well-established multi-attribute utility analysis (MAUA) methodology. MAUA methods facilitate compiling multiple PR characteristics into tiered PRAETOR output PR metrics enabling easier decision making at the analyst, program manager, and policy maker levels. PRAETOR uses intrinsic and extrinsic PR attributes to evaluate NFC systems. The PRAETOR 1.0 code originally had 63 attribute inputs representing the NFC system. The attribute input values assigned by the user are mapped to a utility value between 0 and 1 using utility functions. Each attribute has an associated weight obtained through a survey. Larger PRAETOR utility values indicate higher NFC system PR. An updated version of PRAETOR (Version 2.0) added seven more attribute inputs representing the nuclear security PR aspects of: (1) physical protection systems (PPS) and (2) containment and surveillance (C&S). The applicability of PRAETOR is demonstrated through a set of case studies. Two cases of Pressurized Water Reactor (PWR)used fuel assemblies with different cooling times were considered in this paper: (a) non-cooled fuel assemblies, and (b) 30-year cooled fuel assemblies. The case studies consider the new PPS and C&S attributes with low and high utility values. The PR results for the case studies with the updated PRAETOR were compared with those without the PPS and C&S attributes. The new attributes increased overall PR value by about 10% for case (a) and decreased it by about 3% in case (b). The importance of adding new attributes capturing physical protection and containment & surveillance is established. (author)

  12. BWR internals life assurance

    International Nuclear Information System (INIS)

    Herrera, M.L.; Stancavage, P.P.

    1988-01-01

    Boiling water reactor (BWR) internal components play an important role in power plant life extension. Many important internals were not designed for easy removal and changes in material properties and local environmental effects due to high radiation makes stress corrosion cracking more likely and more difficult to correct. Over the past several years, operating experience has shown that inspection, monitoring and refurbishment can be accomplished for internal structures with existing technology. In addition, mitigation techniques which address the causes of degradation are available to assure that life extension targets can be met. This paper describes the many considerations and aspects when evaluating life extension for reactor vessel internals

  13. BWR condensate filtration studies

    International Nuclear Information System (INIS)

    Wilson, J.A.; Pasricha, A.; Rekart, T.E.

    1993-09-01

    Poor removal of particulate corrosion products (especially iron) from condensate is one of the major problems in BWR systems. The presence of activated corrosion products creates ''hot spots'' and increases piping dose rates. Also, fuel efficiency is reduced and the risk of fuel failure is increased by the deposit of corrosion products on the fuel. Because of these concerns, current EPRI guidelines call for a maximum of 2 ppb of iron in the reactor feedwater with a level of 0.5 ppb being especially desirable. It has become clear that conventional deep bed resins are incapable of meeting these levels. While installation of prefilter systems is an option, it would be more economical for plants with naked deep beds to find an improved bead resin for use in existing systems. BWR condensate filtration technologies are being tested on a condensate side stream at Hope Creek Nuclear Generating Station. After two years of testing, hollow fiber filters (HFF) and fiber matrix filters (FMF), and low crosslink cation resin, all provide acceptable results. The results are presented for pressure drop, filtration efficiency, and water quality measurements. The costs are compared for backwashable non-precoat HFF and FMF. Results are also presented for full deep bed vessel tests of the low crosslink cation resin

  14. BWR type nuclear reactors

    International Nuclear Information System (INIS)

    Yamamoto, Toru.

    1987-01-01

    Purpose: To obtain reactor core characteristics with less changes in the excess reactivity due to fuel burnup even when the operation period varies. Constitution: In a BWR type reactor where fuel assemblies containing fuel rods incorporated with burnable poisons are arranged, the fuel assemblies are grouped into first fuel assemblies and second fuel assemblies. Then, the number of fuel rods incorporated with burnable poisons within the first fuel assemblies is made greater than that of the second fuel rods, while the concentration of the burnable poisons in the fuel rods incorporated with the burnable poisons in the first fuel assemblies is made lower than that of the fuel rods incorporated with the burnable poisons in the second fuel assemblies. In the BWR type reactor constituted in this way, the reactor core characteristics can be improved by changing the ratio between the first fuel assemblies and the second fuel assemblies charged to the reactor core, thereby decreasing the changes in the burnup of the excess reactivity. (Kamimura, M.)

  15. BWR zinc addition Sourcebook

    International Nuclear Information System (INIS)

    Garcia, Susan E.; Giannelli, Joseph F.; Jarvis, Alfred J.

    2014-01-01

    Boiling Water Reactors (BWRs) have been injecting zinc into the primary coolant via the reactor feedwater system for over 25 years for the purpose of controlling primary system radiation fields. The BWR zinc injection process has evolved since the initial application at the Hope Creek Nuclear Station in 1986. Key transitions were from the original natural zinc oxide (NZO) to depleted zinc oxide (DZO), and from active zinc injection of a powdered zinc oxide slurry (pumped systems) to passive injection systems (zinc pellet beds). Zinc addition has continued through various chemistry regimes changes, from normal water chemistry (NWC) to hydrogen water chemistry (HWC) and HWC with noble metals (NobleChem™) for mitigation of intergranular stress corrosion cracking (IGSCC) of reactor internals and primary system piping. While past reports published by the Electric Power Research Institute (EPRI) document specific industry experience related to these topics, the Zinc Sourcebook was prepared to consolidate all of the experience gained over the past 25 years. The Zinc Sourcebook will benefit experienced BWR Chemistry, Operations, Radiation Protection and Engineering personnel as well as new people entering the nuclear power industry. While all North American BWRs implement feedwater zinc injection, a number of other BWRs do not inject zinc. This Sourcebook will also be a valuable resource to plants considering the benefits of zinc addition process implementation, and to gain insights on industry experience related to zinc process control and best practices. This paper presents some of the highlights from the Sourcebook. (author)

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

    International Nuclear Information System (INIS)

    Haukeland, S.; Ivung, B.; Pedersen, T.

    1999-01-01

    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

  17. 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.

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

    International Nuclear Information System (INIS)

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

    2007-11-01

    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

  19. Methodology development for plutonium categorization and enhancement of proliferation resistance by P3 mechanism

    Energy Technology Data Exchange (ETDEWEB)

    Saito, M.; Kimura, Y.; Sagara, H.; Han, C. Y. [Tokyo Institute of Technology, Tokyo (Japan); Koyama, S. [Japan Atomic Energy Agency, Ibaraki (Japan)

    2012-03-15

    'Protected Plutonium Production (P3)' has been proposed to enhance the proliferation resistance of plutonium by the transmutation of Minor Actinides (MA). For example, adding the small amount of Minor Actinides such as {sup 237}Np or {sup 241}Am with large neutron capture cross-section to the uranium fuel to enhance the production of {sup 238}Pu, which has high spontaneous fission neutron rate do deteriorate the quality of the nuclear weapon manufacture and maintenance technologically difficult, is very effective for improving the isotopic barrier for the proliferation of plutonium. To demonstrate the P3 mechanism experimentally, U samples with 2, 5 and 10% {sup 237}Np doping were irradiated in Advanced Thermal Reactor (ATR) of INL. The fuel test samples were removed from the core at 100, 200 and 300 effective full power days (EFPD), and then post irradiation examination was completed at Chemical Lab. in Idaho National Laboratory(INL). The theoretical results of P3 mechanism predict the experimental ones quite well. The evaluation function, 'Attractiveness', was introduced as the ratio of function of Rossi-alpha to the 'Technical Difficulties for Fission Explosive Device Use. 'Rossi-alpha defined as the ratio of super-criticality to prompt neutron lifetime is the meaningful feature of the explosive yield. The Technical Difficulties for Fission Explosive Device Use can be expressed by the function of specific decay heat , spontaneous fission neutron rate and radiation of plutonium metal. Original methodology to evaluate Attractiveness of Plutonium has been improved by considering the effect of the compression of Plutonium isotope and also pre-detonation probability due to spontaneous fission neutron ate, which was applied for the categorization of the plutonium from the conventional reactors and the innovative reactors based on P3 mechanism. In the present paper, the fundamentals of P3 mechanism, the experimental demonstration of P3

  20. Methodology development for plutonium categorization and enhancement of proliferation resistance by P3 mechanism

    International Nuclear Information System (INIS)

    Saito, M.; Kimura, Y.; Sagara, H.; Han, C. Y.; Koyama, S.

    2012-01-01

    'Protected Plutonium Production (P3)' has been proposed to enhance the proliferation resistance of plutonium by the transmutation of Minor Actinides (MA). For example, adding the small amount of Minor Actinides such as 237 Np or 241 Am with large neutron capture cross-section to the uranium fuel to enhance the production of 238 Pu, which has high spontaneous fission neutron rate do deteriorate the quality of the nuclear weapon manufacture and maintenance technologically difficult, is very effective for improving the isotopic barrier for the proliferation of plutonium. To demonstrate the P3 mechanism experimentally, U samples with 2, 5 and 10% 237 Np doping were irradiated in Advanced Thermal Reactor (ATR) of INL. The fuel test samples were removed from the core at 100, 200 and 300 effective full power days (EFPD), and then post irradiation examination was completed at Chemical Lab. in Idaho National Laboratory(INL). The theoretical results of P3 mechanism predict the experimental ones quite well. The evaluation function, 'Attractiveness', was introduced as the ratio of function of Rossi-alpha to the 'Technical Difficulties for Fission Explosive Device Use. 'Rossi-alpha defined as the ratio of super-criticality to prompt neutron lifetime is the meaningful feature of the explosive yield. The Technical Difficulties for Fission Explosive Device Use can be expressed by the function of specific decay heat , spontaneous fission neutron rate and radiation of plutonium metal. Original methodology to evaluate Attractiveness of Plutonium has been improved by considering the effect of the compression of Plutonium isotope and also pre-detonation probability due to spontaneous fission neutron ate, which was applied for the categorization of the plutonium from the conventional reactors and the innovative reactors based on P3 mechanism. In the present paper, the fundamentals of P3 mechanism, the experimental demonstration of P3 mechanism in ATR of INL and the methodology

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

    International Nuclear Information System (INIS)

    Whitlock, J.; Bari, R.; Peterson, P.; Padoani, F.; Cojazzi, G.G.M.; Renda, G.; ); Cazalet, J.; Haas, E.; Hori, K.; Kawakubo, Y.; Chang, S.; Kim, H.; Kwon, E.-H.; Yoo, H.; Chebeskov, A.; Pshakin, G.; Pilat, J.F.; Therios, I.; Bertel, E.

    2015-01-01

    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)

  2. A Comparison of the Safety Analysis Process and the Generation IV Proliferation Resistance/Physical Protection Assessment Methodology

    International Nuclear Information System (INIS)

    T. A. Bjornard; M. D. Zentner

    2006-01-01

    The Generation IV International Forum (GIF) is a vehicle for the cooperative international development of future nuclear energy systems. The Generation IV program has established primary objectives in the areas of sustainability, economics, safety and reliability, and Proliferation Resistance and Physical Protection (PR and PP). In order to help meet the latter objective a program was launched in December 2002 to develop a rigorous means to assess nuclear energy systems with respect to PR and PP. The study of Physical Protection of a facility is a relatively well established methodology, but an approach to evaluate the Proliferation Resistance of a nuclear fuel cycle is not. This paper will examine the Proliferation Resistance (PR) evaluation methodology being developed by the PR group, which is largely a new approach and compare it to generally accepted nuclear facility safety evaluation methodologies. Safety evaluation methods have been the subjects of decades of development and use. Further, safety design and analysis is fairly broadly understood, as well as being the subject of federally mandated procedures and requirements. It is therefore extremely instructive to compare and contrast the proposed new PR evaluation methodology process with that used in safety analysis. By so doing, instructive and useful conclusions can be derived from the comparison that will help to strengthen the PR methodological approach as it is developed further. From the comparison made in this paper it is evident that there are very strong parallels between the two processes. Most importantly, it is clear that the proliferation resistance aspects of nuclear energy systems are best considered beginning at the very outset of the design process. Only in this way can the designer identify and cost effectively incorporate intrinsic features that might be difficult to implement at some later stage. Also, just like safety, the process to implement proliferation resistance should be a dynamic

  3. BWR fuel performance

    International Nuclear Information System (INIS)

    Baily, W.E.; Armijo, J.S.; Jacobson, J.; Proebstle, R.A.

    1979-01-01

    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. BWR type reactors

    International Nuclear Information System (INIS)

    Yano, Ryoichi; Sato, Takashi; Osaki, Masahiko; Hirayama, Fumio; Watabe, Atsushi.

    1980-01-01

    Purpose: To effectively eliminate radioactive substances released upon loss of coolant accidents in BWR type reactors. Constitution: A high pressure gas jetting device having a plurality of small aperture nozzles is provided above a spray nozzle, that is, at the top of a dry well. The jetting device is connected to a vacuum breaker provided in a pressure suppression chamber. Upon loss of coolant accident, coolants are sprayed from the spray nozzle and air or nitrogen is jetted from the gas jetting device as well. Then, the gases in the dry well are disturbed, whereby radioactive iodine at high concentration liable to be accumulated in the dry well is forced downwardly, dissolved in the spray water and eliminated. (Ikeda, J.)

  5. 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

  6. Nuclear data for innovative reactors having persistent resistance to nuclear proliferation

    International Nuclear Information System (INIS)

    Yoshida, Tadashi; Hagura, Naoto; Toshinsky, Vladimir; Saito, Masaki

    2005-01-01

    The Protected Plutonium Production (P 3 ) project has been initiated in order to realize the proliferation-resistant plutonium. The first stage of this project is based on the idea to protect plutonium by 238 Pu generated from 237 Np doped in the fresh uranium fuel. Along with the burnup going on, the doped 237 Np is transmuted via 238 Np to 238 Pu, which is the intense neutron source of spontaneous fission (2.6x10 3 n/g/s). These neutrons will deteriorate the quality of the plutonium as the nuclear explosive material. In addition, high alpha-decay heat of 238 Pu (560 W/kg) makes the processes of the explosive fabrication and its maintenance technologically difficult. In this background we surveyed the status of nuclear data relevant to the P 3 -concept reactor design. The reliability of the capture cross section data for 237 Np and 238 Pu are far from satisfactory and more accurate experiments and evaluations are required. As fissions in these nuclides do not always play any decisive role in the P 3 concept LWR, we can be fairly tolerable about the uncertainty of the fission cross section data, but further study is needed. (author)

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

    International Nuclear Information System (INIS)

    Farmer, M. T.; Hill, D. J.; Sienicki, J. J.; Spencer, B. W.; Wade, D. C.

    1999-01-01

    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

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

    International Nuclear Information System (INIS)

    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

  9. 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.

  10. Acute insulin resistance stimulates and insulin sensitization attenuates vascular smooth muscle cell migration and proliferation.

    Science.gov (United States)

    Cersosimo, Eugenio; Xu, Xiaojing; Upala, Sikarin; Triplitt, Curtis; Musi, Nicolas

    2014-08-01

    Differential activation/deactivation of insulin signaling, PI-3K and MAP-K pathways by high glucose and palmitate, with/out the insulin sensitizer pioglitazone (PIO), have been previously shown in vascular smooth muscle cells (VSMCs). To determine the biological impact of these molecular changes, we examined VSMC migration and proliferation ("M"&"P") patterns in similar conditions. VSMCs from healthy human coronary arteries were incubated in growth medium and "M"&"P" were analyzed after exposure to high glucose (25 mmol/L) ± palmitate (200 μmol/L) and ± PIO (8 μmol/L) for 5 h. "M"&"P" were assessed by: (1) polycarbonate membrane barrier with chemo-attractants and extended cell protrusions quantified by optical density (OD595 nm); (2) % change in radius area (2D Assay) using inverted microscopy images; and (3) cell viability assay expressed as cell absorbance (ABS) in media. "M" in 25 mmol/L glucose media increased by ~25% from baseline and % change in radius area rose from ~20% to ~30%. The addition of PIO was accompanied by a significant decrease in "M" from 0.25 ± 0.02 to 0.19 ± 0.02; a comparable decline from 0.25 ± 0.02 to 0.18 ± 0.02 was also seen with 25 mmol/L of glucose +200 μmol/L of palmitate. When PIO was coincubated with high glucose plus palmitate there was a 50% reduction in % change in radius. A ~10% increase in ABS, reflecting augmented "P" in media with 25 mmol/L glucose versus control was documented. The addition of PIO reduced ABS from 0.208 ± 0.03 to 0.183 ± 0.06. Both high glucose and palmitate showed ABS of ~0.140 ± 0.02, which decreased with PIO to ~0.120 ± 0.02, indicating "P" was reduced. These results confirm that high glucose and palmitate stimulate VSMCs migration and proliferation in vitro, which is attenuated by coincubation with the insulin sensitizer PIO. Although, we cannot ascertain whether these functional changes are coincident with the activation/deactivation of signal molecules, our findings are consistent with the

  11. Intrinsically secure fast reactors for long-lived waste free and proliferation resistant nuclear power

    International Nuclear Information System (INIS)

    Slessarev, Igor

    2008-01-01

    This paper provides description of a nuclear reactor concept aimed towards a radical safety enhancement, an increased proliferation resistance, as well as a realisation of a 'long-lived waste free' NP development. It emphasizes the achievement of considerable reduction ('by design') of residual actinides in the waste streams and of the most hazardous long-lived fission products. It allows to implement only small volume of repositories for the radioactive waste (mostly fission products) and to postpone the technically arduous problems of a large scale disposal of the long-lived wastes until the next millennium, i.e. up to the exhaustion of the fertile natural resources and/or the emergence of more effective technologies of nuclide separation/transmutation. A thorough incineration/transmutation of the wastes (residual actinides in the mixture with lanthanides as well as of the most hazardous fission products) under reactor neutron flux is proposed for their mass reduction. A gradual growth of NP park is necessary for increasing the NP park capacity for waste irradiation. This 'constraint' is not really limiting because it coincides with the permanently growing demands in energy production. The potential of long-lived waste reduction depends on the total fertile fuel resources and on NP growth rate. It was shown that the accumulated actinide long-lived radioactive masses will be reduced significantly: by factor in the range of 10 4 -10 8 in magnitude (compared with LWR once-through cycle) and by 10 2 -10 6 (compared with the ordinary fast reactor park). Thus, the total long-lived waste toxicity pollution source might be comparable with the 'burnt away' toxicity of the natural fertile feed stream. This is quite realistic taking into account the large fertile fuel (U/Th) world-wide resources which provide the NP growth for a sustained time. Along side with the radical intrinsic safety improvement, a further enhancement of core physics ('neutronics') is one of the

  12. BWR type reactors

    International Nuclear Information System (INIS)

    Watanabe, Shoichi

    1986-01-01

    Purpose: To enable to remove water not by way of mechanical operation in a reactor core and improve the fuel economy in BWR type reactors. Constitution: A hollow water removing rod of a cross-like profile made of material having a smaller neutron absorption cross section than the moderator is disposed to the water gap for each of unit structures composed of four fuel assemblies, and water is charged and discharged to and from the water removing rod. Water is removed from the water removing rod to decrease the moderators in the water gap to carry out neutron spectrum shift operation from the initial to the medium stage of reactor core cycles. At the final stage of the cycle, airs in the water removing rod are extracted and the moderator is introduced. The moderator is filled and the criticality is maintained with the accumulated nuclear fission materials. The neutron spectrum shift operation can be attained by eliminating hydrothermodynamic instability and using a water removing rod of a simple structure. (Horiuchi, T.)

  13. BWR emergency procedure guidelines

    International Nuclear Information System (INIS)

    Post, J.S.; Karner, E.F.; Stratman, R.A.

    1984-01-01

    This chapter describes plans for dealing with reactor accidents developed by the Boiling Water Reactor (BWR) Owners' Group in response to post-Three Mile Island US NRC requirements. The devised Emergency Procedure Guidelines (EPGs), applicable to all BWRs, are symptom-based rather than event-based. According to the EPGs, the operator does not need to identify what event is occurring in the plant in order to decide what action to take, but need only observe the symptoms (values and trends of key control parameters) which exist and take appropriate action to control these symptoms. The original objective was to provide reactor operator guidance in responding to a small break loss-of-coolant accident (LOCA), but subsequent revisions have included other types of reactor accidents. Topics considered include the reactor pressure vessel (RPV) control guideline, the primary containment control guideline, the secondary containment control guideline, the radioactivity release control guideline, multiple failures vs. the design basis, safe limits vs. technical specifications, the technical status, licensing, and implementation. The EPGs are based upon maintaining both adequate core cooling and primary containment integrity

  14. BWR type reactor core

    International Nuclear Information System (INIS)

    Tatemichi, Shin-ichiro.

    1981-01-01

    Purpose: To eliminate the variation in the power distribution of a BWR type reactor core in the axial direction even if the flow rate is increased or decreased by providing a difference in the void coefficient between the upper part and the lower parts of the reactor core, and increasing the void coefficient at the lower part of the reactor core. Constitution: The void coefficient of the lower region from the center to the lower part along the axial direction of a nuclear fuel assembly is increased to decrease the dependence on the flow rate of the axial power distribution of the nuclear fuel assembly. That is, a water/fuel ratio is varied, the water in non-boiled region is increased or the neutron spectrum is varied so as to vary the void coefficient. In order to exemplify it, the rate of the internal pellets of the fuel rod of the nuclear fuel assembly or the shape of the channel box is varied. Accordingly, the power does not considerably vary even if the flow rate is altered since the power is varied in the power operation. (Yoshihara, H.)

  15. BWR type reactor

    International Nuclear Information System (INIS)

    Okano, Shigeru.

    1992-01-01

    In a BWR type reactor, control rod drives are disposed in the upper portion of a reactor pressure vessel, and a control rod guide tube is disposed in adjacent with a gas/liquid separator at a same height, as well as a steam separator is disposed in the control rod guide tube. The length of a connection rod can be shortened by so much as the control rod guide tube and the gas/liquid separator overlapping with each other. Since the control rod guide tube and the gas/liquid separator are at the same height, the number of the gas/liquid separators to be disposed is decreased and, accordingly, even if the steam separation performance by the gas/liquid separator is lowered, it can be compensated by the steam separator of the control rod guide tube. In view of the above, since the direction of emergent insertion of the control rod is not against gravitational force but it is downward direction utilizing the gravitational force, reliability for the emergent insertion of the control rod can be further improved. Further, the length of the connection rod can be minimized, thereby enabling to lower the height of the reactor pressure vessel. The construction cost for the nuclear power plant can be reduced. (N.H.)

  16. 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.

  17. High Fidelity BWR Fuel Simulations

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-01

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

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

    Science.gov (United States)

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

    2016-05-10

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

  19. BWR Services maintenance training program

    International Nuclear Information System (INIS)

    Cox, J.H.; Chittenden, W.F.

    1979-01-01

    BWR Services has implemented a five-phase program to increase plant availability and capacity factor in operating BWR's. One phase of this program is establishing a maintenance training program on NSSS equipment; the scope encompasses maintenance on both mechanical equipment and electrical control and instrumentation equipment. The program utilizes actual product line equipment for practical Hands-on training. A total of 23 formal courses will be in place by the end of 1979. The General Electric Company is making a multimillion dollar investment in facilities to support this training. These facilities are described

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

    NARCIS (Netherlands)

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

    2002-01-01

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

  1. Concept of the development of new type reactor and other nuclear technology in which nuclear-proliferation resistivity is taken into consideration

    International Nuclear Information System (INIS)

    Chang, Yoon I.

    2000-01-01

    As we start a new century in a new millennium, it is timely to look back at the last 50 years of nuclear energy development, and take a pause here to formulate the directions we need to take in the nuclear energy development for the next 50 years and beyond. The proliferation resistivity should be given a priority in the development of next-generation advanced reactor concepts, and there are three important principles to be considered in such efforts. The proliferation resistivity should be considered as a part of broader objectives, rather than a design goal by itself; the proliferation resistivity should be considered within the context of global institutional framework; and any technical proliferation resistivity features designed into the reactor system should be made intrinsic in nature. (author)

  2. Alternative nuclear fuel cycle arrangements for proliferation resistance: an overview of regulatory factors

    International Nuclear Information System (INIS)

    O'Brien, J.N.

    1982-08-01

    President Carter proposed the International Fuel Cycle Evaluation to analyze various alternative fuel cycles which could minimize the risk of nuclear weapons proliferation. DOE also initiated the Non-Proliferation Alternative Systems Assessment Program. In response to GAO contentions that NRC was not sufficiently involved in these two assessments, a NRC study was initiated, with emphasis on legal and institutional factors. Objectives were to examine multinational fuel cycle facilities, potential effects on the US/IAEA agreement, development of an algorithm for ranking potential fuel cycles, and potential licensing of candidate fuel cycles. This anthology represents the products of this study which has been conducted between 1979 and 1981

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

    International Nuclear Information System (INIS)

    Blanquart, Christophe; Karouri, Salah-Eddine; Issad, Tarik

    2009-01-01

    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.

  4. Proliferation Resistance and Safeguardability Assessment of a SFR Metal Fuel Manufacturing Facility (SFMF) using the INPRO Methodology

    Energy Technology Data Exchange (ETDEWEB)

    Chang, H. L.; Ko, W. I.; Park, S. H.; Kim, H. D.; Park, G. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    To illustrate the proposed Prosta process, to demonstrate its usefulness, and to provide input to a revision of the INPRO manual in the area of proliferation resistance, a case study has been carried out with a conceptually designed sodium cooled fast reactor (SFR) metal fuel manufacturing facility (SFMF), representing novel technology still in the conceptual design phase. A coarse acquisition path analysis has been carried out of the SFMF to demonstrate the assessment process with identified different target materials. The case study demonstrates the usefulness of the proposed PROSA PR assessment process and the interrelationship of the PR assessment with the safeguards-by-design process, identifying potential R and D needs. The PROSA process has been applied to a conceptually designed SFMF, representing novel technology that is still in the conceptual design phase at KAERI. The case study demonstrated that the proposed PROSA process is simpler and easier to perform than the original INPRO methodology and can be applied from the early stage of design showing the relationship of PR assessment to the safeguard-by-design process. New evaluation questionnaire for UR1 is more logical and comprehensive, and provides the legal basis enabling the IAEA to achieve its safeguards objectives including the detection of undeclared nuclear materials and activities. NES information catalogue replacing UR2 was a useful modification and supports safeguardability assessment at the NES and facility level. The proposed PROSA process is also capable to identify strengths and weaknesses of a system in the area of proliferation resistance in a generally understandable form, including R and D gaps that need to be filled in order to meet the criteria for proliferation resistance of a nuclear energy system.

  5. 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.

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

    International Nuclear Information System (INIS)

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

    2014-01-01

    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

  7. EMMPRIN promotes angiogenesis, proliferation, invasion and resistance to sunitinib in renal cell carcinoma, and its level predicts patient outcome.

    Science.gov (United States)

    Sato, Mototaka; Nakai, Yasutomo; Nakata, Wataru; Yoshida, Takahiro; Hatano, Koji; Kawashima, Atsunari; Fujita, Kazutoshi; Uemura, Motohide; Takayama, Hitoshi; Nonomura, Norio

    2013-01-01

    Extracellular matrix metalloproteinase inducer (EMMPRIN) has been reported to play crucial roles, including in angiogenesis, in several carcinomas. However, the correlation between EMMPRIN levels and angiogenesis expression profile has not been reported, and the role of EMMPRIN in renal cell carcinoma (RCC) is unclear. In the present study, we evaluated the association of EMMPRIN with angiogenesis, its value in prognosis, and its roles in RCC. EMMPRIN expression was examined in 50 RCC patients treated with radical nephrectomy. Angiogenesis, proliferation, and invasion activity were evaluated using EMMPRIN knockdown RCC cell lines. The size of EMMPRIN-overexpressing xenografts was measured and the degree of angiogenesis was quantified. EMMPRIN expression was evaluated in RCC patients who received sunitinib therapy and in sunitinib-resistant cells. Further, the relation between EMMPRIN expression and sensitivity to sunitinib was examined. EMMPRIN score was significantly associated with clinicopathological parameters in RCC patients, as well as being significantly correlated with microvessel area (MVA) in immature vessels and with prognosis. Down-regulation of EMMPRIN by siRNA led to decreased VEGF and bFGF expression, cell proliferation, and invasive potential. EMMPRIN over-expressing xenografts showed accelerated growth and MVA of immature vessels. EMMPRIN expression was significantly increased in patients who received sunitinib therapy as well as in sunitinib-resistant 786-O cells (786-suni). EMMPRIN-overexpressing RCC cells were resistant to sunitinib. Our findings indicate that high expression of EMMPRIN in RCC plays important roles in tumor progression and sunitinib resistance. Therefore, EMMPRIN could be a novel target for the treatment of RCC.

  8. EMMPRIN promotes angiogenesis, proliferation, invasion and resistance to sunitinib in renal cell carcinoma, and its level predicts patient outcome.

    Directory of Open Access Journals (Sweden)

    Mototaka Sato

    Full Text Available Extracellular matrix metalloproteinase inducer (EMMPRIN has been reported to play crucial roles, including in angiogenesis, in several carcinomas. However, the correlation between EMMPRIN levels and angiogenesis expression profile has not been reported, and the role of EMMPRIN in renal cell carcinoma (RCC is unclear. In the present study, we evaluated the association of EMMPRIN with angiogenesis, its value in prognosis, and its roles in RCC.EMMPRIN expression was examined in 50 RCC patients treated with radical nephrectomy. Angiogenesis, proliferation, and invasion activity were evaluated using EMMPRIN knockdown RCC cell lines. The size of EMMPRIN-overexpressing xenografts was measured and the degree of angiogenesis was quantified. EMMPRIN expression was evaluated in RCC patients who received sunitinib therapy and in sunitinib-resistant cells. Further, the relation between EMMPRIN expression and sensitivity to sunitinib was examined.EMMPRIN score was significantly associated with clinicopathological parameters in RCC patients, as well as being significantly correlated with microvessel area (MVA in immature vessels and with prognosis. Down-regulation of EMMPRIN by siRNA led to decreased VEGF and bFGF expression, cell proliferation, and invasive potential. EMMPRIN over-expressing xenografts showed accelerated growth and MVA of immature vessels. EMMPRIN expression was significantly increased in patients who received sunitinib therapy as well as in sunitinib-resistant 786-O cells (786-suni. EMMPRIN-overexpressing RCC cells were resistant to sunitinib.Our findings indicate that high expression of EMMPRIN in RCC plays important roles in tumor progression and sunitinib resistance. Therefore, EMMPRIN could be a novel target for the treatment of RCC.

  9. Improvement for BWR operator training

    International Nuclear Information System (INIS)

    Tsuchiya, Toshio; Masuda, Hisao; Isono, Tomoyuki; Noji, Kunio; Togo, Toshiki

    1989-01-01

    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)

  10. Synergistic failure of BWR internals

    International Nuclear Information System (INIS)

    Ware, A. G.; Chang, T.Y.

    1999-01-01

    Boiling Water Reactor (BWR) core shrouds and other reactor internals important to safety are experiencing intergranular stress corrosion cracking (IGSCC). The United States Nuclear Regulatory Commission has followed the problem, and as part of its investigations, contracted with the Idaho National Engineering and Environmental Laboratory to conduct a risk assessment. The overall project objective is to assess the potential consequences and risks associated with the failure of IGSCC-susceptible BWR vessel internals, with specific consideration given to potential cascading and common mode effects. An initial phase has been completed in which background material was gathered and evaluated, and potential accident sequences were identified. A second phase is underway to perform a simplified, quantitative probabilistic risk assessment on a representative high-power BWR/4. Results of the initial study conducted on the jet pumps show that any cascading failures would not result in a significant increase in the core damage frequency. The methodology is currently being extended to other major reactor internals components

  11. BWR control blade replacement strategies

    Energy Technology Data Exchange (ETDEWEB)

    Kennard, M W [Stoller Nuclear Fuel, NAC International, Pleasantville, NY (United States); Harbottle, J E [Stoller Nuclear Fuel, NAC International, Thornbury, Bristol (United Kingdom)

    2000-02-01

    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 B{sub 4}C 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)

  12. BWR control blade replacement strategies

    International Nuclear Information System (INIS)

    Kennard, M.W.; Harbottle, J.E.

    2000-01-01

    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 B 4 C 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)

  13. Development of next BWR plant

    International Nuclear Information System (INIS)

    Moriya, Kumiaki; Tanikawa, Naoshi; Kinoshita, Shoichiro; Utena, Shunsuke

    1995-01-01

    It is expected that BWR power generation will be main nuclear power generation for long period hereafter, and in the ABWRs being constructed at present, the safety, reliability, operation performance, economical efficiency and so on are further heightend as compared with conventional BWRs. On the other hand, in order to cope with future social change, the move to develop the next reactor type following ABWRs was begun already by the cooperation of electirc power companies and plant manufacturers. Hitachi Ltd. has advanced eagerly the development of new light water reactors. Also the objective of BWR power generation hereafter is to heighten the safety, reliability, operation performance and economical efficiency, and the development has been advanced, aiming at bearing the main roles of nuclear power generation. At present, ABWRs are under construction as No. 6 and 7 plants in Kashiwazaki Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc. In order to let ABWRs take root, the further improvement of economy by the standardization, the rationalization by revising the specification and the improvement of machinery and equipment is necessary. As the needs of the development of next generation BWRs, the increase of power output, the heightening of safety and economical efficiency are discussed. The concept of the next generation BWR plant aiming at the start of operation around 2010 is shown. (K.I.)

  14. Development of next BWR plant

    Energy Technology Data Exchange (ETDEWEB)

    Moriya, Kumiaki; Tanikawa, Naoshi; Kinoshita, Shoichiro; Utena, Shunsuke [Hitachi Ltd., Ibaraki (Japan). Hitachi Works

    1995-04-01

    It is expected that BWR power generation will be main nuclear power generation for long period hereafter, and in the ABWRs being constructed at present, the safety, reliability, operation performance, economical efficiency and so on are further heightend as compared with conventional BWRs. On the other hand, in order to cope with future social change, the move to develop the next reactor type following ABWRs was begun already by the cooperation of electirc power companies and plant manufacturers. Hitachi Ltd. has advanced eagerly the development of new light water reactors. Also the objective of BWR power generation hereafter is to heighten the safety, reliability, operation performance and economical efficiency, and the development has been advanced, aiming at bearing the main roles of nuclear power generation. At present, ABWRs are under construction as No. 6 and 7 plants in Kashiwazaki Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc. In order to let ABWRs take root, the further improvement of economy by the standardization, the rationalization by revising the specification and the improvement of machinery and equipment is necessary. As the needs of the development of next generation BWRs, the increase of power output, the heightening of safety and economical efficiency are discussed. The concept of the next generation BWR plant aiming at the start of operation around 2010 is shown. (K.I.).

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

    International Nuclear Information System (INIS)

    2015-01-01

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-23

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

  17. BWR Steam Dryer Alternating Stress Assessment Procedures

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-01

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

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

    DEFF Research Database (Denmark)

    Mackey, Abigail; Esmarck, B; Kadi, F

    2007-01-01

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

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

    International Nuclear Information System (INIS)

    Wang, Jilu; Mao, Daqing; Mu, Quanhua; Luo, Yi

    2015-01-01

    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 3 to (3.68 ± 0.27) × 10 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 7 to (4.30 ± 0.81) × 10 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 7–308 times higher than

  20. 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

  1. Power oscillations in BWR reactors

    International Nuclear Information System (INIS)

    Espinosa P, G.

    2002-01-01

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

  2. Variations of uranium and plutonium coprocessing as proliferation-resistant alternatives to the classical purex process

    International Nuclear Information System (INIS)

    Buckham, J.A.; Sumner, W.B.

    1979-08-01

    Evaluation of these alternatives for processing LWR fuel has led to the following conclusions: (1) None of the alternaives provide a pure, technical solution which completely eliminates the potential for proliferation of nuclear weapons by utilizing plutonium from the light water reactors. (2) The heat spike alternative appears feasible and provides the most effective method of rendering the LWR plutonim unattractive for weapons use. (3) The low-DF process alternate would require demonstration to: (a) determine the reliability of the in-cell recycle streams which are used to prevent reversion of the process for purification of plutonium, and (b) verify the fission product decontamination factors. (4) The alternates evaluated have no significant impacts on the design of waste treatment facilities, although the required capacities of high-level solid waste processing and high-level liquid waste storage can be significantly altered. (5) The impact of these alternate processes on fuel fabrication and other aspects of the fuel cycle requires additional evaluation

  3. Advanced Construction of Compact Containment BWR

    International Nuclear Information System (INIS)

    Takahashi, M.; Maruyama, T.; Mori, H.; Hoshino, K.; Hijioka, Y.; Heki, H.; Nakamaru, M.; Hoshi, T.

    2006-01-01

    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

  4. Development status of compact containment BWR

    International Nuclear Information System (INIS)

    Heki, H.; Nakamaru, M.; Mori, H.; Sekiguchi, K.; Kuroki, M.; Arai, K.; Hida, T.

    2005-01-01

    In Japan, increase of nuclear plant unit capacity has been promoted to take advantage of economies of scale while further enhancing safety and reliability. As a result, more than 50 units of nuclear power plants are playing important role in electric power generation. However, the factors, such as stagnant growth in the recent electricity demand, limitation in electricity grid capacity and limited in initial investment avoiding risk, will not be in favor of large plant outputs. The reactor concept considered in this paper has a small power output, a compact containment and a simplified BWR configuration with comprehensive safety features. The Compact Containment Boiling Water Reactor (CCR), which 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 ECCS 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

  5. Proliferation resistance of the fuel cycle for the Integral Fast Reactor

    International Nuclear Information System (INIS)

    Burris, L.

    1993-01-01

    Argonne National Laboratory has developed an electrorefining pyrochemical process for recovery and recycle of metal fuel discharged from the Integral Fast Reactor (FR). This inherently low decontamination process has an overall decontamination factor of only about 100 for the plutonium metal product. As a result, all of the fuel cycle operations must be conducted in heavily shielded cells containing a high-purity argon atmosphere. The FR fuel cycle possesses high resistance to clandestine diversion or overt, state- supported removal of plutonium for nuclear weapons production because of two main factors: the highly radioactive product, which is also contaminated with heat- and neutron-producing isotopes of plutonium and other actinide elements, and the difficulty of removing material from the FR facility through the limited number of cell transfer locks without detection

  6. GP88 (PC-Cell Derived Growth Factor, progranulin stimulates proliferation and confers letrozole resistance to aromatase overexpressing breast cancer cells

    Directory of Open Access Journals (Sweden)

    Sabnis Gauri

    2011-06-01

    Full Text Available Abstract Background Aromatase inhibitors (AI that inhibit breast cancer cell growth by blocking estrogen synthesis have become the treatment of choice for post-menopausal women with estrogen receptor positive (ER+ breast cancer. However, some patients display de novo or acquired resistance to AI. Interactions between estrogen and growth factor signaling pathways have been identified in estrogen-responsive cells as one possible reason for acquisition of resistance. Our laboratory has characterized an autocrine growth factor overexpressed in invasive ductal carcinoma named PC-Cell Derived Growth Factor (GP88, also known as progranulin. In the present study, we investigated the role GP88 on the acquisition of resistance to letrozole in ER+ breast cancer cells Methods We used two aromatase overexpressing human breast cancer cell lines MCF-7-CA cells and AC1 cells and their letrozole resistant counterparts as study models. Effect of stimulating or inhibiting GP88 expression on proliferation, anchorage-independent growth, survival and letrozole responsiveness was examined. Results GP88 induced cell proliferation and conferred letrozole resistance in a time- and dose-dependent fashion. Conversely, naturally letrozole resistant breast cancer cells displayed a 10-fold increase in GP88 expression when compared to letrozole sensitive cells. GP88 overexpression, or exogenous addition blocked the inhibitory effect of letrozole on proliferation, and stimulated survival and soft agar colony formation. In letrozole resistant cells, silencing GP88 by siRNA inhibited cell proliferation and restored their sensitivity to letrozole. Conclusion Our findings provide information on the role of an alternate growth and survival factor on the acquisition of aromatase inhibitor resistance in ER+ breast cancer.

  7. BWR water chemistry impurity studies

    International Nuclear Information System (INIS)

    Ljungberg, L.G.; Korhonen, S.; Renstroem, K.; Hofling, C.G.; Rebensdorff, B.

    1990-03-01

    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

  8. The Proliferation Resistance of a Nuclear Fuel Cycle Using Fuel Recovered from the Electrolytic Reduction of Pressurized Water Reactor Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Jung Min; Cochran, Thomas; Mckinzie, Matthew [NRDC, Washington, (United States)

    2016-05-15

    At some points in the fuel cycle, a level of intrinsic or technical proliferation-resistance can be provided by radiation barriers that surround weapons-usable materials. In this report we examine some aspects of intrinsic proliferation resistance of a fuel cycle for a fast neutron reactor that uses fuel recovered from the electrolytic reduction process of pressurized water reactor spent fuel, followed by a melt-refining process. This fuel cycle, proposed by a nuclear engineer at the Korea Advanced Institute of Science and Technology (KAIST), is being examined with respect to its potential merits of higher fuel utilization, lower production of radioactive byproducts, and better economics relative to a pyroprocesing-based fuel cycle. With respect to intrinsic proliferation resistance, however, we show that since europium is separated out during the electrolytic reduction process, this fuel cycle has little merit beyond that of a pyroprocessing-based fuel cycle because of the lower radiation barrier of its recovered materials containing weapons-usable actinides. Unless europium is not separated following voloxidation, the proposed KAIST fuel cycle is not intrinsically proliferation resistant and in this regard does not represent a significant improvement over pyroprocessing. We suggest further modification of the proposed KAIST fuel cycle, namely, omitting electrolytic reduction and melt reduction, and producing the fast reactor fuel directly following voloxidation.

  9. BWR core melt progression phenomena: Experimental analyses

    International Nuclear Information System (INIS)

    Ott, L.J.

    1992-01-01

    In the BWR Core Melt in Progression Phenomena Program, experimental results concerning severe fuel damage and core melt progression in BWR core geometry are used to evaluate existing models of the governing phenomena. These include control blade eutectic liquefaction and the subsequent relocation and attack on the channel box structure; oxidation heating and hydrogen generation; Zircaloy melting and relocation; and the continuing oxidation of zirconium with metallic blockage formation. Integral data have been obtained from the BWR DF-4 experiment in the ACRR and from BWR tests in the German CORA exreactor fuel-damage test facility. Additional integral data will be obtained from new CORA BWR test, the full-length FLHT-6 BWR test in the NRU test reactor, and the new program of exreactor experiments at Sandia National Laboratories (SNL) on metallic melt relocation and blockage formation. an essential part of this activity is interpretation and use of the results of the BWR tests. The Oak Ridge National Laboratory (ORNL) has developed experiment-specific models for analysis of the BWR experiments; to date, these models have permitted far more precise analyses of the conditions in these experiments than has previously been available. These analyses have provided a basis for more accurate interpretation of the phenomena that the experiments are intended to investigate. The results of posttest analyses of BWR experiments are discussed and significant findings from these analyses are explained. The ORNL control blade/canister models with materials interaction, relocation and blockage models are currently being implemented in SCDAP/RELAP5 as an optional structural component

  10. SCORPIO-BWR: status and future plans

    International Nuclear Information System (INIS)

    Porsmyr, Jan; Bodal, Terje; Beere, William H.

    2004-01-01

    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

  11. SCORPIO-BWR: status and future plans

    Energy Technology Data Exchange (ETDEWEB)

    Porsmyr, Jan; Bodal, Terje; Beere, William H. (and others)

    2004-07-01

    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

  12. Advances in BWR water chemistry

    International Nuclear Information System (INIS)

    Garcia, Susan E.; Giannelli, Joseph F.; Jarvis, Mary L.

    2012-09-01

    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

  13. Repair and preventive maintenance technology for BWR reactor internals and piping

    International Nuclear Information System (INIS)

    Ootsubo, Tooru; Itou, Takashi; Sakashita, Akihiro

    2009-01-01

    Stress corrosion cracking of welding portion has found in many domestic and foreign BWR reactor internals and Primary Loop Recirculation piping. Also, repair and preventive maintenance technologies for SCC has been developed and/or adopted to BWRs in recent years. This paper introduces the sample of these technologies, such as seal-welding for SCC on BWR reactor internals, preventive maintenance technology for PLR piping such as Corrosion Resistant Cladding, Internal Polishing and Induction Heating Stress Improvement. These technologies are introduced on 'E-Journal of Advanced Maintenance', which is an international journal on a exclusive website of Japan Society of Maintenology. (author)

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

    DEFF Research Database (Denmark)

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

    2017-01-01

    The objective of this study was to evaluate bone resistance after in ovo administration of copper nanoparticles (NanoCu) and to determine the number of cells positive for proliferating cell nuclear antigen (PCNA) in the femoral bones of broiler chickens (n = 12 per group). The study demonstrated...... that femoral bones from the NanoCu group were characterised by a higher weight and volume and by significantly greater resistance to fractures compared to the Control group. NanoCu promoted the proliferation of PCNA-positive cells in the long bones of chickens. A significantly higher number of PCNA......-positive cells in the bones of birds in the NanoCu group compared with the Control group (137 and 122, respectively) indicate a stimulatory effect during embryogenesis. Considering the improvement in bone resistance to fractures and the effect of NanoCu on the number of PCNA-positive cells in femoral bones, Nano...

  15. 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).

  16. BWR type nuclear power plant

    International Nuclear Information System (INIS)

    Matsumoto, Kosuke.

    1991-01-01

    In a BWR type nuclear power plant in which reactor water in a reactor pressure vessel can be drained to a waste processing system by way of reactor recycling pipeways and remaining heat removal system pipeways, a pressurized air supply device is disposed for supplying air for pressurizing reactor water to the inside of the reactor pressure vessel by way of an upper head. With such a constitution, since the pressurized air sent from the pressurized air supply device above the reactor pressure vessel for the reactor water discharging pressure upon draining, the water draining pressure is increased compared with a conventional case and, accordingly, the amount of drained water is not reduced even in the latter half of draining. Accordingly, the draining efficiency can be improved and only a relatively short period of time is required till the completion of the draining, which can improve safety and save labors. (T.M.)

  17. 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.

  18. 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)

    Ahn, Hee-Jin; Kim, Gwangil; Park, Kyung-Soon

    2013-01-01

    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

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

    International Nuclear Information System (INIS)

    Yang, Myung Seung; Ahn, D. H.; Ko, W. I.

    2007-03-01

    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

  20. The different expression of TRPM7 and MagT1 impacts on the proliferation of colon carcinoma cells sensitive or resistant to doxorubicin.

    Science.gov (United States)

    Cazzaniga, Alessandra; Moscheni, Claudia; Trapani, Valentina; Wolf, Federica I; Farruggia, Giovanna; Sargenti, Azzurra; Iotti, Stefano; Maier, Jeanette A M; Castiglioni, Sara

    2017-01-17

    The processes leading to anticancer drug resistance are not completely unraveled. To get insights into the underlying mechanisms, we compared colon carcinoma cells sensitive to doxorubicin with their resistant counterpart. We found that resistant cells are growth retarded, and show staminal and ultrastructural features profoundly different from sensitive cells. The resistant phenotype is accompanied by the upregulation of the magnesium transporter MagT1 and the downregulation of the ion channel kinase TRPM7. We demonstrate that the different amounts of TRPM7 and MagT1 account for the different proliferation rate of sensitive and resistant colon carcinoma cells. It remains to be verified whether they are also involved in the control of other "staminal" traits.

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

    Science.gov (United States)

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

    2013-08-01

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

  2. Simplified compact containment BWR plant

    International Nuclear Information System (INIS)

    Heki, H.; Nakamaru, M.; Tsutagawa, M.; Hiraiwa, K.; Arai, K.; Hida, T.

    2004-01-01

    The reactor concept considered in this paper has a small power output, a compact containment and a simplified BWR configuration with comprehensive safety features. The Compact Containment Boiling Water Reactor (CCR), which is being developed with matured BWR technologies together with innovative systems/components, is expected to prove attractive in the world energy markets due to its flexibility in regard to both energy demands and site conditions, its high potential for reducing investment risk and its safety features facilitating public acceptance. The flexibility is achieved by CCR's small power output of 300 MWe class and capability of long operating cycle (refueling intervals). CCR is expected to be attractive from view point of investment due to its simplification/innovation in design such as natural circulation core cooling with the bottom located short core, internal upper entry control rod drives (CRDs) with ring-type dryers and simplified ECCS system with high pressure containment concept. The natural circulation core eliminates recirculation pumps and the maintenance of such pumps. The internal upper entry CRDs reduce the height of the reactor vessel (RPV) and consequently reduce the height of the primary containment vessel (PCV). The safety features mainly consist of large water inventory above the core without large penetration below the top of the core, passive cooling system by isolation condenser (IC), passive auto catalytic recombiner and in-vessel retention (IVR) capability. The large inventory increases the system response time in the case of design-base accidents, including loss of coolant accidents. The IC suppresses PCV pressure by steam condensation without any AC power. The recombiner decreases hydrogen concentration in the PCV in the case of a severe accident. Cooling the molten core inside the RPV if the core should be damaged by loss of core coolability could attain the IVR. The feasibility of CCR safety system has been confirmed by LOCA

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

    International Nuclear Information System (INIS)

    Jia, Yanhan; Zhang, Yan; Qiao, Chunxia; Liu, Guijun; Zhao, Qing; Zhou, Tingting; Chen, Guojiang; Li, Yali; Feng, Jiannan; Li, Yan; Zhang, Qiuping; Peng, Hui

    2013-01-01

    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

  4. 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.

  5. Improvement for BWR operator training

    International Nuclear Information System (INIS)

    Kurisu, Takanori; Takahashi, Yoshitaka; Harada, Mitsuhiro; Takahashi, Iwao.

    1988-01-01

    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.)

  6. Fundamentals of boiling water reactor (BWR)

    International Nuclear Information System (INIS)

    Bozzola, S.

    1982-01-01

    These lectures on fundamentals of BWR reactor physics are a synthesis of known and established concepts. These lectures are intended to be a comprehensive (even though descriptive in nature) presentation, which would give the basis for a fair understanding of power operation, fuel cycle and safety aspects of the boiling water reactor. The fundamentals of BWR reactor physics are oriented to design and operation. In the first lecture general description of BWR is presented, with emphasis on the reactor physics aspects. A survey of methods applied in fuel and core design and operation is presented in the second lecture in order to indicate the main features of the calculational tools. The third and fourth lectures are devoted to review of BWR design bases, reactivity requirements, reactivity and power control, fuel loading patterns. Moreover, operating limits are reviewed, as the actual limits during power operation and constraints for reactor physics analyses (design and operation). The basic elements of core management are also presented. The constraints on control rod movements during the achieving of criticality and low power operation are illustrated in the fifth lecture. Some considerations on plant transient analyses are also presented in the fifth lecture, in order to show the impact between core and fuel performance and plant/system performance. The last (sixth) lecture is devoted to the open vessel testing during the startup of a commercial BWR. A control rod calibration is also illustrated. (author)

  7. BWR Refill-Reflood Program. Final report

    International Nuclear Information System (INIS)

    Myers, L.L.

    1983-09-01

    The BWR Refill-Reflood Program is part of the continuing Loss of Coolant Accident (LOCA) research in the United States which is jointly sponsored by the Nuclear Regulatory Commission, the Electric Power Research Institute, and the General Electric Company. The current program expanded the focus of this research to include full scale experimental evaluations of multidimensional and multichannel effects during system refill. The program has also made major contributions to the BWR version of the Transient Reactor Analysis Code (TRAC) which has been developed cooperatively with the Idaho National Engineering Laboratory (INEL) for application to BWR transients. A summary description of the complete program is provided including the principal findings and main conclusions of the program. The results of the program have shown that multidimensional and parallel channel effects have the potential to significantly improve the system response over that observed in single channel tests

  8. Interpretation of incore noise measurements in BWR's

    International Nuclear Information System (INIS)

    Dam, H. van

    1983-01-01

    A survey is given of the main incentives for power reactor noise research, and the differences and similarities of noise in power and zero power systems are shown. After a short outline of historical developments the basic characteristics of the adjoint method in reactor noise theory are dealt with. The detector adjoint functions describe the transfer functions between spatially distributed noise sources and a (neutron or gamma) detector. In particular, the spatial dependence of these functions explains the 'local' and 'global' effects in BWR noise measurements. By including thermal hydraulic feedback effects in the adjoint analysis, it is shown that the common idea of a dominant global effect at low frequencies, which should result in point kinetic behaviour, is erroneous. The same analysis provides a method for nonperturbing on-line measurements on a BWR in The Netherlands. In the final part of the paper some ideas are given for further research in the field of BWR noise. (author)

  9. ABB advanced BWR and PWR fuel

    International Nuclear Information System (INIS)

    Junkrans, S.; Helmersson, S.; Andersson, S.

    1999-01-01

    Fuel designed and fabricated by ABB is now operating in 40 PWRs and BWRs in Europe, the United States and Korea. An excellent fuel reliability track record has been established. High burnups are proven for both BWR and PWR. Thermal margin improving features and advanced burnable absorber concepts enable the utilities to adopt demanding duty cycles to meet new economic objectives. In particular we note the excellent reliability record of ABB PWR fuel equipped with Guardian TM debris filter, proven to meet the -6 rod-cycles fuel failure goal, and the out-standing operating record of the SVEA 10x10 BWR fuel, where ABB is the only vendor to date with multi batch experience to high burnup. ABB is dedicated to maintain high fuel reliability as well as continually improve and develop a broad line of BWR and PWR products. ABB's development and fuel follow-up activities are performed in close co-operation with its customers. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-15

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

  11. BWR radiation exposure--experience and projection

    International Nuclear Information System (INIS)

    Falk, C.F.; Wilkinson, C.D.; Hollander, W.R.

    1979-01-01

    The BWR/6 Mark III radiation exposures are projected to be about half of those of current average operating experience of 725 man-rem. These projections are said to be realistic and based on current achievements and not on promises of future development. The several BWRs operating with low primary system radiation levels are positive evidence that radiation sources can be reduced. Improvements have been made in reducing the maintenance times for the BWR/6, and further improvements can be made by further attention to cost-effective plant arrangement and layout during detail design to improve accessibility and maintainability of each system and component

  12. General Electric's training program for BWR chemists

    International Nuclear Information System (INIS)

    Osborn, R.N.; Lim, W.

    1981-01-01

    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

  13. BWR plant analyzer development at BNL

    International Nuclear Information System (INIS)

    Cheng, H.S.; Wulff, W.; Mallen, A.N.; Lekach, S.V.; Stritar, A.; Cerbone, R.J.

    1985-01-01

    Advanced technology for high-speed interactive nuclear power plant simulations is of great value for timely resolution of safety issues, for plant monitoring, and for computer-aided emergency responses to an accident. Presented is the methodology employed at BNL to develop a BWR plant analyzer capable of simulating severe plant transients at much faster than real-time process speeds. Five modeling principles are established and a criterion is given for selecting numerical procedures and efficient computers to achieve the very high simulation speeds. Typical results are shown to demonstrate the modeling fidelity of the BWR plant analyzer

  14. The BWR Hybrid 4 control rod

    International Nuclear Information System (INIS)

    Gross, H.; Fuchs, H.P.; Lippert, H.J.; Dambietz, W.

    1988-01-01

    The service life of BWR control rods designed in the past has been unsatisfactory. The main reason was irradiation assisted stress corrosion cracking of B 4 C rods caused by external swelling of the B 4 C powder. By this reason KWU developed an improved BWR control rod (Hybrid 4 control rod) with extended service life and increased control rod worth. It also allows the procedure for replacing and rearranging fuel assemblies to be considerably simplified. A complete set of Hydbrid 4 control rods is expected to last throughout the service life of a plant (assumption: ca. 40 years) if an appropriate control rod reshuffling management program is used. (orig.)

  15. Improvement for BWR operator training, 3

    International Nuclear Information System (INIS)

    Noji, Kunio; Toeda, Susumu; Saito, Genhachi; Suzuki, Koichi

    1990-01-01

    BWR Operator Training Center Corporation (BTC) is conducting training for BWR plant operators using Full-scope Simulators. There are several courses for individual operators and one training course for shift crew (Family Training Course) in BTC. Family Training is carried out by all members of the operating shift-crew. BTC has made efforts to improve the Family Training in order to acquire more effective training results and contribute to up-grade team performance of all crews. This paper describes some items of our efforts towards Family Training improvement. (author)

  16. Efforts for optimization of BWR core internals replacement

    International Nuclear Information System (INIS)

    Iizuka, N.

    2000-01-01

    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)

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

    International Nuclear Information System (INIS)

    Cole, Steven E.; Garner, Norman L.; Lippert, Hans-Joachim; Graebert, Rüdiger; Mollard, Pierre; Hahn, Gregory C.

    2014-01-01

    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)

  18. Panorama of the BWR reactors - Evolution of the concept

    Energy Technology Data Exchange (ETDEWEB)

    Novotny, C.; Uhrig, E. [AREVA NP GmbH, Safety Engineering Department - PEPS-G (Germany)

    2012-01-15

    Nowadays, a fleet of more than 50 boiling water reactors (BWR) are in operation in the world. This article gives a short overview on the developments of nuclear power plants of the BWR type, with a focus on the European builds. It describes the technical bases from the early designs in the fifties, sketches the innovations of the sixties and seventies in the types BWR 69 and 72 (Baulinie 69 and 72) and gives an outlook of a possible next generation BWR. A promising approach in recent BWR developments is the the combination of passive safety systems with established design basis

  19. Fissile material proliferation risk

    International Nuclear Information System (INIS)

    Dreicer, J.S.; Rutherford, D.A.

    1996-01-01

    The proliferation risk of a facility depends 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 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. 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 nuclear related sites and facilities in different countries and still ensure a global decrease in proliferation risk for fissile material (plutonium and highly enriched uranium)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-11-27

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

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

    International Nuclear Information System (INIS)

    Villagran, Marcelo A.; Gutierrez-Castro, Francisco A.; Pantoja, Diego F.; Alarcon, Jose C.; Fariña, Macarena A.; Amigo, Romina F.; Muñoz-Godoy, Natalia A.; Pinilla, Mabel G.; Peña, Eduardo A.; Gonzalez-Chavarria, Ivan; Toledo, Jorge R.; Rivas, Coralia I.; Vera, Juan C.; McNerney, Eileen M.; Onate, Sergio A.

    2015-01-01

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

  2. Moderator temperature coefficient in BWR core

    International Nuclear Information System (INIS)

    Naito, Yoshitaka

    1977-01-01

    Temperature dependences of infinite multiplication factor k sub(infinity) and neutron leakage from the core must be examined for estimation of moderator temperature coefficient. Temperature dependence on k sub(infinity) has been investigated by many researchers, however, the dependence on neutron leakage of a BWR with cruciformed control rods has hardly been done. Because there are difficulties and necessity on calculations of three space dimensional and multi-energy groups neutron distribution in a BWR core. In this study, moderator temperature coefficients of JPDR-II (BWR) core were obtained by calculation with DIFFUSION-ACE, which is newly developed three-dimensional multi-group computer code. The results were compared with experimental data measured from 20 to 275 0 C of the moderator temperature and the good agreement was obtained between calculation and measurement. In order to evaluate neutron leakage from the core, the other two calculations were carried out, adjusting criticality by uniform absorption rate and by material buckling. The former underestimated neutron leakage and the latter overestimated it. Discussion on the results shows that in order to estimate the temperature coefficient of BWR, neutron leakage must be evaluated precisely, therefore the calculation at actual pattern of control rods is necessary. (auth.)

  3. BWR vessel and internals project (BWRVIP)

    International Nuclear Information System (INIS)

    Bilanin, W.J.; Dyle, R.L.

    1996-01-01

    Recent Boiling Water Reactor (BWR) inspections indicate that Intergranular Stress Corrosion Cracking (IGSCC) is a significant technical issue for some BWR internals. IN response, the Boiling Water Reactor Vessel and Internals Project (BWRVIP) was formed by an associated of domestic and international utilities which own and operate BWRs. The project is identifying or developing generic, cost-effective strategies for managing degradation of reactor internals from which each utility can select the alternative most appropriate for their plant. The Electric Power Research Institute manages the technical program, implementing the utility defined programs. The BWRVIP is organized into four technical tasks: Assessment, Inspection, Repair and Mitigation. An Integration task coordinates the work. The goal of the Assessment task is to develop methodologies for evaluation of vessel and internal components in support of decisions for operation, inspection, mitigation or repair. The goal of the Inspection task is to develop and assess effective and predictable inspection techniques which can be used to determine the condition of BWR vessel and internals that are potentially susceptible to service-related SCC degradation. The goal of the Repair task is to assure the availability of cost-effective repair/replacement alternatives. The goal of the Mitigation task is to develop and demonstrate countermeasures for SCC degradation. This paper summarizes the BWRVIP approach for addressing BWR internals SCC degradation and illustrates how utilities are utilizing BWRVIP products to successfully manage the effect of SCC on core shrouds

  4. Delivering high performance BWR fuel reliably

    International Nuclear Information System (INIS)

    Schardt, J.F.

    1998-01-01

    Utilities are under intense pressure to reduce their production costs in order to compete in the increasingly deregulated marketplace. They need fuel, which can deliver high performance to meet demanding operating strategies. GE's latest BWR fuel design, GE14, provides that high performance capability. GE's product introduction process assures that this performance will be delivered reliably, with little risk to the utility. (author)

  5. Secondary systems of PWR and BWR

    International Nuclear Information System (INIS)

    Schindler, N.

    1981-01-01

    The secondary systems of a nuclear power plant comprises the steam, condensate and feedwater cycle, the steam plant auxiliary or ancillary systems and the cooling water systems. The presentation gives a general review about the main systems which show a high similarity of PWR and BWR plants. (orig./RW)

  6. Assessment of two BWR accident management strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Petek, M.

    1991-01-01

    Candidate mitigative strategies for management of in-vessel events during the late phase (after core degradation has occurred) of postulated 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 additional assessment. The first is 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 pertains to the opposite case, for which vessel injection would be restored after control blade melting had begun; its purpose is 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 have been performed during 1991 under the auspices of the Detailed Assessment of BWR In-Vessel Strategies Program. This paper provides a discussion of the motivation for and purpose of these strategies and the potential for their success. 33 refs., 9 figs

  7. BWR stability analysis at Brookhaven National Laboratory

    International Nuclear Information System (INIS)

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

    1991-01-01

    Following the unexpected, but safely terminated, power and flow oscillations in the LaSalle-2 Boiling Water Reactor (BWR) on March 9, 1988, the Nuclear Regulatory Commission (NRC) Offices of Nuclear Reactor Regulation (NRR) and of Analysis and Evaluation of Operational Data (AEOD) requested that the Office of Nuclear Regulatory Research (RES) carry out BWR stability analyses, centered around fourteen specific questions. Ten of the fourteen questions address BWR stability issues in general and are dealt with in this paper. The other four questions address local, out-of-phase oscillations and matters of instrumentation; they fall outside the scope of the work reported here. It was the purpose of the work documented in this report to answer ten of the fourteen NRC-stipulated questions. Nine questions are answered by analyzing the LaSalle-2 instability and related BWR transients with the BNL Engineering Plant Analyzer (EPA) and by performing an uncertainty assessment of the EPA predictions. The tenth question is answered on the basis of first principles. The ten answers are summarized

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

    International Nuclear Information System (INIS)

    Ohyama, Nobuhiro; Suda, Kenichi; Ogata, Hiroshi; Matsuda, Shinichi; Nagasaka, Kazuhiro; Fujii, Toshi; Nozawa, Toshiya; Ishihama, Kiyoshi; Higuchi, Tomokazu

    2004-01-01

    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. IGSCC in cold worked austenitic stainless steel in BWR environment

    International Nuclear Information System (INIS)

    Persson, B.; Lindblad, B.

    1989-09-01

    The survey shows that austenitic stainless steels in a cold worked condition can exhibit IGSCC in BWR environment. It is also found that IGSCC often is initiated as a transgranular crack. Local stresses and surface defects very often acts as starting points for IGSCC. IGSCC due to cold working requires a cold working magnitude of at leas 5%. During cold working a formation of mechanical martensite can take place. The transgranular corrosion occurs in the martensitic phase due to sensitation. The crack propagates integranularly due to anodic solvation of α'-martensite. Sensitation of the martensitic phase is fasten in BCC-structures than in a FCC-structures mainly due to faster diffusion of chromium and carbon which cause precipitation of chromium carbides. Experiments show that a carbon content as low as 0.008% is enough for the formation of 68% martensite and for sensitation. Hydrogen induced cracking is regarded as a mechanism which can accelerate IGSCC. Such cracking requires a hydrostatic stress near the crack tip. Since the oxide in the crack tip is relatively impermeable to hydrogen, cracks in the oxide layer are required for such embrittlement. Hydrogen induced embrittlement of the martensitic phase, at the crack tip, can cause crack propagation. Solution heat treated unstabilized stainless steels are regarded to have a good resistance to IGSCC if they have not undergone cold working. In general, though, Mo-alloyed steels have a better resistance to IGSCC in BWR environment. Regarding the causes for IGSCC, the present literature survey shows that many mechanisms are suggested. To provide a safer ground for the estimation of crack propagation rates, SA recommends SKI to finance a project with the aim to determine the crack propagation rate on proper material. (authors) (65 refs.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-03-01

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

  11. BWR NSSS design basis documentation

    International Nuclear Information System (INIS)

    Vij, R.S.; Bates, R.E.

    2004-01-01

    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. (author)

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

    Science.gov (United States)

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

    2017-12-01

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

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

    DEFF Research Database (Denmark)

    Krejsgaard, T; Helvad, Rikke; Ralfkiær, Elisabeth

    2010-01-01

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

  14. EPRI BWR Water Chemistry Guidelines Revision

    International Nuclear Information System (INIS)

    Garcia, Susan E.; Giannelli, Joseph F.

    2014-01-01

    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

  15. Options for small and medium sized reactors (SMRs) to overcome loss of economies of scale and incorporate increased proliferation resistance and energy security

    International Nuclear Information System (INIS)

    Kuznetsov, Vladimir

    2008-01-01

    The designers of innovative small and medium sized reactors pursue new design and deployment strategies making use of certain advantages provided by smaller reactor size and capacity to achieve reduced design complexity and simplified operation and maintenance requirements, and to provide for incremental capacity increase through multi-module plant clustering. Competitiveness of SMRs depends on the incorporated strategies to overcome loss of economies of scale but equally it depends on finding appropriate market niches for such reactors. For many less developed countries, these are the features of enhanced proliferation resistance and increased robustness of barriers for sabotage protection that may ensure the progress of nuclear power. For such countries, small reactors without on-site refuelling, designed for infrequent replacement of well-contained fuel cassette(s) in a manner that impedes clandestine diversion of nuclear fuel material, may provide a solution. Based on the outputs of recent IAEA activities for innovative SMRs, the paper provides a summary of the state-of-the-art in approaches to improve SMR competitiveness and incorporate enhanced proliferation resistance and energy security. (author)

  16. Design and deployment strategies for small and medium sized reactors (SMRs) to overcome loss of economies of scale and incorporate increased proliferation resistance

    International Nuclear Information System (INIS)

    Kuznetsov, V.

    2007-01-01

    The designers of innovative small and medium sized reactors pursue new design and deployment strategies making use of certain advantages provided by smaller reactor size and capacity to achieve reduced design complexity and simplified operation and maintenance requirements, and to provide for incremental capacity increase through multi-module plant clustering. Competitiveness of SMRs (Small and Medium size Reactor) depends on the incorporated strategies to overcome loss of economies of scale but equally it depends on finding appropriate market niches for such reactors. For many less developed countries, these are the features of enhanced proliferation resistance and increased robustness of barriers for sabotage protection that may ensure the progress of nuclear power. For such countries, small reactors without on-site refuelling, designed for infrequent replacement of well-contained fuel cassette(s) in a manner that impedes clandestine diversion of nuclear fuel material, may provide a solution. Based on the outputs of recent IAEA activities for innovative SMRs, the paper provides a summary of the state-of-the-art in approaches to improve SMR competitiveness and incorporate enhanced proliferation resistance and energy security. (author)

  17. Development of a BWR core burn-up calculation code COREBN-BWR

    International Nuclear Information System (INIS)

    Morimoto, Yuichi; Okumura, Keisuke

    1992-05-01

    In order to evaluate core performances of BWR type reactors, the three dimensional core burnup calculation code COREBN-BWR and the fuel management code HIST-BWR have been developed. In analyses of BWR type reactors, thermal hydraulics calculations must be coupled with neutronics calculations to evaluate core performances, because steam void distribution changes according to the change of the power distribution. By installing new functions as follows to the three dimensional core burnup code COREBN2 developed in JAERI for PWR type reactor analyses, the code system becomes to be applicable to burnup analyses of BWR type reactors. (1) Macroscopic cross section calculation function taking into account of coolant void distribution. (2) Thermal hydraulics calculation function to evaluate core flow split, coolant void distribution and thermal margin. (3) Burnup calculation function under the Haling strategy. (4) Fuel management function to incorporate the thermal hydraulics information. This report consists of the general description, calculational models, input data requirements and their explanations, detailed information on usage and sample input. (author)

  18. Hsp90 inhibitor 17-AAG reduces ErbB2 levels and inhibits proliferation of the trastuzumab resistant breast tumor cell line JIMT-1.

    Science.gov (United States)

    Zsebik, Barbara; Citri, Ami; Isola, Jorma; Yarden, Yosef; Szöllosi, János; Vereb, György

    2006-04-15

    ErbB2, a member of the EGF receptor family of tyrosine kinases is overexpressed on many tumor cells of epithelial origin and is the molecular target of trastuzumab (Herceptin), the first humanized antibody used in the therapy of solid tumors. Trastuzumab, which is thought to act, at least in part, by downregulating ErbB2 expression is only effective in approximately 30-40% of ErbB2 positive breast tumors. Geldanamycin and its derivative 17-AAG are potential antitumor agents capable of downregulating client proteins of Hsp90, including ErbB2. To investigate the ability of 17-AAG to downregulate ErbB2 in trastuzumab resistant breast cancer cells and the possibility of 17-AAG and trastuzumab potentiating each other's effect, the recently established trastuzumab resistant breast cancer cell line, JIMT-1 was compared to the known trastuzumab sensitive SKBR-3 line. Baseline and stimulus-evoked dimerization and activation levels of ErbB2, and the effects of trastuzumab and 17-AAG alone and in combination on cell proliferation and apoptosis, as well as on ErbB2 expression and phosphorylation have been measured. Baseline activation and amenability to activation and downregulation by trastuzumab was much lower in the resistant line. However, 17-AAG enhanced ErbB2 homodimerization after 5-10 min of treatment in both cell lines, and decreased proliferation with an IC50 of 70 nM for SKBR-3 and 10nM for JIMT-1. Thus, 17-AAG may be a useful drug in trastuzumab resistant ErbB2 overexpressing tumors. The antiproliferative effect of 17-AAG was positively correlated with phosphorylation and downregulation of ErbB2 and was dominated by apoptosis, although, especially at higher doses, necrosis was also present. Interestingly, IC50 values for ErbB2 downregulation and phosphorylation, in the 30-40 nM range, were not significantly different for the two cell lines. This observation and the negative correlation between resting ErbB2 levels and the antiproliferative effect of 17-AAG may

  19. Crud deposition modeling on BWR fuel rods

    International Nuclear Information System (INIS)

    Kucuk, Aylin; Cheng, Bo; Potts, Gerald A.; Shiralkar, Bharat; Morgan, Dave; Epperson, Kenny; Gose, Garry

    2014-01-01

    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)

  20. BWR radiation buildup control with ionic zinc

    International Nuclear Information System (INIS)

    Marble, W.J.; Wood, C.J.; Leighty, C.E.; Green, T.A.

    1986-01-01

    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

  1. BWR mechanics and materials technology update

    International Nuclear Information System (INIS)

    Kiss, E.

    1983-01-01

    This paper discusses technical results obtained from a variety of important programs underway at General Electric's Nuclear Engineering Division. The principal objective of these programs is to qualify and improve BWR product related technologies that fall broadly under the disciplines of Applied Mechanics and Materials Engineering. The paper identifies and deals with current technical issues that are of general importance to the LWR industry albeit the specific focus is directed to the development and qualification of analytical predictive methods and criteria, and improved materials for use in the design of the BWR. In this paper, specific results and accomplishments are summarized to provide a braod perspective of technology advances. Results are presented in sections which discuss: dynamic analysis and modeling; fatigue and fracture evaluation; materials engineering advances; and flow induced vibration. (orig.)

  2. Parallel channel effects under BWR LOCA conditions

    International Nuclear Information System (INIS)

    Suzuki, H.; Hatamiya, S.; Murase, M.

    1988-01-01

    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.)

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

    Directory of Open Access Journals (Sweden)

    Sanchita Das

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

  4. Eulerian fluid-structure analysis of BWR

    International Nuclear Information System (INIS)

    McMaster, W.H.

    1979-05-01

    A fluid-structure-interaction algorithm is developed for the analysis of the dynamic response of a BWR pressure-suppression pool and containment structure. The method is incorporated into a two-dimensional semi-implicit Eulerian hydrodynamics code, PELE-IC, for the solution of incompressible flow coupled to flexible structures. The fluid, structure, and coupling algorithms have been verified by calculation of solved problems from the literature and by comparison with air and steam blowdown experiments

  5. Methyl Iodide Decomposition at BWR Conditions

    International Nuclear Information System (INIS)

    Pop, Mike; Bell, Merl

    2012-09-01

    Based on favourable results from short-term testing of methanol addition to an operating BWR plant, AREVA has performed numerous studies in support of necessary Engineering and Plant Safety Evaluations prior to extended injection of methanol. The current paper presents data from a study intended to provide further understanding of the decomposition of methyl iodide as it affects the assessment of methyl iodide formation with the application of methanol at BWR Plants. This paper describes the results of the decomposition testing under UV-C light at laboratory conditions and its effect on the subject methyl iodide production evaluation. The study as to the formation and decomposition of methyl iodide as it is effected by methanol addition is one phase of a larger AREVA effort to provide a generic plant Safety Evaluation prior to long-term methanol injection to an operating BWR. Other testing phases have investigated the compatibility of methanol with fuel construction materials, plant structural materials, plant consumable materials (i.e. elastomers and coatings), and ion exchange resins. Methyl iodide is known to be very unstable, typically preserved with copper metal or other stabilizing materials when produced and stored. It is even more unstable when exposed to light, heat, radiation, and water. Additionally, it is known that methyl iodide will decompose radiolytically, and that this effect may be simulated using ultra-violet radiation (UV-C) [2]. In the tests described in this paper, the use of a UV-C light source provides activation energy for the formation of methyl iodide. Thus is similar to the effect expected from Cherenkov radiation present in a reactor core after shutdown. Based on the testing described in this paper, it is concluded that injection of methanol at concentrations below 2.5 ppm in BWR applications to mitigate IGSCC of internals is inconsequential to the accident conditions postulated in the FSAR as they are related to methyl iodide formation

  6. Delivering high performance BWR fuel reliably

    Energy Technology Data Exchange (ETDEWEB)

    Schardt, J.F. [GE Nuclear Energy, Wilmington, NC (United States)

    1998-07-01

    Utilities are under intense pressure to reduce their production costs in order to compete in the increasingly deregulated marketplace. They need fuel, which can deliver high performance to meet demanding operating strategies. GE's latest BWR fuel design, GE14, provides that high performance capability. GE's product introduction process assures that this performance will be delivered reliably, with little risk to the utility. (author)

  7. The HAMBO BWR simulator of HAMMLAB

    International Nuclear Information System (INIS)

    Karlsson, Tommy; Jokstad, Haakon; Meyer, Brita D.; Nihlwing, Christer; Norrman, Sixten; Puska, Eija Karita; Raussi, Pekka; Tiihonen, Olli

    2001-02-01

    Modernisation of control rooms of the nuclear power plants has been a major issue in Sweden and Finland the last few years, and this will continue in the years to come. As an aid in the process of introducing new technology into the control rooms, the benefit of having an experimental simulator where proto typing of solutions can be performed, has been emphasised by many plants. With this as a basis, the BWR plants in Sweden and Finland decided to fund, in co-operation with the Halden Project, an experimental BWR simulator based on the Forsmark 3 plant in Sweden. The BWR simulator development project was initiated in January 1998. VTT Energy in Finland developed the simulator models with the aid of their APROS tool, while the operator interface was developed by the Halden Project. The simulator was thoroughly tested by experienced HRP personnel and professional Forsmark 3 operators, and accepted by the BWR utilities in June 2000. The acceptance tests consisted of 19 well-defined transients, as well as the running of the simulator from full power down to cold shutdown and back up again with the use of plant procedures. This report describes the HAMBO simulator, with its simulator models, the operator interface, and the underlying hardware and software infrastructure. The tools used for developing the simulator, APROS, Picasso-3 and the Integration Platform, are also briefly described. The acceptance tests are described, and examples of the results are presented, to illustrate the level of validation of the simulator. The report concludes with an indication of the short-term usage of the simulator. (Author)

  8. PREDICTIVE METHODS FOR STABILITY MARGIN IN BWR

    OpenAIRE

    MELARA SAN ROMÁN, JOSÉ

    2016-01-01

    [EN] Power and flow oscillations in a BWR are very undesirable. One of the major concerns is to ensure, during power oscillations, compliance with GDC 10 and 12. GDC 10 requires that the reactor core be designed with appropriate margin to assure that specified acceptable fuel design limits will not be exceeded during any condition of normal operation, including the effects of anticipated operational occurrences. GDC 12 requires assurance that power oscillations which can result in conditions ...

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

    International Nuclear Information System (INIS)

    Sekimoto, H.

    2010-01-01

    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

  10. BWR level estimation using Kalman Filtering approach

    International Nuclear Information System (INIS)

    Garner, G.; Divakaruni, S.M.; Meyer, J.E.

    1986-01-01

    Work is in progress on development of a system for Boiling Water Reactor (BWR) vessel level validation and failure detection. The levels validated include the liquid level both inside and outside the core shroud. This work is a major part of a larger effort to develop a complete system for BWR signal validation. The demonstration plant is the Oyster Creek BWR. Liquid level inside the core shroud is not directly measured during full power operation. This level must be validated using measurements of other quantities and analytic models. Given the available sensors, analytic models for level that are based on mass and energy balances can contain open integrators. When such a model is driven by noisy measurements, the model predicted level will deviate from the true level over time. To validate the level properly and to avoid false alarms, the open integrator must be stabilized. In addition, plant parameters will change slowly with time. The respective model must either account for these plant changes or be insensitive to them to avoid false alarms and maintain sensitivity to true failures of level instrumentation. Problems are addressed here by combining the extended Kalman Filter and Parity Space Decision/Estimator. The open integrator is stabilized by integrating from the validated estimate at the beginning of each sampling interval, rather than from the model predicted value. The model is adapted to slow plant/sensor changes by updating model parameters on-line

  11. Utility experience with BWR-PSMS

    International Nuclear Information System (INIS)

    Bond, G.R.

    1986-01-01

    The BWR Power Shape Monitoring System (BWR-PSMS) has proven to be an effective and versatile tool for core monitoring. GPU Nuclear Corporation's (GPUN) Oyster Creek plant has been involved in the PSMS development since its inception, having been selected by EPRI as the initial demonstration site. Beginning with Cycle 10, Oyster Creek has been applying the BWR-PSMS as the primary core monitoring tool. Although the system has been in operation at Oyster Creek for the past several cycles, this is the first time the PSMS was used to monitor compliance to the plant technical specifications, to guide adherence to vendore fuel maneuvering recommendations and to develop data for certain performance records such as fuel burnup, isotopic accounting, etc. This paper will discuss the bases for the decision to apply PSMS as the fundamental core monitoring system, the experience in implementing the PSMS in this mode, activities currently underway or planned related to PSMS, and potential future extensions and applications of PSMS at Oyster Creek

  12. Economic analysis of hydride fueled BWR

    International Nuclear Information System (INIS)

    Ganda, F.; Shuffler, C.; Greenspan, E.; Todreas, N.

    2009-01-01

    The economic implications of designing BWR cores with hydride fuels instead of conventional oxide fuels are analyzed. The economic analysis methodology adopted is based on the lifetime levelized cost of electricity (COE). Bracketing values (1970 and 3010 $/kWe) are used for the overnight construction costs and for the power scaling factors (0.4 and 0.8) that correlate between a change in the capital cost to a change in the power level. It is concluded that a newly constructed BWR reactor could substantially benefit from the use of 10 x 10 hydride fuel bundles instead of 10 x 10 oxide fuel bundles design presently in use. The cost saving would depend on the core pressure drop constraint that can be implemented in newly constructed BWRs - it is between 2% and 3% for a core pressure drop constraint as of the reference BWR, between 9% and 15% for a 50% higher core pressure drop, and between 12% and 21% higher for close to 100% core pressure. The attainable cost reduction was found insensitive to the specific construction cost but strongly dependent on the power scaling factor. The cost advantage of hydride fuelled cores as compared to that of the oxide reference core depends only weakly on the uranium and SWU prices, on the 'per volume base' fabrication cost of hydride fuels, and on the discount rate used. To be economically competitive, the uranium enrichment required for the hydride fuelled core needs to be around 10%.

  13. Investigation of BWR stability in Forsmark 2

    International Nuclear Information System (INIS)

    Oguma, R.; Reisch, F.; Bergdahl, B.G.; Lorenzen, J.; Aakerhielm, F.; Kellner, S.

    1988-01-01

    A series of noise measurements have been conducted at the Forsmark-2 reactor during its start-up operation after the revision in 1987. The main purpose was to investigate the BWR stability problem based on noise analysis, i.e. the problem of resonant power oscillation with frequency of about 0.5 Hz, which tends to arise at high power and low core flow condition. The noise analysis was performed to estimate the noise source which gives rise to the power oscillation, to evaluate the stability condition of the Forsmark-2 reactor in terms of the decay ratio (DR), as well as to investigate a safety related problem in connection with the BWR stability. The results indicate that the power oscillation is due to dynamic coupling between the neutron kinetics and thermal-hydraulics via void reactivity feedback. The DR reached as high as ≅ 0.7 at 63% of the rated power and 4100 kg/s of the total core flow. An investigation was made for the noise recording which represents a strong pressure oscillation with a peak frequency at 0.33 Hz. The result suggests that such pressure oscillation, if the peak frequency coincided with that of the resonant power oscillation, might become a cause of scram. The present noise analysis indicates the importance of a BWR on-line surveillance system with functions like stability condition monitoring and control system diagnosis. (orig.)

  14. Assessment of two BWR accident management strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Petek, M.

    1994-01-01

    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.))

  15. Scaling and uncertainty in BWR instability problems

    International Nuclear Information System (INIS)

    Di Auria, F.; Pellicoro, V.

    1995-01-01

    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)

  16. BWR recirculation pump diagnostic expert system

    International Nuclear Information System (INIS)

    Chiang, S.C.; Morimoto, C.N.; Torres, M.R.

    2004-01-01

    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)

  17. Interpretation of incore noise measurements in BWR's

    International Nuclear Information System (INIS)

    Dam, H. van

    1982-01-01

    A survey is given of the main incentives for power reactor noise research and the differences and similarities of noise in power and zero power systems are touched on. The basic characteristics of the adjoint method in reactor noise theory are treated. The detector adjoint functions describe the transfer functions between spatially distributed noise sources and a (neutron or gamma) detector. In particular, the spatial dependence of these functions explains the 'local' and 'global' effects in BWR noise measurements. By including thermal hydraulic feedback effects in the adjoint analysis, it is shown that the common idea of a dominant global effect at low frequencies which should result in point kinetic behaviour, is erroneous. The same analysis provides a method for nonperturbing on-line measurement of the reactor transfer function, which is demonstrated by results from measurements on a BWR in the Netherlands. In the final part of the paper some ideas are given for further research in the field of BWR noise. (author)

  18. 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.

  19. Stress corrosion cracking of L-grade stainless steels in boiling water reactor (BWR) plants

    International Nuclear Information System (INIS)

    Suzuki, Shunichi; Fukuda, Toshihiko; Yamashita, Hironobu

    2004-01-01

    L-grade stainless steels as 316NG, SUS316L and SUS304L have been used for the BWR reactor internals and re-circulation pipes as SCC resistant materials. However, SCC of the L-grade material components were reported recently in many Japanese BWR plants. The detail investigation of the components showed the fabrication process such as welding, machining and surface finishing strongly affected SCC occurrence. In this paper, research results of SCC of L-grade stainless steels, metallurgical investigation of core shrouds and re-circulation pipings, and features of SCC morphology were introduced. Besides, the structural integrity of components with SCC, countermeasures for SCC and future R and D planning were introduced. (author)

  20. BWR Refill-Reflood Program, Task 4.7 - model development: TRAC-BWR component models

    International Nuclear Information System (INIS)

    Cheung, Y.K.; Parameswaran, V.; Shaug, J.C.

    1983-09-01

    TRAC (Transient Reactor Analysis Code) is a computer code for best-estimate analysis for the thermal hydraulic conditions in a reactor system. The development and assessment of the BWR component models developed under the Refill/Reflood Program that are necessary to structure a BWR-version of TRAC are described in this report. These component models are the jet pump, steam separator, steam dryer, two-phase level tracking model, and upper-plenum mixing model. These models have been implemented into TRAC-B02. Also a single-channel option has been developed for individual fuel-channel analysis following a system-response calculation

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

    International Nuclear Information System (INIS)

    Todosow, M.; Galperin, A.

    2002-01-01

    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- 233 U 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

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

    Directory of Open Access Journals (Sweden)

    Pu Li

    2017-01-01

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

  3. Strategies of operation cycles in BWR type reactors

    International Nuclear Information System (INIS)

    Molina, D.; Sendino, F.

    1996-01-01

    The article analyzes the operation cycles in BWR type reactors. The cycle size of operation is the consequence on the optimization process of the costs with the technical characteristics of nuclear fuel and the characteristics of demand and production. The authors analyze the cases of Garona NP and Cofrentes NP, both with BWR reactors. (Author)

  4. GPE-BWR and the containment venting and filtering issue

    International Nuclear Information System (INIS)

    Palomo, J.; Santiago, J. de

    1988-01-01

    The Spanish Boiling Water Reactor Owner's Group (GPE-BWR) is formed by three utilities, owning four units: Santa Maria de Garona (46 MWe, BWR3, Mark I containment), Cofrentes (975 MWe, BWR6, Mark III containment) and Valdecaballeros (2x975 MWe, BWR6, Mark III containment) - all of the reactors having been supplied by General Electric. One of the GPE-BWR's several committees is the Safety and Licensing Committee, which follows up the evolution of severe accident topics and particularly the containment venting and filtering issue. In September 1987, the Consejo de Seguridad Nuclear (CSN), the Spanish Regulatory Body, asked the GPE-BWR to define its position on the installation of a containment venting system. The GPE-BWR created a Working Group which presented a Report on Containment Venting to the CSN in January 1987 gathered from: the US Nuclear Regulatory Commission (NRC); some US utilities; and several European countries, especially France, Germany and Sweden. CSN's review of the containment venting Report and the Action Plan proposed by the GPE-BWR finished in April 1988. The conclusion of the Report and the proposed Action Plan take into account the US NRC's identified open items on severe accidents and the R and D programs scheduled to close these items

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

    International Nuclear Information System (INIS)

    Sutherland, W.A.; Alamgir, M.; Findlay, J.A.; Hwang, W.S.

    1985-10-01

    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

  6. 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.

  7. 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.

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

    International Nuclear Information System (INIS)

    Zhao, L.; Li, N.; Yu, J.K.; Tang, H.T.; Li, Y.L.; He, M.; Yu, Z.J.; Bai, X.F.; Zheng, Z.H.; Wang, E.H.; Wei, M.J.

    2013-01-01

    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

  9. Development of the COMPRE-A Program for Evaluating Proliferation Resistance (PR) and Physical Protection (PP) of the Nuclear Facility

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sukhoon; Kim, Juyub; Shin, Kwangyoung [FNC Technology Co., Yongin (Korea, Republic of); Seo, Janghoon; Jang, Sungsoon; Yoo, Hosik [Korea Institute of Nuclear Non-proliferation and Control, Daejeon (Korea, Republic of)

    2016-10-15

    In order to make up for the limitation inherent in these methodologies, a new methodology called 'Comprehensive Methodology for PR and PP Evaluation (COMPRE)' is being developed by the Korea Institute of Nuclear Non-proliferation and Control (KINAC). Currently, a project for improving a computational program implementing the COMPRE methodology is ongoing in KINAC, and the 'COMPRE-A (i.e. COMPRE-Advanced model)' program was developed as part of this project. This paper describes details of the COMPRE-A program developed for establishing the platform corresponding to features of the evaluation factor derived for COMPRE, and for enhancing easiness of the result analysis through embodying the visualization and comparison tools for the evaluation result. As part of the project for improving a computational program implementing the COMPRE methodology, the COMPRE-A program was developed. This program provides not only the platform corresponding to features of the evaluation factor derived for COMPRE but also the visualization and comparison tools for enhancing easiness of the result analysis. The V and V for the developed program was performed through comparison with the result obtained by using MS-Excel for a sample case consisting of three (3) measures, six (6) sections and twenty-six (26) attributes. From the comparison result, it was confirmed that the risk in aspects of PR and PP at nuclear facilities could be quantitatively evaluated within the acceptable range using the COMPRE-A program.

  10. 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.

  11. Prevention of organic iodide formation in BWR's

    International Nuclear Information System (INIS)

    Karjunen, T.; Laitinen, T.; Piippo, J.; Sirkiae, P.

    1996-01-01

    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

  12. A simplified spatial model for BWR stability

    International Nuclear Information System (INIS)

    Berman, Y.; Lederer, Y.; Meron, E.

    2012-01-01

    A spatial reduced order model for the study of BWR stability, based on the phenomenological model of March-Leuba et al., is presented. As one dimensional spatial dependence of the neutron flux, fuel temperature and void fraction is introduced, it is possible to describe both global and regional oscillations of the reactor power. Both linear stability analysis and numerical analysis were applied in order to describe the parameters which govern the model stability. The results were found qualitatively similar to past results. Doppler reactivity feedback was found essential for the explanation of the different regions of the flow-power stability map. (authors)

  13. Level 2 PRA for a German BWR

    International Nuclear Information System (INIS)

    Sassen, F.; Rapp, W.; Tietsch, W.; Roess, P.

    2007-01-01

    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)

  14. Recycling systems for BWR type reactors

    International Nuclear Information System (INIS)

    Takagi, Akio; Yamamoto, Fumiaki; Fukumoto, Ryuji.

    1986-01-01

    Purpose: To stabilize the coolant flowing characteristics and reactor core reactivity. Constitution: The recycling system in a BWR type reactor comprises a recycling pump disposed to the outside of a reactor pressure vessel, a ring header connected to the recycling pump through main pipe ways, and a plurality of pipes branched from and connected with the ring header and connected to a plurality of jet pumps within the pressure vessel. Then, by making the diameter for the pipeways of each of the branched pipes different from each other, the effective cross-sectional area is varied to thereby average the coolant flow rate supplied to each of the jet pumps. (Seki, T.)

  15. Maintenance of BWR control rod drive mechanisms

    International Nuclear Information System (INIS)

    Greene, R.H.

    1991-01-01

    Control rod drive mechanism (CRDM) replacement and rebuilding is one of the highest dose, most physically demanding, and complicated maintenance activities routinely accomplished by BWR utilities. A recent industry workshop sponsored by the Oak Ridge National Laboratory, which dealt with the effects of CRDM aging, revealed enhancements in maintenance techniques and tooling which have reduced ALARA, improved worker comfort and productivity, and have provided revised guidelines for CRDM changeout selection. Highlights of this workshop and ongoing research on CRDM aging are presented in this paper

  16. 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.

  17. ECP measurements in the BWR-1 water loop relative to water composition changes

    Energy Technology Data Exchange (ETDEWEB)

    Kus, P.; Vsolak, R.; Kysela, J., E-mail: ksp@ujv.cz [Nuclear Research Inst. Rez plc, Husinec - Rez (Czech Republic); Hanawa, S.; Nakamura, T.; Uchida, S., E-mail: hanawa.satoshi@jaea.go.jp [Japan Atomic Energy Agency, Tokai-mura, Ibaraki (Japan)

    2010-07-01

    The goal of this study is to investigate the usage of ECP sensors in nuclear power plants. ECP sensors were tested using the LVR-15 reactor at the Nuclear Research Institute Rez plc (NRI) in the Czech Republic. The experiment took place on the BWR-1 loop, which was designed for investigating the behaviour of structural materials and radioactivity transport under BWR conditions. The BWR-1 loop facilitates irradiation experiments within a wide range of operating parameters (max. pressure of 10 MPa, max. temperature of 573 K and a neutron flux of 1.0* 10{sup 18} n/m{sup 2}s). This study involves the measurement of electrochemical potential (ECP). Corrosion potential is the main parameter for monitoring of water composition changes in nuclear power plants (NPP). The electrochemical potentials of stainless steel were measured under high temperatures in a test loop (BWR-1) under different water composition conditions. Total neutron flux was ∼10{sup -3} to ∼10{sup 12} n/cm{sup 2}s (>0.1 MeV) at a temperature of 560K, neutral pH, and water resistivity of 18.2 MOhm. ECP sensor response related to changes in water composition was monitored. Switching from NWC (normal water conditions) to HWC (hydrogen water conditions) was controlled using oxygen dosage. Water chemistry was monitored approx. 50 meters from the active channel. The active channel temperature was maintained within a range of 543 - 561 K from the start of irradiation for the entire duration of the experiment. A total of 24 reference electrodes composed of platinum (Pt), silver/silver chloride (Ag/AgCl) and a zircon membrane containing silver oxide (Ag{sub 2}O) powder were installed inside the active channel of the LVR-15 test reactor. The active channel (Field tube) was divided into four zones, with each zone containing six sensors. A mathematical radiolysis code model was created in cooperation with the Japan Atomic Energy Agency. (author)

  18. Phenomenology of BWR fuel assembly degradation

    Science.gov (United States)

    Kurata, Masaki; Barrachin, Marc; Haste, Tim; Steinbrueck, Martin

    2018-03-01

    Severe accidents occurred at the Fukushima-Daiichi Nuclear Power Station (FDNPS) which required an immediate re-examination of fuel degradation phenomenology. The present paper reviews the updated knowledge on the phenomenology of the fuel degradation, focusing mainly on the BWR fuel assembly degradation at the macroscopic scale and that of the individual interactions at the meso-scale. Oxidation of boron carbide (B4C) control rods potentially generates far larger amounts of heat and hydrogen under BWR accident conditions. All integral tests with B4C control rods or control blades have shown early failure, liquefaction, relocation and oxidation of B4C starting at temperatures around 1250 °C, well below the significant interaction temperatures of UO2-Zry. These interactions or reactions potentially influence the progress of fuel degradation in the early phase. The steam-starved conditions, which are being discussed as a likely scenario at the FDNPS accident, highly influence the individual interactions and potentially lead the fuel degradation in non-prototypical directions. The detailed phenomenology of individual interactions and their influence on the transient and on the late phase of the severe accidents are also discussed.

  19. BWR normal water chemistry guidelines: 1986 revision

    International Nuclear Information System (INIS)

    1988-09-01

    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

  20. BWR Assembly Optimization for Minor Actinide Recycling

    International Nuclear Information System (INIS)

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

    2010-01-01

    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.

  1. Recent BWR fuel management reactor physics advances

    International Nuclear Information System (INIS)

    Crowther, R.L.; Congdon, S.P.; Crawford, B.W.; Kang, C.M.; Martin, C.L.; Reese, A.P.; Savoia, P.J.; Specker, S.R.; Welchly, R.

    1982-01-01

    Improvements in BWR fuel management have been under development to reduce uranium and separative work (SWU) requirements and reduce fuel cycle costs, while also maintaining maximal capacity factors and high fuel reliability. Improved reactor physics methods are playing an increasingly important role in making such advances feasible. The improved design, process computer and analysis methods both increase knowledge of the thermal margins which are available to implement fuel management advance, and improve the capability to reliably and efficiently analyze and design for fuel management advances. Gamma scan measurements of the power distributions of advanced fuel assembly and advanced reactor core designs, and improved in-core instruments also are important contributors to improving 3-d predictive methods and to increasing thermal margins. This paper is an overview of the recent advances in BWR reactor physics fuel management methods, coupled with fuel management and core design advances. The reactor physics measurements which are required to confirm the predictions of performance fo fuel management advances also are summarized

  2. CECP, Decommissioning Costs for PWR and BWR

    International Nuclear Information System (INIS)

    Bierschbach, M.C.

    1997-01-01

    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)

  3. A BWR licensing experience in the USA

    International Nuclear Information System (INIS)

    Powers, J.; Ogura, C.; Arai, K.; Thomas, S.; Mookhoek, B.

    2015-09-01

    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)

  4. BWR-stability investigation at Forsmark 1

    International Nuclear Information System (INIS)

    Bergdahl, B.G.; Reisch, F.; Oguma, R.; Lorenzen, J.; Aakerhielm, F.

    1988-01-01

    A series of noise measurements have been conducted at Forsmark 1 during start-up operation after the revision summer '87. The main purpose was to investigate BWR-stability problems, i.e. resonant power oscillations of 0.5 Hz around 65% power and 4100 kg/s core flow, which tend to arise at high power and low core flow conditions. The analysis was performed to estimate the noise source which gives rise to the oscillation, to evaluate the measure of stability, i.e. the Decay Ratio (Dr) as well as to investigate other safety related problems. The result indicates that the oscillation is due to the dynamic coupling between the neutron kinetics and thermal hydraulics via void reactivity feedback. The Dr ranged between values of 0.7 and > 0.9, instead of expected 0.6 (Dr=1 is defined as instability). These high values imply that the core cannot suppress oscillations fast enough and a small perturbation can cause scram. Further it was found that the entire core is oscillating in phase (LPRM's) with varying strength where any connection to the consequences of different fuel (8x8, 9x9) being present simultaneously cannot be excluded. This report elucidates the importance of an on-line BWR-stability surveillance system with functions like stability condition monitoring and control system diagnosis. (orig.)

  5. Hydrogen injection device in BWR type reactor

    International Nuclear Information System (INIS)

    Takagi, Jun-ichi; Kubo, Koji.

    1988-01-01

    Purpose: To reduce the increasing ratio of main steam system dose rate due to N-16 activity due to excess hydrogen injection in the hydrogen injection operation of BWR type reactors. Constitution: There are provided a hydrogen injection mechanism for injecting hydrogen into primary coolants of a BWR type reactor, and a chemical injection device for injecting chemicals such as methanol, which makes nitrogen radioisotopes resulted in the reactor water upon hydrogen injection non-volatile, into the pressure vessel separately from hydrogen. Injected hydrogen and the chemicals are not reacted in the feedwater system, but the reaction proceeds due to the presence of radioactive rays after the injection into the pressure vessel. Then, hydrogen causes re-combination in the downcomer portion to reduce the dissolved oxygen concentration. Meanwhile, about 70 % of the chemicals is supplied by means of a jet pump directly to the reactor core, thereby converting the chemical form of N-16 in the reactor core more oxidative (non-volatile). (Kawakami, Y.)

  6. 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.

  7. 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)

  8. LBB application in Swedish BWR design

    International Nuclear Information System (INIS)

    Kornfeldt, H.; Bjoerk, K.O.; Ekstroem, P.

    1997-01-01

    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

  9. BWR fuel experience with zinc injection

    International Nuclear Information System (INIS)

    Levin, H.A.; Garcia, S.E.

    1995-01-01

    In 1982 a correlation between low primary recirculation system dose rates in BWR's and the presence of ionic zinc in reactor water was identified. The source of the zinc was primarily from Admiralty brass condensers. Plants with brass condensers are called ''natural zinc'' plants. Brass condensers were also a source of copper that was implicated in crude induced localized corrosion (CILC) fuel failures. In 1986 the first BWR intentionally injected zinc for the benefits of dose rate control. Although zinc alone was never implicated in fuel degradation of failures, a comprehensive fuel surveillance program was initiated to monitor fuel performance. Currently there are 14 plants that are injecting zinc. Six of these plants are also on hydrogen water chemistry. This paper describes the effect on both Zircaloy corrosion and the cruding characteristics as a result of these changes in water chemistry. Fuel rod corrosion was found to be independent of the specific water chemistry of the plants. The corrosion behavior was the same with the additions of zinc alone or zinc plus hydrogen and well within the operating experience for fuel without either of these additions. No change was observed in the amounts of crude deposited on the fuel rods, both for the adherent and loosely held deposits. One of the effects of the zinc addition was the trend to form more of the zinc rich iron spinel in the fuel deposits rather than the hematite deposits that are predominantly formed with non additive water chemistry

  10. Safety assessment for a potential SNF repository and its implication to the proliferation resistance nuclear fuel cycle

    International Nuclear Information System (INIS)

    Hwang, Y.; Jeong, M.S.; Seo, C.S.

    2007-01-01

    decay of I-129. Also, Sr-90 and Cs-137 shares 89 % of the total decay heat. In this sense, the pyro-processing contributes significantly to reduce the risks and doses of disposal. Also, the separation of decay heat generating nuclides significantly reduces the disposal areas. Results indicate the followings: (1) To minimize the annual dose, it is important to properly manage the release mechanism of I-129 by pyro-processing. (2) To minimize the underground repository area, it is important to remove the decay heat sources, Cs-137 and Sr-90. (3) To minimize the burden to monitor a repository, it i s important to control long lived radionuclides such as I-129 and TRU's. If instantaneous release is gone, the source term controlled by congruent release will diminish quite significantly. Also, if heat sources are safety stored after removal above the ground for a certain period of time, then the underground repository area will be shrunken quite significantly also. These two factors are clear advantages of the pyro-processing to preserve environment and a future generation without significant worry over nuclear proliferation. The electro refining processes in combination with winning technologies will give additional benefits to the environment by removing and recycling long lived TRU's. To fully analyze the benefits of these two processes for environmental protection more detailed researches are needed: (1) To identify the accurate inventories from these processes. (2) To understand the dissolution mechanism of solidified wastes. Also, the new management concept development is recommended to effectively dispose of solidified wastes and possibly metal ingots and store heat generating waste above the ground. (authors)

  11. Development of new irradiation facility for BWR safety research

    International Nuclear Information System (INIS)

    Okada, Yuji; Magome, Hirokatsu; Iida, Kazuhiro; Hanawa, Hiroshi; Ohmi, Masao

    2013-01-01

    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)

  12. Residual stress analysis in BWR pressure vessel attachments

    International Nuclear Information System (INIS)

    Dexter, R.J.; Leung, C.P.; Pont, D.

    1992-06-01

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

  13. Best-estimate analysis development for BWR systems

    International Nuclear Information System (INIS)

    Sutherland, W.A.; Alamgir, M.; Kalra, S.P.; Beckner, W.D.

    1986-01-01

    The Full Integral Simulation Test (FIST) Program is a three pronged approach to the development of best-estimate analysis capability for BWR systems. An experimental program in the FIST BWR system simulator facility extends the LOCA data base and adds operational transients data. An analytical method development program with the BWR-TRAC computer program extends the modeling of BWR specific components and major interfacing systems, and improves numerical techniques to reduce computer running time. A method qualification program tests TRAC-B against experiments run in the FIST facility and extends the results to reactor system applications. With the completion and integration of these three activities, the objective of a best-estimate analysis capability has been achieved. (author)

  14. Boiling water system of nuclear power plants (BWR)

    International Nuclear Information System (INIS)

    Martias Nurdin

    1975-01-01

    About 85% of the world electric generators are light water reactors. It shows that LWR is technologically and economically competitive with other generators. The Boiling Water Reactor (BWR) is one of the two systems in the LWR group. The techniques of BWR operation in several countries, especially low and moderate power BWR, are presented. The discussion is made in relation with the interconnection problems of electric installation in developing countries, including Indonesia, where the total electric energy installation is low. The high reliability and great flexibility of the operation of a boiling water reactor for a sufficiently long period are also presented. Component standardization for BWR system is discussed to get a better technological and economical performance for further development. (author)

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

    International Nuclear Information System (INIS)

    Yadigaroglu, George

    2014-01-01

    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

  16. Level controlling system in BWR type reactors

    International Nuclear Information System (INIS)

    Joge, Toshio; Higashigawa, Yuichi; Oomori, Takashi.

    1981-01-01

    Purpose: To reasonably attain fully automatic water level control in the core of BWR type nuclear power plants. Constitution: A feedwater flow regulation valve for reactor operation and a feedwater flow regulation valve for starting are provided at the outlet of a motor-driven feedwater pump in a feedwater system, and these valves are controlled by a feedwater flow rate controller. While on the other hand, a damp valve for reactor clean up system is controlled either in ''computer'' mode or in ''manual'' mode selected by a master switch, that is, controlled from a computer or the ON-OFF switch of the master switch by way of a valve control analog memory and a turn-over switch. In this way, the water level in the nuclear reactor can be controlled in a fully automatic manner reasonably at the starting up and shutdown of the plant to thereby provide man power saving. (Seki, T.)

  17. Pressure vessel for a BWR type reactor

    International Nuclear Information System (INIS)

    Shimamoto, Yoshiharu.

    1980-01-01

    Purpose: To prevent the retention of low temperature water and also prevent the thermal fatigue of the pressure vessel by making large the curvature radius of a pressure vessel of a feed water sparger fitting portion and accelerating the mixing of low-temperature water at the feed water sparger base and in-pile hot water. Constitution: The curvature radius of the corner of the feed water sparger fitting portion in a pressure vessel is formed largely. In-pile circulating water infiltrates up to the base portion of the feed water sparger to carry outside low-temperature water at the base part, which is mixed with in-pile hot water. Accordingly, low temperature water does not stay at the base portion of the feed water sparger and generation of thermal fatigue in the pressure vessel can be prevented and the safety of the BWR type reactor can be improved. (Yoshino, Y.)

  18. Development of long operating cycle simplified BWR

    International Nuclear Information System (INIS)

    Heki, H.; Nakamaru, M.; Maruya, T.; Hiraiwa, K.; Arai, K.; Narabayash, T.; Aritomi, M.

    2002-01-01

    This paper describes an innovative plant concept for long operating cycle simplified BWR (LSBWR) In this plant concept, 1) Long operating cycle ( 3 to 15 years), 2) Simplified systems and building, 3) Factory fabrication in module are discussed. Designing long operating core is based on medium enriched U-235 with burnable poison. Simplified systems and building are realized by using natural circulation with bottom located core, internal CRD and PCV with passive system and an integrated reactor and turbine building. This LSBWR concept will have make high degree of safety by IVR (In Vessel Retention) capability, large water inventory above the core region and no PCV vent to the environment due to PCCS (Passive Containment Cooling System) and internal vent tank. Integrated building concept could realize highly modular arrangement in hull structure (ship frame structure), ease of seismic isolation capability and high applicability of standardization and factory fabrication. (authors)

  19. BWR plant advanced central control panel PODIA

    International Nuclear Information System (INIS)

    Fujii, K.; Hayakawa, H.; Ikeda, Y.; Neda, T.; Suto, O.; Takamiya, S.

    1983-01-01

    BWR plant central control panels have become more and more enlarged and complicated recently due to the magnification of the scale of a plant and the requirement to reinforce safety. So, it is important to make communication between men and the complicated central control panel smooth. Toshiba has developed an advanced central control panel, named PODIA, which uses many computers and color CRTs, and PODIA is now in the stage of application to practical plants. In this article, the writers first touch upon control functions transition in the central control room, the PODIA position concerning the world-wide trend in this technology phase and the human engineering on the design. Then they present concrete design concepts for the control board and computer system which constitute PODIA

  20. Evaluation of internal flooding in a BWR

    International Nuclear Information System (INIS)

    Shiu, K.; Papazoglou, I.A.; Sun, Y.H.; Anavim, E.; Ilberg, D.

    1985-01-01

    Flooding inside a nuclear power station is capable of concurrently disabling redundant safety systems. This paper presents the results of a recent review study performed on internally-generated floods inside a boiling water reactor (BWR) reactor building. The study evaluated the flood initiator frequency due to either maintenance or ruptures using Markovian models. A time phased event tree approach was adopted to quantify the core damage frequency based on the flood initiator frequency. It is found in the study that the contribution to the total core damage due to internal flooding events is not insignificant and is comparable to other transient contributors. The findings also indicate that the operator plays an important role in the prevention as well as the mitigation of a flooding event

  1. Seismic risk assessment of a BWR

    International Nuclear Information System (INIS)

    Wells, J.E.; Bernreuter, D.L.; Chen, J.C.; Lappa, D.A.; Chuang, T.Y.; Murray, R.C.; Johnson, J.J.

    1987-01-01

    The simplified seismic risk methodology developed in the USNRC Seismic Safety Margins Research Program (SSMRP) was demonstrated by its application to the Zion nuclear power plant (PWR). The simplified seismic risk methodology was developed to reduce the costs associated with a seismic risk analysis while providing adequate results. A detailed model of Zion, including systems analysis models (initiating events, event trees, and fault trees), SSI and structure models, and piping models, was developed and used in assessing the seismic risk of the Zion nuclear power plant (FSAR). The simplified seismic risk methodology was applied to the LaSalle County Station nuclear power plant, a BWR; to further demonstrate its applicability, and if possible, to provide a basis for comparing the seismic risk from PWRs and BWRs. (orig./HP)

  2. A BWR Safety and Operability Improvements

    International Nuclear Information System (INIS)

    Sawyer, Craig D.

    1993-01-01

    The A BWR is the culmination of 30 years of design, development and operating experience of BWRs around the world. It represents across the board improvements is safety, operation and maintenance practices (O and M), economics, radiation exposure and rad waste generation. More than ten years and $20m5 went into the design and development of its new features, and it is now under construction in Japan. This paper concentrates on the safety and operability improvements. In the safety area, more than a decade improvement in core damage frequency (CDFR) has been assessed by formal PIRA techniques, with CDFR less than 10 -6 /year. Severe accident mitigation has also been formally addressed in the design. Plant operations were simplified by incorporation of better materials, optimum use of redundancy in mechanical and electrical equipment so that on-line maintenance can be performed, by better arrangements which account for required maintenance practices, and by an advanced control room

  3. BWR stability using a reducing dynamical model

    International Nuclear Information System (INIS)

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

    1990-01-01

    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)

  4. BWR stability using a reduced dynamical model

    International Nuclear Information System (INIS)

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

    1990-01-01

    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

  5. Advanced technology for BWR operator training simulator

    International Nuclear Information System (INIS)

    Shibuya, Akira; Fujita, Eimitsu; Nakao, Toshihiko; Nakabaru, Mitsugu; Asaoka, Kouchi.

    1991-01-01

    This paper describes an operator training simulator for BWR nuclear power plants which went into service recently. The simulator is a full scope replica type simulator which faithfully replicates the control room environment of the reference plant with six main control panels and twelve auxiliary ones. In comparison with earlier simulators, the scope of the simulation is significantly extended in both width and depth. The simulation model is also refined in order to include operator training according to sympton-based emergency procedure guidelines to mitigate the results in accident cases. In particular, the core model and the calculational model of the radiation intensity distribution, if radioactive materials were released, are improved. As for simulator control capabilities by which efficient and effective training can be achieved, various advanced designs are adopted allowing easy use of the simulators. (author)

  6. Manufacturing technology and process for BWR fuel

    International Nuclear Information System (INIS)

    Kato, Shigeru

    1996-01-01

    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)

  7. Facility of BWR type nuclear power plant

    International Nuclear Information System (INIS)

    Kubo, Mitsuji

    1998-01-01

    A condensate filtering device for cleaning condensate flown from a low pressure turbine and a condensate desalting device are connected by way of a condensate pipeline. Control rod drives (CRD) are disposed to the lower portion of BWR. A CRD pump and one end of a CRD feedwater pipeline are connected in series to the upstream of CRD. The other end of the CRD feedwater pipeline is connected to a CRD water taking pipeline branched from the condensate pipeline. Water is taken to the CRD from downstream of the condensate filtering device and upstream of a connecting portion between a low pressure heater drain pipeline and the condensate pipeline. Flow of impurities leached out of the condensate desalting device to the reactor can be suppressed, and rising of temperature of CRD water by the low pressure heater drain water is prevented. In addition, flowing of dissolved oxygen to the CRD system can be suppressed. (I.N.)

  8. BWR alloy 182 stress Corrosion Cracking Experience

    International Nuclear Information System (INIS)

    Horn, R.M.; Hickling, J.

    2002-01-01

    Modern Boiling Water Reactors (BWR) have successfully operated for more than three decades. Over that time frame, different materials issues have continued to arise, leading to comprehensive efforts to understand the root cause while concurrently developing different mitigation strategies to address near-term, continued operation, as well as provide long-term paths to extended plant life. These activities have led to methods to inspect components to quantify the extent of degradation, appropriate methods of analysis to quantify structural margin, repair designs (or strategies to replace the component function) and improved materials for current and future application. The primary materials issue has been the occurrence of stress corrosion cracking (SCC). While this phenomenon has been primarily associated with austenitic stainless steel, it has also been found in nickel-base weldments used to join piping and reactor internal components to the reactor pressure vessel consistent with fabrication practices throughout the nuclear industry. The objective of this paper is to focus on the history and learning gained regarding Alloy 182 weld metal. The paper will discuss the chronology of weld metal cracking in piping components as well as in reactor internal components. The BWR industry has pro-actively developed inspection processes and procedures that have been successfully used to interrogate different locations for the existence of cracking. The recognition of the potential for cracking has also led to extensive studies to understand cracking behavior. Among other things, work has been performed to characterize crack growth rates in both oxygenated and hydrogenated environments. The latter may also be relevant to PWR systems. These data, along with the understanding of stress corrosion cracking processes, have led to extensive implementation of appropriate mitigation measures. (authors)

  9. Assessment of two BWR accident management strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Petek, M.

    1991-01-01

    A recently completed Oak Ridge effort proposes two management strategies for mitigation of the events that might occur in-vessel after the onset of significant core damage in a BWR severe accident. While the probability of such an accident is low, there may be effective yet inexpensive mitigation measures that could be implemented employing the existing plant equipment and requiring only additions to the plant emergency procedures. In this spirit, accident management strategies have been proposed for use of a borated solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and for containment flooding to maintain the core debris within the reactor vessel if injection systems cannot be restored. The proposed strategy for poisoning of the water used for vessel reflood should injection systems be restored after control blade damage has occurred has great promise, using only the existing plant equipment but employing a different chemical form for the boron poison. The dominant BWR severe accident sequence is Station Blackout and without means for mechanical stirring or heating of the storage tank, the question of being able to form the poisoned solution under accident conditions becomes of supreme importance. On the other hand, the proposed strategy for drywell flooding to cool the reactor vessel bottom head and prevent the core and structure debris from escaping to the drywell holds less promise. This strategy does, however, have potential for future plant designs in which passive methods might be employed to completely submerge the reactor vessel under severe accident conditions without the need for containment venting

  10. Seismic PRA of a BWR plant

    International Nuclear Information System (INIS)

    Nishio, Masahide; Fujimoto, Haruo

    2014-01-01

    Since the occurrence of nuclear power plant accidents in the Fukushima Daichi nuclear power station, the regulatory framework on severe accident (SA) has been discussed in Japan. The basic concept is to typify and identify the accident sequences leading to core/primary containment vessel (PCV) damage and to implement SA measures covering internal and external events extensively. As Japan is an earthquake-prone country and earthquakes and tsunami are important natural external events for nuclear safety of nuclear power plants, JNES performed the seismic probabilistic risk assessment (PRA) on a typical nuclear power plant and evaluated the dominant accident sequences leading to core/PCV damage to discuss dominant scenarios of severe accident (SA). The analytical models and the results of level-1 seismic PRA on a 1,100 MWe BWR-5 plant are shown here. Seismic PRA was performed for a typical BWR5 plant. Initiating events with large contribution to core damage frequency are the loss of all AC powers (station blackout) and the large LOCA. The top of dominant accident sequences is the simultaneous occurrence of station blackout and large LOCA. Important components to core damage frequency are electric power supply equipment. It needs to keep in mind that the results are influenced on site geologic characteristic to a greater or lesser. In the process of analysis, issues such as conservative assumptions related to damages of building or structure and success criteria for excessive LOCA are left to be resolved. These issues will be further studied including thermal hydric analysis in the future. (authors)

  11. BWR startup and shutdown activity transport control

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, S.E., E-mail: sgarcia@epri.com [Electric Power Research Inst. (EPRI), Palo Alto, California (United States); Giannelli, J.F.; Jarvis, A.J., E-mail: jgiannelli@finetech.com, E-mail: ajarvis@finetech.com [Finetech, Inc., Parsippany, New Jersey (United States)

    2010-07-01

    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 {sup o

  12. BWR Radiation Assessment and Control Program: assessment and control of BWR radiation fields. Volume 1. Executive summary

    International Nuclear Information System (INIS)

    Anstine, L.D.

    1983-05-01

    This report covers work on the BWR Radiation Assessment and Control (BRAC) Program from 1978 to 1982. The major activities during this report period were assessment of the radiation-level trends in BWRs, evaluation of the effects of forward-pumped heater drains on BWR water quality, installation and operation of a corrosion-product deposition loop in an operating BWR, and analyzation of fuel-deposit samples from two BWRs. Radiation fields were found to be controlled by cobalt-60 and to vary from as low as 50 mr/hr to as high as 800 mr/hr on the recirculation-system piping. Detailed information on BWR corrosion films and system deposits is presented in the report. Additionally, the results of an oxygen-injection experiment and recontamination monitoring studies are provided

  13. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-11-15

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

  15. Proliferation Resistant Nuclear Reactor Fuel

    International Nuclear Information System (INIS)

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

    2011-01-01

    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.

  16. Improvement technique of sensitized HAZ by GTAW cladding applied to a BWR power plant

    International Nuclear Information System (INIS)

    Tujimura, Hiroshi; Tamai, Yasumasa; Furukawa, Hideyasu; Kurosawa, Kouichi; Chiba, Isao; Nomura, Keiichi.

    1995-01-01

    A SCC(Stress Corrosion Cracking)-resistant technique, in which the sleeve installed by expansion is melted by GTAW process without filler metal with outside water cooling, was developed. The technique was applied to ICM (In-Core Monitor) housings of a BWR power plant in 1993. The ICM housings of which materials are type 304 Stainless Steels are sensitized with high tensile residual stresses by welding to the RPV (Reactor Pressure Vessel). As the result, ICM housings have potential of SCC initiation. Therefore, the improvement technique resistant to SCC was needed. The technique can improve chemical composition of the housing inside and residual stresses of the housing outside at the same time. Sensitization of the housing inner surface area is eliminated by replacing low-carbon with proper-ferrite microstructure clad. High tensile residual stresses of housing outside surface area is improved into compressive side. Compressive stresses of outside surface are induced by thermal stresses which are caused by inside cladding with outside water cooling. The clad is required to be low-carbon metal with proper ferrite and not to have the new sensitized HAZ (Heat Affected Zone) on the surface by cladding. The effect of the technique was qualified by SCC test, chemical composition check, ferrite content measurement and residual stresses measurement etc. All equipment for remote application were developed and qualified, too. The technique was successfully applied to a BWR plant after sufficient training

  17. Construction of an external electrode for determination of electrochemical corrosion potential in normal operational conditions of an BWR type reactor for hot cells

    International Nuclear Information System (INIS)

    Aguilar T, J.A.; Rivera M, H.; Hernandez C, R.

    2001-01-01

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

  18. 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

  19. Proliferation risks

    International Nuclear Information System (INIS)

    Carchon, R.

    1998-09-01

    The report gives an overview of different aspects related to safeguards of fissile materials. Existing treaties including the Non-Proliferation Treaty, and the Tlatelolco and the Rarotonga Treaties are discussed. An overview of safeguards systems for the control of fissile materials as well as the role of various authorities is given. An overall overview of proliferation risks, the physical protection of fissile materials and the trade in fissile materials is given. Finally, the status in problem countries and de facto nuclear weapon states is discussed

  20. Water level measurement uncertainty during BWR instability

    International Nuclear Information System (INIS)

    Torok, R.C.; Derbidge, T.C.; Healzer, J.M.

    1994-01-01

    This paper addresses the performance of the water-level measurement system in a boiling water reactor (BWR) during severe instability oscillations which, under some circumstances, can occur during an anticipated transient without SCRAM (ATWS). Test data from a prototypical mock-up of the water-level measurement system was used to refine and calibrate a water-level measurement system model. The model was then used to predict level measurement system response, using as boundary conditions vessel pressures calculated by ppercase RETRAN for an ATWS/instability event.The results of the study indicate that rapid pressure changes in the reactor pressure vessel which cause oscillations in downcomer water level, coupled with differences in instrument line lengths, can produce errors in the sensed water level. Using nominal parameters for the measurement system components, a severe instability transient which produced a 0.2 m peak-to-minimum water-level oscillation in the vessel downcomer was predicted to produce pressure difference equivalent to a 0.7 m level oscillation at the input to the differential pressure transmitter, 0.5 m oscillation at the output of the transmitter, and an oscillation of 0.3 m on the water-level indicator in the control room. The level measurement system error, caused by downcomer water-level oscillations and instrument line length differential, is mitigated by damping both in the differential pressure transmitter used to infer level and in the control room display instrument. ((orig.))

  1. BWR fuel clad behaviour following LOCA

    International Nuclear Information System (INIS)

    Chaudhry, S.M.; Vyas, K.N.; Dinesh Babu, R.

    1996-01-01

    Flow and pressure through the fuel coolant channel reduce rapidly following a loss of coolant accident. Due to stored energy and decay heat, fuel and cladding temperatures rise rapidly. Increase in clad temperature causes deterioration of mechanical properties of clad material. This coupled with increase of pressure inside the cladding due to accumulation of fission gases and de-pressurization of coolant causes the cladding to balloon. This phenomenon is important as it can reduce or completely block the flow passages in a fuel assembly causing reduction of emergency coolant flow. Behaviour of a BWR clad is analyzed in a design basis LOCA. Fuel and clad temperatures following a LOCA are calculated. Fission gas release and pressure is estimated using well established models. An elasto-plastic analysis of clad tube is carried out to determine plastic strains and corresponding deformations using finite-element technique. Analysis of neighbouring pins gives an estimate of flow areas available for emergency coolant flow. (author). 7 refs, 6 figs, 3 tabs

  2. Recent technology for BWR operator training simulators

    International Nuclear Information System (INIS)

    Sato, Takao; Hashimoto, Shigeo; Kato, Kanji; Mizuno, Toshiyuki; Asaoka, Koichi.

    1990-01-01

    As one of the important factors for maintaining the high capacity ratio in Japanese nuclear power stations, the contribution of excellent operators is pointed out. BWR Operation Training Center has trained many operators using two full scope simulators for operation training modeling BWRs. But in order to meet the demands of the recent increase of training needs and the upgrading of the contents, it was decided to install the third simulator, and Hitachi Ltd. received the order to construct the main part, and delivered it. This simulator obtained the good reputation as its range of simulation is wide, and the characteristics resemble very well those of the actual plants. Besides, various new designs were adopted in the control of the simulator, and its handling became very easy. Japanese nuclear power plants are operated at constant power output, and the unexpected stop is very rare, therefore the chance of operating the plants by operators is very few. Accordingly, the training using the simulators which can simulate the behavior of the plants with computers, and can freely generate abnormal phenomena has become increasingly important. The mode and positioning of the simulators for operation training, the full scope simulator BTC-3 and so on are reported. (K.I.)

  3. Method of operating BWR type reactors

    International Nuclear Information System (INIS)

    Sekimizu, Koichi

    1980-01-01

    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.)

  4. Method of operating BWR type power plants

    International Nuclear Information System (INIS)

    Koyama, Kazuaki.

    1981-01-01

    Purpose: To improve the operation efficiency of BWR type reactors by reducing the time from the start-up of the reactor to the start-up of the turbine and electrical generator, as well as decrease the pressure difference in each of the sections of the pressure vessel to thereby extend its life span. Method: The operation comprises switching the nuclear reactor from the shutdown mode to the start-up mode, increasing the reactor power to a predetermined level lower than a rated power while maintaining the reactor pressure to a predetermined level lower than a rated pressure, starting up a turbine and an electrical generator in the state of the predetermined reactor pressure and the reactor power to connect the electrical generator to the power transmission system and, thereafter, increasing the reactor pressure and the reactor power to the predetermined rated pressure and rated power respectively. This can shorten the time from the start-up of the reactor to the start of the power transmission system, whereby the operation efficiency of the power plant can be improved. (Moriyama, K.)

  5. Experience on a BWR plant diagnosis system

    International Nuclear Information System (INIS)

    Tanabe, A.; Kawai, K.; Hashimoto, Y.

    1981-01-01

    It is important to watch plant dynamics and equipment condition for avoiding a big transient or avoiding damage to a system by equipment failure. After the TMI accident the necessity of a diagnosis system has been recognized and such development activities have become of primary importance in many organizations. A diagnosis system has two kinds of function. One is the early detection of an anomaly before detection by a conventional instrumentation system. The other is appropriate instruction after alarm or scram has occurred. The authors have been developing the former system by a noise analysis technique and a feasibility study has been undertaken in recent years as a joint research programme of several electric power companies and the Toshiba Corporation. A prototype diagnosis system has been installed on a BWR plant in Japan. This diagnosis system concerns reactor core, jet pumps and three main control systems. Many data from normal operation have been accumulated using this system and a variation pattern of normal noise data is clarified. On this basis, anomally detection criteria have been determined using statistical decision theory. It is confirmed that this system performance is satisfactory, and that the system will be of great use for surveillance of core and control systems without artificial disturbances. (author)

  6. Power control system in BWR type reactors

    International Nuclear Information System (INIS)

    Nishizawa, Yasuo.

    1980-01-01

    Purpose: To control the reactor power so that the power distribution can satisfy the limiting conditions, by regulating the reactor core flow rate while monitoring the power distribution in the reactor core of a BWR type reactor. Constitution: A power distribution monitor determines the power distribution for the entire reactor core based on the data for neutron flux, reactor core thermal power, reactor core flow rate and control rod pattern from the reactor and calculates the linear power density distribution. A power up ratio computing device computes the current linear power density increase ratio. An aimed power up ratio is determined by converting the electrical power up ratio transferred from a load demand input device into the reactor core thermal power up ratio. The present reactor core thermal power up ratio is subtracted from the limiting power up ratio and the difference is sent to an operation amount indicator and the reactor core flow rate is changed in a reactor core flow rate regulator, by which the reactor power is controlled. (Moriyama, K.)

  7. Fuel assemblies for BWR type reactors

    International Nuclear Information System (INIS)

    Ishizuka, Takao.

    1981-01-01

    Purpose: To enable effective failed fuel detection by the provision of water rod formed with a connecting section connected to a warmed water feed pipe of a sipping device at the lower portion and with a warmed water jetting port in the lower portion in a fuel assembly of a BWR type reactor to thereby carry out rapid sipping. Constitution: Fuel rods and water rods are contained in the channel box of a fuel assembly, and the water rod is provided at its upper portion with a connecting section connected to the warmed water feed pipe of the sipping device and formed at its lower portion with a warmed water jetting port for jetting warmed water fed from the warmed water feed pipe. Upon detection of failed fuels, the reactor operation is shut down and the reactor core is immersed in water. The cover for the reactor container is removed and the cap of the sipping device is inserted to connect the warmed water feed pipe to the connecting section of the water rod. Then, warmed water is fed to the water rod and jetted out from the warmed water jetting port to cause convection and unify the water of the channel box in a short time. Thereafter, specimen is sampled and analyzed for the detection of failed fuels. (Moriyama, K.)

  8. Primary coolant system of BWR type reactor

    International Nuclear Information System (INIS)

    Ibe, Hidefumi; Takahashi, Masanori; Aoki, Yasuko

    1997-01-01

    The present invention provides a water quality control system for preventing corrosion and for extending working life of structural materials of a BWR-type reactor. Namely, a sensor group 1 and a sensor group 2 are disposed at different positions such as in a feedwater system, a recycling system, main steam pipes, and a pressure vessel, respectively. Each sensor group can record and generate alarms independently. The sensor group 1 for usual monitoring is connected to a calculation device by way of a switch to confirm that the monitored values are within a proper range by the injection of a water quality moderating agent. The sensor group 2 is caused to stand alone or connected with the calculation device by way of a switch optionally. When abnormality should occur in the sensor group 1, the sensor group 2 determines the limit for the increase/decrease of controlling amount of the moderating agent at a portion where the conditions are changed to the most severe direction by using data base. The moderating agent is injected and controlled based on the controlling amount. The system of the present invention can optionally cope with a new sensor and determination for new water quality standards. Then the evaluation/control accuracy of the entire system can be improved while covering up the errors of each sensor. (I.S.)

  9. Feedwater control system in BWR type reactor

    International Nuclear Information System (INIS)

    Tanji, Jun-ichi; Oomori, Takashi.

    1980-01-01

    Purpose: To improve the water level control performance in BWR type reactor by regulating the water level set to the reactor depending on the rate of change in the recycling amount of coolant to thereby control the fluctuations in the water level resulted in the reactor within an aimed range even upon significant fluctuations in the recycling flow rate. Constitution: The recycling flow rate of coolant in the reactor is detected and the rate of its change with time is computed to form a rate of change signal. The rate of change signal is inputted to a reactor level setter to amend the actual reactor water level demand signal and regulate the water level set to the reactor water depending on the rate of change in the recycling flow rate. Such a regulation method for the set water level enables to control the water level fluctuation resulted in the reactor within the aimed range even upon the significant fluctuation in the recycling flow rate and improve the water level control performance of the reactor, whereby the operationability for the reactor is improved to enhance the operation rate. (Moriyama, K.)

  10. Power generator in BWR type reactors

    International Nuclear Information System (INIS)

    Yoshida, Kenji.

    1984-01-01

    Purpose: To enable to perform stable and dynamic conditioning operation for nuclear fuels in BWR type reactors. Constitution: The conditioning operation for the nuclear fuels is performed by varying the reactor core thermal power in a predetermined pattern by changing the predetermined power changing pattern of generator power, the rising rate of the reactor core thermal power and the upper limit for the rising power of the reactor core thermal power are calculated and the power pattern for the generator is corrected by a power conditioning device such that the upper limit for the thermal power rising rate and the upper limit for the thermal power rising rate are at the predetermined levels. Thus, when the relation between the reactor core thermal power and the generator electrical power is fluctuated, the fluctuation is detected based on the variation in the thermal power rising rate and the limit value for the thermal power rising rate, and the correction is made to the generator power changing pattern so that these values take the predetermined values to thereby perform the stable conditioning operation for the nuclear fuels. (Moriyama, K.)

  11. Burnup credit feasibility for BWR spent fuel shipments

    International Nuclear Information System (INIS)

    Broadhead, B.L.

    1990-01-01

    Considerable interest in the allowance of reactivity credit for the exposure history of power reactor fuel currently exists. This ''burnup credit'' issue has the potential to greatly reduce risk and cost when applied to the design and certification of spent of fuel casks used for transportation and storage. Analyses 1 have shown the feasibility estimated the risk and economic incentives for allowing burnup credit in pressurized water reactor (PWR) spent fuel shipping cask applications. This paper summarizes the extension of the previous PWR feasibility assessments to boiling water reactor (BWR) fuel. As with the PWR analysis, the purpose was not verification of burnup credit (see ref. 2 for ongoing work in this area) but a reasonable assessment of the feasibility and potential gains from its use in BWR applications. This feasibility analysis aims to apply simple methods that adequately characterize the time-dependent isotopic compositions of typical BWR fuel. An initial analysis objective was to identify a simple and reliable method for characterizing BWR spent fuel. The method includes characterization of a typical pin-cell spectrum, using a one-dimensional (1-D) model of a BWR assembly. The calculated spectrum allows burnup-dependent few-group material constants to be generated. Point depletion methods were then used to obtain the time-varying characteristics of the fuel. These simple methods were validated, where practical, with multidimensional methods. 6 refs., 1 tab

  12. Application of gadolinia credit to cask transportation of BWR-STEP3 SFAs

    International Nuclear Information System (INIS)

    Kikuchi, Tsukasa; Mitsuhashi, Ishi; Ito, Dai-ichiro; Nakamura, Yu

    2003-01-01

    Instead of the fresh-fuel assumption, the application of gadolinia credit to cask transportation of BWR SFAs is studied. Its efficacy for BWR-STEP2 SFAs had already been estimated. This paper reports on the application of gadolinia credit to cask transportation of BWR-STEP3 SFAs. (author)

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

    International Nuclear Information System (INIS)

    Linde, A. van der.

    1989-04-01

    Comparative testing of UO 2 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 UO 2 in LWR fuel rods offers better resistance against stress corrosion cracking of the cladding, but that under fast, single step, power ramping conditions pellet UO 2 in LWR fuel rods has a better resistance against hoop stress failure of the cladding. 128 figs., 36 refs., 19 tabs

  14. Seismic risk assessment of a BWR: status report

    International Nuclear Information System (INIS)

    Chuang, T.Y.; Bernreuter, D.L.; Wells, J.E.; Johnson, J.J.

    1985-02-01

    The seismic risk methodology developed in the US NRC Seismic Safety Margins Research Program (SSMRP) was demonstrated by its application to the Zion nuclear power plant, a pressurized water reactor (PWR). A detailed model of Zion, including systems analysis models (initiating events, event trees, and fault trees), SSI and structure models, and piping models was developed and analyzed. The SSMRP methodology can equally be applied to a boiling water reactor (BWR). To demonstrate its applicability, to identify fundamental differences in seismic risk between a PWR and a BWR, and to provide a basis of comparison of seismic risk between a PWR and a BWR when analyzed with comparable methodology and assumptions, a seismic risk analysis is being performed on the LaSalle County Station nuclear power plant

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

    International Nuclear Information System (INIS)

    Smith, S.K.; Lehnert, D.F.; Locke, R.K.

    1991-01-01

    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)

  16. Simulation of decreasing reactor power level with BWR simulator

    International Nuclear Information System (INIS)

    Suwoto; Zuhair; Rivai, Abu Khalid

    2002-01-01

    Study on characteristic of BWR using Desktop PC Based Simulator Program was analysed. This simulator is more efficient and cheaper for analyzing of characteristic and dynamic respond than full scope simulator for decreasing power level of BW. Dynamic responses of BWR reactor was investigated during the power level reduction from 100% FP (Full Power) which is 3926 MWth to 0% FP with 25% steps and 1 % FP/sec rate. The overall results for core flow rate, reactor steam flow, feed-water flow and turbine-generator power show tendency proportional to reduction of reactor power. This results show that reactor power control in BWR could be done by control of re-circulation flow that alter the density of water used as coolant and moderator. Decreasing the re-circulation flow rate will decrease void density which has negative reactivity and also affect the position of control rods

  17. An overview of the BWR ECCS strainer blockage issues

    International Nuclear Information System (INIS)

    Serkiz, A.W.; Marshall, M.L. Jr.; Elliott, R.

    1996-01-01

    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

  18. AREVA 10x10 BWR fuel experience feedback and on going upgrading

    International Nuclear Information System (INIS)

    Lippert, Hans Joachim; Rentmeister, Thomas; Garner, Norman; Tandy, Jay; Mollard, Pierre

    2008-01-01

    Established with engineering and manufacturing operations in the US and Europe, AREVA NP has been and is supplying nuclear fuel assemblies and associated core components to boiling water reactors worldwide, representing today more than 63 000 fuel assemblies. The evolution of BWR fuel rod arrays from early 6x6 designs to the 10x10 designs first introduced in the mid 1990's yielded significant improvements in thermal mechanical operating limits, critical power level, cold shutdown margin, discharge burnup, as well as other key operational capabilities. Since first delivered in 1992, ATRIUM T M 1 0 fuel assemblies have now been supplied to a total of 32 BWR plants in the US, Europe, and Asia resulting in an operating experience over 20 000 fuel assemblies. This article presents in detail the operational experience consolidated by these more than 20 000 ATRIUM T M 1 0 BWR assemblies already supplied to utilities. Within the different 10x10 fuel assemblies available, the Fuel Assembly design is chosen and tailored to the operating strategies of each reactor. Among them, the latest versions of ATRIUM T M a re ATRIUM T M 1 0XP and ATRIUM T M 1 0XM fuel assemblies which have been delivered to several utilities worldwide. The article details key aspects of ATRIUM T M 1 0 fuel assemblies in terms of reliability and performance. Special attention is paid to key proven features, ULTRAFLOW T M s pacer grids, the use of part length fuel rods (PLFRs) and their geometrical optimization, water channel and load chain, upgraded features available for inclusion with most advanced designs. Regular upgrading of the product has been made possible thanks to a continuous improvement process with the aim of further upgrading BWR fuel assembly performance and reliability. Regarding thermal mechanical behavior of fuel rods, chromia (Cr2O3) doped fuel pellets, described in Reference 1, well illustrate this improvement strategy to reduce fission gas release, increase power thresholds for PCI

  19. Nuclear proliferation

    International Nuclear Information System (INIS)

    Stencel, S.

    1978-01-01

    The terms and reactions to President Carter's nuclear policy, culminating in the 1978 Nuclear Non-Proliferation Act, are reviewed and analyzed. The new law increases restrictions on nuclear exports, encourages continued use of light water reactors in preference to plutonium-fueled reactors, and emphasizes technical solutions to proliferation problems. Critics of the law point out that it will hurt U.S. trade unfairly, that other countries do not have as many fuel options as the U.S. has, and that nuclear sales have as many political and economic as technical solutions. Compromise areas include new international safety guidelines, the possibility of an international nuclear fuel bank, and a willingness to consider each case on its merits. 21 references

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

    International Nuclear Information System (INIS)

    Kohsaka, Atsuo; Akimoto, Masayuki; Asahi, Yoshiro; Abe, Kiyoharu; Muramatsu, Ken; Araya, Fumimasa; Sato, Kazuo

    1982-12-01

    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)

  1. The BWR VIP role in license renewal

    International Nuclear Information System (INIS)

    Dyle, R.

    2001-01-01

    The full text follows. The Boiling Water Reactor Vessel and Internals Project (BWRVIP) was started in response to an increase in the occurrence of intergranular stress corrosion cracking (IGSCC) in BWR reactor pressure vessel (RPV) internal components. The BWRVIP first evaluated the internals to determine which components were necessary to assure safe operation. An assessment of the relative significance of the internals was then performed to establish the priority in which components would be evaluated. Once this was determined, each safety-related component was evaluated to determine what, if any, inspections or tests were necessary to assure component integrity. Although IGSCC was the initial degradation mechanism of concern for RPV internals, the individual component evaluations considered all known modes of failure (fatigue, stress corrosion cracking, neutron embrittlement, etc). The component evaluations also considered all potential failure locations and the susceptibility to degradation. Once the potential failure locations and mechanisms were identified, the BWRVIP developed inspection criteria to assess component condition. The BWRVIP also developed flaw evaluation methodologies that could be used to determine the integrity and remaining life of each component. All of this information was consolidated into a document called an Inspection and Flaw Evaluation (IE) Guideline for each component. At the same time the BWRVIP was developing its program to assure internals integrity, utilities began to seriously consider measures necessary to extend the life of the plants. In the United States, the USNRC promulgated rules to allow the renewal of a license to allow plant operation for an additional 20 years. One aspect of the rule was that management of age-related degradation in the renewal period must be performed. The timing of this ''license renewal rule'' was advantageous in that it allowed the BWRVIP to address the requirements of the rule in the development

  2. Peach Bottom transient analysis with BWR TRACB02

    International Nuclear Information System (INIS)

    Alamgir, M.; Sutherland, W.A.

    1984-01-01

    TRAC calculations have been performed for a Turbine Trip transient (TT1) in the Peach Bottom BWR power plant. This study is a part of the qualification of the BWR-TRAC code. The simulation is aimed at reproducing the observed thermal hydraulic behavior in a pressurization transient. Measured core power is an input to the calculation. Comparison with data show the code reasonably well predicts the generation and propagation of the pressure waves in the main steam line and associated pressurization of the reactor vessel following the closure of the turbine stop valve

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

    International Nuclear Information System (INIS)

    Yokobayashi, Masao; Yoshida, Kazuo; Fujiki, Kazuo

    1989-06-01

    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)

  4. Limerick BWR turbine control and protection system upgrade success

    International Nuclear Information System (INIS)

    Tang, C.K.; Pietryka, T.S.; Federico, P.A.; Williams, J.C.

    2015-01-01

    Westinghouse and Exelon have successfully implemented a digital electro-hydraulic control (DEHC) at Limerick BWR Unit 1 Station to perform the turbine control, protection and reactor pressure functions. The DEHC replaces analog controls and addressed system performance, obsolescence and reliability. This was a first-of-a-kind application for control and protection of the main turbine and BWR pressure control for the distributed control system utilized. The demolition of analog equipment, main control room and front standard modifications, and acceptance testing were completed on schedule during the normal 2014 outage. Key aspects of the project that facilitated this success will be discussed and presented. (author)

  5. Limerick BWR turbine control and protection system upgrade success

    Energy Technology Data Exchange (ETDEWEB)

    Tang, C.K.; Pietryka, T.S.; Federico, P.A., E-mail: tangck@westinghouse.com, E-mail: pietryt@westinghouse, E-mail: federipa@westinghouse.com [Westinghouse Electric Company, LLC, Cranberry Township, PA (United States); Williams, J.C., E-mail: Jonathan.Williams@exeloncorp.com [Exelon Nuclear, Warrenville, IL (United States)

    2015-07-01

    Westinghouse and Exelon have successfully implemented a digital electro-hydraulic control (DEHC) at Limerick BWR Unit 1 Station to perform the turbine control, protection and reactor pressure functions. The DEHC replaces analog controls and addressed system performance, obsolescence and reliability. This was a first-of-a-kind application for control and protection of the main turbine and BWR pressure control for the distributed control system utilized. The demolition of analog equipment, main control room and front standard modifications, and acceptance testing were completed on schedule during the normal 2014 outage. Key aspects of the project that facilitated this success will be discussed and presented. (author)

  6. BWR shutdown analyzer using artificial intelligence (AI) techniques

    International Nuclear Information System (INIS)

    Cain, D.G.

    1986-01-01

    A prototype alarm system for detecting abnormal reactor shutdowns based on artificial intelligence technology is described. The system incorporates knowledge about Boiling Water Reactor (BWR) plant design and component behavior, as well as knowledge required to distinguish normal, abnormal, and ATWS accident conditions. The system was developed using a software tool environment for creating knowledge-based applications on a LISP machine. To facilitate prototype implementation and evaluation, a casual simulation of BWR shutdown sequences was developed and interfaced with the alarm system. An intelligent graphics interface for execution and control is described. System performance considerations and general observations relating to artificial intelligence application to nuclear power plant problems are provided

  7. Modern technology applied in the advanced BWR (ABWR)

    International Nuclear Information System (INIS)

    Hucik, S.A.

    1988-01-01

    The advanced boiling water reactor (ABWR) represents the next generation of light water reactors (LWR) to be introduced into commercial operation in the 1990's. The ABWR is the result of the continuing evolution of the BWR, incorporating state-of-the-art technology and improvements based on worldwide experience, and extensive design and test and development programs. This paper discusses how the ABWR development objective focused on an optimized selection of advanced technologies and proven BWR technologies. A technical evaluation of the ABWR shows its superiority in terms of performance characteristics and economics relative to current LWR designs

  8. 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)

  9. The different expression of TRPM7 and MagT1 impacts on the proliferation of colon carcinoma cells sensitive or resistant to doxorubicin

    OpenAIRE

    Cazzaniga, Alessandra; Moscheni, Claudia; Trapani, Valentina; Wolf, Federica I.; Farruggia, Giovanna; Sargenti, Azzurra; Iotti, Stefano; Maier, Jeanette A. M.; Castiglioni, Sara

    2017-01-01

    The processes leading to anticancer drug resistance are not completely unraveled. To get insights into the underlying mechanisms, we compared colon carcinoma cells sensitive to doxorubicin with their resistant counterpart. We found that resistant cells are growth retarded, and show staminal and ultrastructural features profoundly different from sensitive cells. The resistant phenotype is accompanied by the upregulation of the magnesium transporter MagT1 and the downregulation of the ion chann...

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

    International Nuclear Information System (INIS)

    Zhang, Zi-wei; Guo, Rui-wei; Lv, Jin-lin; Wang, Xian-mei; Ye, Jin-shan; Lu, Ni-hong; Liang, Xing; Yang, Li-xia

    2017-01-01

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

  11. Evaluation of PWR and BWR pin cell benchmark results

    International Nuclear Information System (INIS)

    Pijlgroms, B.J.; Gruppelaar, H.; Janssen, A.J.; Hoogenboom, J.E.; Leege, P.F.A. de; Voet, J. van der; Verhagen, F.C.M.

    1991-12-01

    Benchmark results of the Dutch PINK working group on PWR and BWR pin cell calculational benchmark as defined by EPRI are presented and evaluated. The observed discrepancies are problem dependent: a part of the results is satisfactory, some other results require further analysis. A brief overview is given of the different code packages used in this analysis. (author). 14 refs., 9 figs., 30 tabs

  12. Evaluation of PWR and BWR pin cell benchmark results

    Energy Technology Data Exchange (ETDEWEB)

    Pijlgroms, B.J.; Gruppelaar, H.; Janssen, A.J. (Netherlands Energy Research Foundation (ECN), Petten (Netherlands)); Hoogenboom, J.E.; Leege, P.F.A. de (Interuniversitair Reactor Inst., Delft (Netherlands)); Voet, J. van der (Gemeenschappelijke Kernenergiecentrale Nederland NV, Dodewaard (Netherlands)); Verhagen, F.C.M. (Keuring van Electrotechnische Materialen NV, Arnhem (Netherlands))

    1991-12-01

    Benchmark results of the Dutch PINK working group on PWR and BWR pin cell calculational benchmark as defined by EPRI are presented and evaluated. The observed discrepancies are problem dependent: a part of the results is satisfactory, some other results require further analysis. A brief overview is given of the different code packages used in this analysis. (author). 14 refs., 9 figs., 30 tabs.

  13. Requests on domestic nuclear data library from BWR design

    International Nuclear Information System (INIS)

    Maruyama, Hiromi

    2003-01-01

    Requests on the domestic nuclear data library JENDL and activities of the Nuclear Data Center have been presented from the perspective of BWR design and design code development. The requests include a standard multi-group cross section library, technical supports, and clarification of advantage of JENDL as well as requests from physical aspects. (author)

  14. BWR ATWS mitigation by Fine Motion Control Rod

    International Nuclear Information System (INIS)

    Rohatgi, U.S.; Cheng, H.S.; Khan, H.; Mallen, A.; Diamond, D.

    1994-01-01

    Two main methods of ATWS mitigation in a SBWR are: fine Motion control Rods (FMCRD) and Boron injection via the Standby Liquid control System (SLCS). This study has demonstrated that the use of FMCRD along with feedwater runback mitigated the conditions due to reactivity insertion and possible ATWS in a BWR which is similar to SBWR

  15. Material operating behaviour of ABB BWR control rods

    International Nuclear Information System (INIS)

    Rebensdorff, B.; Bart, G.

    2000-01-01

    The BWR control rods made by ABB use boron carbide (B 4 C 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 B 4 C 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)

  16. Detection of failed fuel rods in shrouded BWR fuel assemblies

    International Nuclear Information System (INIS)

    Baero, G.; Boehm, W.; Goor, B.; Donnelly, T.

    1988-01-01

    A manipulator and an ultrasonic testing (UT) technique were developed to identify defective fuel rods in shrouded BWR fuel assemblies. The manipulator drives a UT probe axially through the bottom tie plate into the water channels between the fuel rods. The rotating UT probe locates defective fuel rods by ingressed water which attenuates the UT-signal. (author)

  17. Transmutation of minor actinide using thorium fueled BWR core

    International Nuclear Information System (INIS)

    Susilo, Jati

    2002-01-01

    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 6 t h 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

  18. Physical model of nonlinear noise with application to BWR stability

    International Nuclear Information System (INIS)

    March-Leuba, J.; Perez, R.B.

    1983-01-01

    Within the framework of the present model it is shown that the BWR reactor cannot be unstable in the linear sense, but rather it executes limited power oscillations of a magnitude that depends on the operating conditions. The onset of these oscillations can be diagnosed by the decrease in stochasticity in the power traces and by the appearance of harmonics in the PSD

  19. Power plant design: ESBWR - the latest passive BWR

    International Nuclear Information System (INIS)

    Arnold, H.; Yadigaroglu, G.; Stoop, P.C.

    1997-01-01

    When General Electric said it would end development of its 670 MWe SBWR (Simplified Boiling Water Reactor), it was not quite the end of the story. Also on the drawing board at the time was the larger ESBWR (standing for either European or Economic Simplified BWR) whose goal was to provide the improved economic performance that the SBWR could not. (UK)

  20. 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)

  1. Advanced methods for BWR transient and stability analysis

    International Nuclear Information System (INIS)

    Schmidt, A.; Wehle, F.; Opel, S.; Velten, R.

    2008-01-01

    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)

  2. Decay ratio studies in BWR and PWR using wavelet

    International Nuclear Information System (INIS)

    Ciftcioglu, Oe.

    1996-10-01

    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.)

  3. Operator training simulator for BWR nuclear power plant

    International Nuclear Information System (INIS)

    Watanabe, Tadasu

    1988-01-01

    For the operation management of nuclear power stations with high reliability and safety, the role played by operators is very important. The effort of improving the man-machine interface in the central control rooms of nuclear power stations is energetically advanced, but the importance of the role of operators does not change. For the training of the operators of nuclear power stations, simulators have been used from the early stage. As the simulator facilities for operator training, there are the full scope simulator simulating faithfully the central control room of an actual plant and the small simulator mainly aiming at learning the plant functions. For BWR nuclear power stations, two full scope simulators are installed in the BWR Operator Training Center, and the training has been carried out since 1974. The plant function learning simulators have been installed in respective electric power companies as the education and training facilities in the companies. The role of simulators in operator training, the BTC No.1 simulator of a BWR-4 of 780 MWe and the BTC No.2 simulator of a BWR-5 of 1,100 MWe, plant function learning simulators, and the design of the BTC No.2 simulator and plant function learning simulators are reported. (K.I.)

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

    International Nuclear Information System (INIS)

    Ortiz-Villafuerte, Javier; Castillo-Duran, Rogelio; Palacios-Hernandez, Javier C.

    2011-01-01

    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

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

    International Nuclear Information System (INIS)

    Takahashi, Shiro; Okuyama, Keita; Tamura, Akinori

    2009-01-01

    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)

  6. Metallurgical factors that contribute to cracking in BWR piping

    International Nuclear Information System (INIS)

    Weeks, J.R.

    1975-01-01

    During the fall of 1974 and early winter of 1975, cracks have been discovered in the 4 in. bypass lines of several Boiling Water Reactors (BWR's) in the United States. Further, similar cracks were discovered at two BWR's in Japan during the same period. More recently, cracks have been discovered in the core spray piping and in a furnace-sensitized ''safe end'' and adjacent ''dutchman'' at the Dresden Nuclear Power Station, Unit No. 2. Although inspections at all other U.S. BWR's have not disclosed further instances of cracking in core spray piping, leaking cracks have been found in the core spray piping of two BWR's overseas. Metallurgical examinations of these cracks are not yet complete. The following observations have been made to date. All cracks (except those in the furnace-sensitized safe end and dutchman) occurred in seamless type 304 stainless steel piping or in elbows fabricated from such piping, in the outer heat affected zone of either field or shop welds, in lines isolated from the main primary coolant flow during full power operation, except for the not yet examined cracks in the Monticello bypass lines. The cracks are exclusively intergranular, and occur in metal that has been lightly sensitized by the welding process, with only intermittent grain boundary carbides. They developed in the areas of peak axial residual stresses from welding rather than in the most heavily sensitized areas. No fatigue striations have been found on the fracture surfaces. The evidence received to date strongly indicates that these cracks were caused by intergranular stress corrosion of weld-sensitized stainless steel by BWR water containing greater than 0.2 ppM oxygen. The possible role of fatigue or alternating stresses in this corrosion is not clear. Further, not all the cracks detected to date necessarily have occurred by the same mechanism

  7. Preventing proliferation

    International Nuclear Information System (INIS)

    Rathjens, G.

    1983-01-01

    Challenging the argument that nuclear proliferation may be stabilizing, the author cites the Israeli attack on Iraq as evidence that emergent nuclear states may be moved to attack their adversaries.The larger the number of decision makers who can unleash nuclear weapons, the greater the liklihood of their use. Several reasons are cited for nations to seek nuclear capability: the accelerated spread of technology, the deterioration in US-Soviet relations and strength relative to their nations, the high cost of conventional weapons, and a loss of confidence in the international safeguards system. The imposition of constraints, such as a Comprehensive Test Ban Treaty, on nuclear trade and technology transfer are likely to have a high cost. The US position on this issue is likely to be determined by the balance of power with the Soviet Union. 5 references

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

    International Nuclear Information System (INIS)

    Watanabe, Masato; Tuji, Kenji; Ito, Keisuke

    2010-01-01

    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)

  9. Animation of Antimicrobial Resistance

    Medline Plus

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

  10. Animation of Antimicrobial Resistance

    Medline Plus

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

  11. Animation of Antimicrobial Resistance

    Medline Plus

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

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

    International Nuclear Information System (INIS)

    Peng, Honghai; Du, Bin; Jiang, Huili; Gao, Jun

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-01-22

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

  14. LAPUR5 BWR stability analysis in Kuosheng nuclear power plant

    International Nuclear Information System (INIS)

    Kunlung Wu; Chunkuan Shih; Wang, J.R.; Kao, L.S.

    2005-01-01

    Full text of publication follows: Unstable oscillation of a nuclear power reactor core is one of the main reasons that causes minor core damage. Stability analysis needs to be performed to predict the potential problem as early as possible and to prevent core instability events from happening. Nuclear Regulatory Commission (NRC) requests all BWR licensees to examine each core reload and to impose operating limitations, as appropriate, to ensure compliance with GDC 10 and 12. GDC 10 requires that the reactor core be designed with appropriate margin to assure that specified acceptable fuel design limits will not be exceeded during any condition of normal operation, including anticipated operational occurrences. GDC 12 requires assurance that power oscillations which can result in conditions exceeding specified acceptable fuel design limits are either not possible or can be reliably and readily detected and suppressed. Therefore, the core instability is directly related to the fuel design limits. The core and channel DR (decay ratio) calculation are commonly performed to determine system's stability when new fuel designs are introduced in the core. In order to establish the independent analysis technology for BWR licensees and verifications, the Institute of Nuclear Energy Research (INER) has obtained agreement from NRC and implemented the 'Methodology and Procedure for Calculation of Core and Channel Decay Ratios with LAPUR', which was developed by the IBERINCO in 2001. LAPUR5 uses a multi-nodal description of the neutron dynamics, together with a distributed parameter model of the core thermal hydrodynamics to produce a space-dependent representation of the dynamics of a BWR in the frequency domain for small perturbations around a steady state condition. From the output of LAPUR5, the following results are obtained: global core decay ratio, out-of phase core decay ratio, and channel decay ratio. They are key parameters in the determination of BWR core stability

  15. 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)

    2008-11-01

    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

  16. Impact of advanced BWR core physics method on BWR core monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Moon, H; Wells, A [Siemens Power Corporation, Richland (United States)

    2000-07-01

    Siemens Power Corporation recently initiated development of POWERPLEX{sup TM}-III for delivery to the Grand Gulf Nuclear Power Station. The main change introduced in POWERPLEX{sup TM}-III as compared to its predecessor POWERPLEX{sup TM}-II is the incorporation of the advances BWR core simulator MICROBURN-B2. A number of issues were identified and evaluated relating to the implementation of MICROBURN-B2 and its impact on core monitoring. MICROBURN-B2 demands about three to five times more memory and two to three times more computing time than its predecessor MICROBURN-B in POWERPLEX {sup TM}-II. POWERPLEX{sup TM}-III will improve thermal margin prediction accuracy and provide more accurate plant operating conditions to operators than POWERPLEX{sup TM}-II due to its improved accuracy in predicted TIP values and critical k-effective. The most significant advantage of POWERPLEX{sup TM}-III is its capability to monitor a relaxed rod sequence exchange operation. (authors)

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

    Science.gov (United States)

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

  18. Fissile material disposition and proliferation risk

    Energy Technology Data Exchange (ETDEWEB)

    Dreicer, J.S.; Rutherford, D.A. [Los Alamos National Lab., NM (United States). NIS Div.

    1996-05-01

    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.

  19. Fissile material disposition and proliferation risk

    International Nuclear Information System (INIS)

    Dreicer, J.S.; Rutherford, D.A.

    1996-01-01

    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

  20. Comparative analysis of mechanical characteristics of solidified concentrates from BWR system using Yugoslav and Italian cements

    International Nuclear Information System (INIS)

    Plecas, I.; Peric, A.; Drljaca, J.; Kostadinovic, A.

    1987-01-01

    In this paper, properties of Italian and Yugoslav cement mixture with BWR evaporation concentrates were compared, research was held upon fifteen samples, according to the adequate formulations. Samples were made in standard cube form, side 10 cm. Functional relationship between decreasing the compressive strength and amount of incorporated BWR concentrate cement mixture was developed. The results of research showed nearly the same mechanical properties of solidified BWR concentrate with Italian and Yugoslav cements. (author)

  1. Kinematics of two-phase mixture level motion in BWR pressure vessels

    International Nuclear Information System (INIS)

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

    1985-01-01

    A model is presented for predicting two-phase mixture level elevations in BWR systems. The model accounts for the particular geometry and conditions in a BWR system during Small-Break Loss of Coolant Accidents. The model presented here is particularly suitable for efficient, high-speed simulations on small minicomputers. The model has been implemented and tested. Results are shown from BWR ATWS simulations

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

    Science.gov (United States)

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

    2014-03-01

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

  3. Construction techniques and management methods for BWR plants

    International Nuclear Information System (INIS)

    Shimizu, Yohji; Tateishi, Mizuo; Hayashi, Yoshishige

    1989-01-01

    Toshiba is constantly striving for safer and more efficient plant construction to realize high-quality BWR plants within a short construction period. To achieve these aims, Toshiba has developed and improved a large number of construction techniques and construction management methods. In the area of installation, various techniques have been applied such as the modularization of piping and equipment, shop installation of reactor internals, etc. Further, installation management has been upgraded by the use of pre-installation review programs, the development of installation control systems, etc. For commissioning, improvements in commissioning management have been achieved through the use of computer systems, and testing methods have also been upgraded by the development of computer systems for the recording and analysis of test data and the automatic adjustment of controllers in the main control system of the BWR. This paper outlines these construction techniques and management methods. (author)

  4. Specifications of the BWR simulator for HAMMLAB 2000

    International Nuclear Information System (INIS)

    Grini, Rolf-Einar; Miettinen, Jaakko; Nurmilaukas, Pekka; Raussi; Pekka; Saarni, Ray; Stokke; Egil; Soerensen, Aimar; Tiihonen, Olli

    1998-02-01

    The Boiling Water Reactor (BWR) simulator for HAMMLAB 2000 will be a model of the Swedish plant Forsmark-3. This report gives the specifications of the BWR simulator. The bulk of the report is a copy of the relevant addendum to the contract with the developer, and to the contract with the group of utilities and with ABB Atom. After a general overview, each plant system is described one after the other (using the reference plant system coding), and the simulation of each system is specified. Even the systems that shall not be simulated are included; in those cases the specification is: It is not required that ... is simulated. A list of malfunctions is given, as well as a list of validation transients. Finally the operator interface is specified. (author)

  5. BWR Fuel Assemblies Physics Analysis Utilizing 3D MCNP Modeling

    International Nuclear Information System (INIS)

    Chiang, Ren-Tai; Williams, John B.; Folk, Ken S.

    2008-01-01

    MCNP is used to model a partially controlled BWR fresh fuel four assemblies (2x2) system for better understanding BWR fuel behavior and for benchmarking production codes. The impact of the GE14 plenum regions on axial power distribution is observed by comparing against the GE13 axial power distribution, in which the GE14 relative power is lower than the GE13 relative power at the 15. node and at the 16. node due to presence of the plenum regions in GE14 fuel in these two nodes. The segmented rod power distribution study indicates that the azimuthally dependent power distribution is very significant for the fuel rods next to the water gap in the uncontrolled portion. (authors)

  6. BWR Fuel Assemblies Physics Analysis Utilizing 3D MCNP Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Chiang, Ren-Tai [University of Florida, Gainesville, Florida 32611 (United States); Williams, John B.; Folk, Ken S. [Southern Nuclear Company, Birmingham, Alabama 35242 (United States)

    2008-07-01

    MCNP is used to model a partially controlled BWR fresh fuel four assemblies (2x2) system for better understanding BWR fuel behavior and for benchmarking production codes. The impact of the GE14 plenum regions on axial power distribution is observed by comparing against the GE13 axial power distribution, in which the GE14 relative power is lower than the GE13 relative power at the 15. node and at the 16. node due to presence of the plenum regions in GE14 fuel in these two nodes. The segmented rod power distribution study indicates that the azimuthally dependent power distribution is very significant for the fuel rods next to the water gap in the uncontrolled portion. (authors)

  7. Initiation model for intergranular stress corrosion cracking in BWR pipes

    International Nuclear Information System (INIS)

    Hishida, Mamoru; Kawakubo, Takashi; Nakagawa, Yuji; Arii, Mitsuru.

    1981-01-01

    Discussions were made on the keys of intergranular stress corrosion cracking of austenitic stainless steel in high-temperature water in laboratories and stress corrosion cracking incidents in operating plants. Based on these discussions, a model was set up of intergranular stress corrosion cracking initiation in BWR pipes. Regarding the model, it was presumed that the intergranular stress corrosion cracking initiates during start up periods whenever heat-affected zones in welded pipes are highly sensitized and suffer dynamic strain in transient water containing dissolved oxygen. A series of BWR start up simulation tests were made by using a flowing autoclave system with slow strain rate test equipment. Validity of the model was confirmed through the test results. (author)

  8. A BWR 24-month cycle analysis using multicycle techniques

    International Nuclear Information System (INIS)

    Hartley, K.D.

    1993-01-01

    Boiling water reactor (BWR) fuel cycle design analyses have become increasingly challenging in the past several years. As utilities continue to seek improved capacity factors, reduced power generation costs, and reduced outage costs, longer cycle lengths and fuel design optimization become important considerations. Accurate multicycle analysis techniques are necessary to determine the viability of fuel designs and cycle operating strategies to meet reactor operating requirements, e.g., meet thermal and reactivity margin constraints, while minimizing overall fuel cycle costs. Siemens Power Corporation (SPC), Nuclear Division, has successfully employed multi-cycle analysis techniques with realistic rodded cycle depletions to demonstrate equilibrium fuel cycle performance in 24-month cycles. Analyses have been performed by a BWR/5 reactor, at both rated and uprated power conditions

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

    International Nuclear Information System (INIS)

    1992-09-01

    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 specifications. This report contains three volumes. This document, Volume 2, contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS

  10. Neutron noise analysis of BWR using time series analysis

    International Nuclear Information System (INIS)

    Fukunishi, Kohyu

    1976-01-01

    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) [de

  11. Condensate polishing guidelines for PWR and BWR plants

    International Nuclear Information System (INIS)

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

    1993-03-01

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

  12. Vertical Drop of 44-BWR Waste Package With Lifting Collars

    Energy Technology Data Exchange (ETDEWEB)

    A.K. Scheider

    2005-08-23

    The objective of this calculation is to determine the structural response of a waste package (WP) dropped flat on its bottom from a specified height. The WP used for that purpose is the 44-Boiling Water Reactor (BWR) WP. The scope of this document is limited to reporting the calculation results in terms of stress intensities. The Uncanistered Waste Disposal Container System is classified as Quality Level 1 (Ref. 4, page 7). Therefore, this calculation is subject to the requirements of the Quality Assurance Requirements and Description (Ref. 16). AP-3. 12Q, Design Calculations and Analyses (Ref. 11) is used to perform the calculation and develop the document. The information provided by the sketches attached to this calculation is that of the potential design of the type of 44-BWR WP considered in this calculation and provides the potential dimensions and materials for that design.

  13. Study on thermal performance and margins of BWR fuel elements

    International Nuclear Information System (INIS)

    Stosic, Zoran

    1999-01-01

    This paper contributes to developing a methodology of predicting and analyzing thermal performance and margins of Boiling Water Reactor (BWR) fuel assemblies under conditions of reaching high quality Boiling Crisis and subsequent post-dryout thermal hydraulics causing temperature excursion of fuel cladding. Operational margins against dryout and potential for increasing fuel performance with appropriate benefits are discussed. The philosophy of modeling with its special topics are demonstrated on the HECHAN (HEated CHannel ANalyzer) model as the state-of-art for thermal-hydraulics analysis of BWR fuel assemblies in pre- and post-dryout two-phase flow regimes. The scope of further work either being or has to be performed concerning implementation of new physical aspects, including domain extension of HECHAN model applications to the Pressurized Water Reactors (PWRs), is discussed. Finally, a comprehensive overview of the literature dealing with development of the model is given. (author)

  14. Paired replacement fuel assemblies for BWR-type reactor

    International Nuclear Information System (INIS)

    Oguchi, Kazushige.

    1997-01-01

    There are disposed a large-diameter water rod constituting a non-boiling region at a central portion and paired replacement fuel assemblies for two streams having the same average enrichment degree and different amount of burnable poisons. The paired replacement fuel assemblies comprise a first fuel assembly having a less amount of burnable poisons and a second fuel assembly having a larger amount of burnable poisons. A number of burnable poison-containing fuel rods in adjacent with the large diameter water rod is increased in the second fuel assembly than the first fuel assembly. Then, the poison of the paired replacement fuel assemblies for the BWR type reactor can be annihilated simultaneously at the final stage of the cycle. Accordingly, fuels for a BWR type reactor excellent in economical property and safety and facilitating the design of the replacement reactor core can be obtained. (N.H.)

  15. Review of international solutions to NEACRP benchmark BWR lattice cell problems

    International Nuclear Information System (INIS)

    Halsall, M.J.

    1977-12-01

    This paper summarises international solutions to a set of BWR benchmark problems. The problems, posed as an activity sponsored by the Nuclear Energy Agency Committee on Reactor Physics, were as follows: 9-pin supercell with central burnable poison pin, mini-BWR with 4 pin-cells and water gaps and control rod cruciform, full 7 x 7 pin BWR lattice cell with differential U 235 enrichment, and full 8 x 8 pin BWR lattice cell with water-hole, Pu-loading, burnable poison, and homogenised cruciform control rod. Solutions have been contributed by Denmark, Japan, Sweden, Switzerland and the UK. (author)

  16. Evaluation of PWR and BWR pin cell benchmark results

    Energy Technology Data Exchange (ETDEWEB)

    Pilgroms, B.J.; Gruppelaar, H.; Janssen, A.J. (Netherlands Energy Research Foundation (ECN), Petten (Netherlands)); Hoogenboom, J.E.; Leege, P.F.A. de (Interuniversitair Reactor Inst., Delft (Netherlands)); Voet, J. van der (Gemeenschappelijke Kernenergiecentrale Nederland NV, Dodewaard (Netherlands)); Verhagen, F.C.M. (Keuring van Electrotechnische Materialen NV, Arnhem (Netherlands))

    1991-12-01

    Benchmark results of the Dutch PINK working group on the PWR and BWR pin cell calculational benchmark as defined by EPRI are presented and evaluated. The observed discrepancies are problem dependent: a part of the results is satisfactory, some other results require further analysis. A brief overview is given of the different code packages used in this analysis. (author). 14 refs.; 9 figs.; 30 tabs.

  17. Latest experiences in inspecting the inside of BWR vessel shields

    Energy Technology Data Exchange (ETDEWEB)

    Alberdi, R.; Gonzalez, E.

    2001-07-01

    In the last few years, the owners of BWR nuclear power plants have been forced to address new fuel shield inspection requirements, TECNATOM has responded to this situation by launching the TEIDE projects, which include development of an inspection machine and the corresponding Non-Destructive Tests to examine the inside of this shield. With these projects, TECNATOM has performed more than 12 fuel shield inspections in different countries. This article describes the experience gained in the last three years. (Author)

  18. Results of modeling advanced BWR fuel designs using CASMO-4

    International Nuclear Information System (INIS)

    Knott, D.; Edenius, M.

    1996-01-01

    Advanced BWR fuel designs from General Electric, Siemens and ABB-Atom have been analyzed using CASMO-4 and compared against fission rate distributions and control rod worths from MCNP. Included in the analysis were fuel storage rack configurations and proposed mixed oxide (MOX) designs. Results are also presented from several cycles of SIMULATE-3 core follow analysis, using nodal data generated by CASMO-4, for cycles in transition from 8x8 designs to advanced fuel designs. (author)

  19. Thermohydraulic stability coupled to the neutronic in a BWR

    International Nuclear Information System (INIS)

    Calleros M, G.; Zapata Y, M.; Gomez H, R.A.; Mendez M, A.; Castlllo D, R.

    2006-01-01

    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 Laguna Verde

  20. TVA experience in BWR reload design and licensing

    International Nuclear Information System (INIS)

    Robertson, J.D.

    1986-01-01

    TVA has developed and implemented the capability to perform BWR reload core design and licensing analyses. The advantages accruing from this capability include the tangible cost-savings from performing reload analyses in-house. Also, ''intangible'' benefits such as increased operating flexibility and the ability to accommodate multivendor fuel designs have been demonstrated. The major disadvantage with performing in-house analyses is the cost associated with development and maintenance of the analytical methods and staff expertise

  1. Maximum thermal loading test of BWR fuel assembly

    International Nuclear Information System (INIS)

    Nakajima, Yoshitaka; Yoshimura, Kunihiro; Nakamura, Satoshi; Ishizuka, Takao.

    1987-01-01

    Various proving tests on the reliability of nuclear power plants have been conducted at the Nuclear Power Engineering Test Center and at the Japan Power Plant Engineering and Inspection Corporation. The tests were initiated at the request of the Ministry of International Trade and Industry (MITI). Toshiba undertook one of the proving tests on the reliability of nuclear fuel assembly; the maximum thermal loading test of BWR fuel assembly from the Nuclear Power Engineering Test Center. These tests are part of the proving tests mentioned above, and their purpose is to confirm the reliability of the thermal hydraulic engineering techniques. Toshiba has been engaged for the past nine years in the design, fabrication and testing of the equipment. For the project, a test model fuel assembly was used to measure the critical power of the BWR fuel assembly and the void and fluidity of the coolant. From the test results, it has been confirmed that the heat is transferred safely from the fuel assembly to the coolant in the BWR nuclear power plant. In addition, the propriety and reliability of the thermal hydraulic engineering techniques for the fuel assembly have been proved. (author)

  2. The BWR vessel and internals project - 2001 and beyond

    International Nuclear Information System (INIS)

    Wagoner, V.; Mulford, T.

    2001-01-01

    The BWR Vessel and Internals Project (BWRVIP) is an international association of utilities owning boiling water reactors (BWRs). Figure 1 contains a list of current BWRVIP member utilities. The association was formed in 1994 with the following objectives: to lead the BWR industry toward generic resolution of reactor pressure vessel and internals material condition issues; to identify or develop generic, cost-effective strategies from which each operating plant will select the most appropriate alternative; to serve as the focal point for the regulatory interface with the industry on BWR vessel and internals issues; and to share information and promote communication and cooperation among participating utilities. The initial issue faced by the BWRVIP was core shroud cracking that had been observed in a number of BWRs. The BWRVIP reacted by quickly developing a set of industry guidelines to assist utilities in inspecting, evaluating, and, if necessary, repairing cracked shrouds. Subsequently, the BWRVIP pro-actively developed a comprehensive set of guidelines for managing degradation in other reactor internal components, including the reactor pressure vessel itself. The major components addressed by the BWRVIP are included. (author)

  3. The BWR vessel and internals project - 2001 and beyond

    Energy Technology Data Exchange (ETDEWEB)

    Wagoner, V. [Carolina Power and Light, Progress Energy Building, NC (United States); Mulford, T. [Electric Power Research Institute, Palo Alto, CA (United States)

    2001-07-01

    The BWR Vessel and Internals Project (BWRVIP) is an international association of utilities owning boiling water reactors (BWRs). Figure 1 contains a list of current BWRVIP member utilities. The association was formed in 1994 with the following objectives: to lead the BWR industry toward generic resolution of reactor pressure vessel and internals material condition issues; to identify or develop generic, cost-effective strategies from which each operating plant will select the most appropriate alternative; to serve as the focal point for the regulatory interface with the industry on BWR vessel and internals issues; and to share information and promote communication and cooperation among participating utilities. The initial issue faced by the BWRVIP was core shroud cracking that had been observed in a number of BWRs. The BWRVIP reacted by quickly developing a set of industry guidelines to assist utilities in inspecting, evaluating, and, if necessary, repairing cracked shrouds. Subsequently, the BWRVIP pro-actively developed a comprehensive set of guidelines for managing degradation in other reactor internal components, including the reactor pressure vessel itself. The major components addressed by the BWRVIP are included. (author)

  4. Seismic proving test of BWR primary loop recirculation system

    International Nuclear Information System (INIS)

    Sato, H.; Shigeta, M.; Karasawa, Y.

    1987-01-01

    The seismic proving test of BWR Primary Loop Recirculation system is the second test to use the large-scale, high-performance vibration table of Tadotsu Engineering Laboratory. The purpose of this test is to prove the seismic reliability of the primary loop recirculation system (PLR), one of the most important safety components in the BWR nuclear plants, and also to confirm the adequacy of seismic analysis method used in the current seismic design. To achieve the purpose, the test was conducted under conditions and scale as near as possible to actual systems. The strength proving test was carried out with the test model mounted on the vibration table in consideration of basic design earthquake ground motions and other conditions to confirm the soundness of structure and the strength against earthquakes. Detailed analysis and analytic evaluation of the data obtained from the test was conducted to confirm the adequacy of the seismic analysis method and earthquake response analysis method used in the current seismic design. Then, on the basis of the results obtained, the seismic safety and reliability of BWR primary loop recirculation of the actual plants was fully evaluated

  5. Novel modular natural circulation BWR design and safety evaluation

    International Nuclear Information System (INIS)

    Ishii, Mamoru; Shi, Shanbin; Yang, Won Sik; Wu, Zeyun; Rassame, Somboon; Liu, Yang

    2015-01-01

    Highlights: • Introduction of BWR-type natural circulation small modular reactor preliminary design (NMR-50). • Design of long fuel cycle length for the NMR-50. • Design of double passive safety systems for the NMR-50. • RELAP5 analyses of design basis accidents for the NMR-50. - Abstract: The Purdue NMR (Novel Modular Reactor) represents a BWR-type small modular reactor with a significantly reduced reactor pressure vessel (RPV) height. Specifically, it has one third the height of a conventional BWR RPV with an electrical output of 50 MWe. The preliminary design of the NMR-50 including reactor, fuel cycle, and safety systems is described and discussed. The improved neutronics design of the NMR-50 extends the fuel cycle length up to 10 years. The NMR-50 is designed with double passive engineering safety system, which is intended to withstand a prolonged station black out with loss of ultimate heat sink accident such as experienced at Fukushima. In order to evaluate the safety features of the NMR-50, two representative design basis accidents, i.e. main steam line break (MSLB) and bottom drain line break (BDLB), are simulated by using the best-estimate thermal–hydraulic code RELAP5. The RPV water inventory, containment pressure, and the performance of engineering safety systems are investigated for about 33 h after the initiation of the accidents

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

    International Nuclear Information System (INIS)

    Nakajima, Tsuyoshi; Iwamoto, Tatsuya; Kumanomido, Hironori

    1996-01-01

    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 140 La 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

  7. Sensitivity of BWR shutdown margin tests to local reactivity anomalies

    International Nuclear Information System (INIS)

    Cokinos, D.M.; Carew, J.F.

    1987-01-01

    Successful shutdown margin (SDM) demonstration is a required procedure in the startup of a newly configured boiling water reactor (BWR) core. In its most reactive condition throughout a cycle, a BWR core must be capable of being made subcritical by a specified margin with the highest worth control rod fully withdrawn and all other rods at their fully inserted positions. Two different methods are used to demonstrate SDM: (a) the adjacent-rod test and (b) the in-sequence test. In the adjacent-rod test, the strongest rod is fully withdrawn and an adjacent rod is withdrawn to reach criticality. In the in-sequence test, control rods spread throughout the core are withdrawn in a predetermined sequence of withdrawals. Larger than expected core k/sub eff/ values have been observed during the performance of BWR SDM tests. The purpose of the work summarized in this paper has been to investigated and quantify the sensitivity of both the adjacent-rod and in-sequence SDM tests to local reactivity anomalies. This was accomplished by introducing reactivity perturbations at selected four-bundle cell locations and by evaluating their effect on core reactivity in each of the two tests

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

    International Nuclear Information System (INIS)

    Morikawa, Yoshitake

    1995-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-03-01

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

  10. Advanced chemistry management system to optimize BWR chemistry control

    International Nuclear Information System (INIS)

    Maeda, K.; Nagasawa, K.

    2002-01-01

    BWR plant chemistry control has close relationships among nuclear safety, component reliability, radiation field management and fuel integrity. Advanced technology is required to improve chemistry control [1,3,6,7,10,11]. Toshiba has developed TACMAN (Toshiba Advanced Chemistry Management system) to support BWR chemistry control. The TACMAN has been developed as response to utilities' years of requirements to keep plant operation safety, reliability and cost benefit. The advanced technology built into the TACMAN allows utilities to make efficient chemistry control and to keep cost benefit. TACMAN is currently being used in response to the needs for tools those plant chemists and engineers could use to optimize and identify plant chemistry conditions continuously. If an incipient condition or anomaly is detected at early stage, root causes evaluation and immediate countermeasures can be provided. Especially, the expert system brings numerous and competitive advantages not only to improve plant chemistry reliability but also to standardize and systematize know-how, empirical knowledge and technologies in BWR chemistry This paper shows detail functions of TACMAN and practical results to evaluate actual plant. (authors)

  11. Large bundle BWR test CORA-18: Test results

    International Nuclear Information System (INIS)

    Hagen, S.; Hofmann, P.; Noack, V.; Sepold, L.; Schanz, G.; Schumacher, G.

    1998-04-01

    The CORA out-of-pile experiments are part of the international Severe Fuel Damage (SFD) Program. They were performed to provide information on the damage progression of Light Water Reactor (LWR) fuel elements in Loss-of-coolant Accidents in the temperature range 1200 C to 2400 C. CORA-18 was the large BWR bundle test corresponding to the PWR test CORA-7. It should investigate if there exists an influence of the BWR bundle size on the fuel damage behaviour. Therefore, the standard-type BWR CORA bundle with 18 fuel rod simulators was replaced by a large bundle with two additional surrounding rows of 30 rods (48 rods total). Power input and steam flow were increased proportionally to the number of fuel rod simulators to give the same initial heat-up rate of about 1 K/s as in the smaller bundles. Emphasis was put on the initial phase of the damage progression. More information on the chemical composition of initial and intermediate interaction products and their relocation behaviour should be obtained. Therefore, power and steam input were terminated after the onset of the temperature escalation. (orig.) [de

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

    Directory of Open Access Journals (Sweden)

    Mariko Abe

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

  13. Parametric tests of the effects of water chemistry impurities on corrosion of Zr-alloys under simulated BWR condition

    Energy Technology Data Exchange (ETDEWEB)

    Shimada, S; Ito, K [Nippon Nuclear Fuel Development Co. Ltd., Oarai, Ibaraki (Japan); Lin, C C [GE Nucklear Energy (United States); Cheng, B [Electric Power Research Inst. (United States); Ikeda, T [Toshiba Corp. (Japan); Oguma, M [Hitachi, Ltd (Japan); Takei, T [Tokyo Electric Power Co., Inc. (Japan); Vitanza, C; Karlsen, T M [Institutt for Energiteknikk, Halden (Norway). OECD Halden Reaktor Projekt

    1997-02-01

    The Halden BWR corrosion test loop was constructed to evaluate the impact of water chemistry variables, heat flux and boiling condition on corrosion performance of Zr-alloys in a simulated BWR environment. The loop consists of two in-core rigs, one for testing fuel rod segments and the other for evaluating water chemistry variables utilizing four miniautoclaves. Ten coupon specimens are enclosed in each miniautoclave. The Zr-alloys for the test include Zircaloy-2 having different nodular corrosion resistance and five new alloys. The first and second of the six irradiation tests planned in this program were completed. Post-irradiation examination of those test specimens have shown that the test loop is capable of producing nodular corrosion on the fuel rod cladding tested under the reference chemistry condition. The miniautoclave tests showed that nodular corrosion could be formed without flux and boiling under some water chemistry conditions and the new alloys, generally, had higher corrosion resistance than the Zircaloy in high oxygen environments. (author). 5 refs, 4 figs, 5 tabs.

  14. 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

  15. Increased expression of bHLH transcription factor E2A (TCF3) in prostate cancer promotes proliferation and confers resistance to doxorubicin induced apoptosis

    International Nuclear Information System (INIS)

    Patel, Divya; Chaudhary, Jaideep

    2012-01-01

    Highlights: ► E2A, considered as a tumor suppressor is highly expressed in prostate cancer. ► Silencing of E2A attenuates cell proliferation and promotes apoptosis. ► E2A regulates c-myc, Id1, Id3 and CDKN1A expression. ► Loss of E2A promotes doxorubicin dependent apoptosis in prostate cancer cells. ► Results suggest that E2A acts as a tumor promoter at least in prostate cancer. -- Abstract: E2A (TCF3) is a multifunctional basic helix loop helix (bHLH), transcription factor. E2A regulates transcription of target genes by homo- or heterodimerization with cell specific bHLH proteins. In general, E2A promotes cell differentiation, acts as a negative regulator of cell proliferation in normal cells and cancer cell lines and is required for normal B-cell development. Given the diverse biological pathways regulated/influenced by E2A little is known about its expression in cancer. In this study we investigated the expression of E2A in prostate cancer. Unexpectedly, E2A immuno-histochemistry demonstrated increased E2A expression in prostate cancer as compared to normal prostate. Silencing of E2A in prostate cancer cells DU145 and PC3 led to a significant reduction in proliferation due to G1 arrest that was in part mediated by increased CDKN1A(p21) and decreased Id1, Id3 and c-myc. E2A silencing in prostate cancer cell lines also resulted in increased apoptosis due to increased mitochondrial permeability and caspase 3/7 activation. Moreover, silencing of E2A increased sensitivity to doxorubicin induced apoptosis. Based on our results, we propose that E2A could be an upstream regulator of Id1 and c-Myc which are highly expressed in prostate cancer. These results for the first time demonstrate that E2A could in fact acts as a tumor promoter at least in prostate cancer.

  16. BWR chemistry control status: a summary of industry chemistry status relative to the BWR water chemistry guidelines

    International Nuclear Information System (INIS)

    Garcia, S.E.; Giannelli, J.F.; Jarvis, M.L.

    2010-01-01

    The EPRI Boiling Water Reactor (BWR) Water Chemistry Guidelines were revised and issued in October 2008. The 2008 Revision of the Guidelines continues to focus on intergranular stress corrosion cracking (IGSCC), which can limit the service life of susceptible materials and components exposed to water chemistry environments. The 2008 Revision also places increased emphasis on fuel performance and meeting the industry goal of zero fuel failures by 2010. As an industry consensus document, the Guidelines were created to provide proactive water chemistry control strategies for mitigating IGSCC, maintaining fuel integrity and controlling radiation fields. The Guidelines provide a technically-based framework for an effective BWR water chemistry program. This paper provides an overview of industry experience relative to the Guidelines. Over the past few years, many BWR units have implemented noble metal chemical application technologies either during plant hot or cold shutdown or at normal power operating conditions. This paper explores plant experience with optimized water chemistry, implementation of various additive chemistries such as noble metal application and zinc addition, and compliance with the Guidelines recommendations. Depleted zinc oxide addition has been broadly applied across the BWR fleet since the 1980s. The guidance for zinc addition has been revised in the Guidelines to reflect concerns with fuel performance. While zinc addition is a successful method for shutdown dose rate control, concerns still exist for high zinc deposition on fuel surfaces, especially when feedwater iron is elevated and as fuel cores are being driven to provide maximum power output over longer fuel cycles. Recent plant experience has shown that the utilization of online noble metal application and continued zinc addition may provide additional benefits for radiation control. Dose rate experiences at plants utilizing the online noble metal application technology and zinc addition

  17. BWR chemistry control status: a summary of industry chemistry status relative to the BWR water chemistry guidelines

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, S.E., E-mail: sgarcia@epri.com [Electric Power Research Inst. (EPRI), Palo Alto, California (United States); Giannelli, J.F.; Jarvis, M.L., E-mail: jgiannelli@finetech.com [Finetech, Inc., Parsippany, NJ (United States)

    2010-07-01

    The EPRI Boiling Water Reactor (BWR) Water Chemistry Guidelines were revised and issued in October 2008. The 2008 Revision of the Guidelines continues to focus on intergranular stress corrosion cracking (IGSCC), which can limit the service life of susceptible materials and components exposed to water chemistry environments. The 2008 Revision also places increased emphasis on fuel performance and meeting the industry goal of zero fuel failures by 2010. As an industry consensus document, the Guidelines were created to provide proactive water chemistry control strategies for mitigating IGSCC, maintaining fuel integrity and controlling radiation fields. The Guidelines provide a technically-based framework for an effective BWR water chemistry program. This paper provides an overview of industry experience relative to the Guidelines. Over the past few years, many BWR units have implemented noble metal chemical application technologies either during plant hot or cold shutdown or at normal power operating conditions. This paper explores plant experience with optimized water chemistry, implementation of various additive chemistries such as noble metal application and zinc addition, and compliance with the Guidelines recommendations. Depleted zinc oxide addition has been broadly applied across the BWR fleet since the 1980s. The guidance for zinc addition has been revised in the Guidelines to reflect concerns with fuel performance. While zinc addition is a successful method for shutdown dose rate control, concerns still exist for high zinc deposition on fuel surfaces, especially when feedwater iron is elevated and as fuel cores are being driven to provide maximum power output over longer fuel cycles. Recent plant experience has shown that the utilization of online noble metal application and continued zinc addition may provide additional benefits for radiation control. Dose rate experiences at plants utilizing the online noble metal application technology and zinc addition

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

    DEFF Research Database (Denmark)

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

    2017-01-01

    . However, the role of SHP-1 in intimal hyperplasia and restenosis has not been clarified in insulin resistance and diabetes. METHODS: We used a femoral artery wire injury mouse model, rodent models with insulin resistance and diabetes, and patients with type 2 diabetes. Further, we modulated SHP-1...... expression using a transgenic mouse that overexpresses SHP-1 in VSMCs (Shp-1-Tg). SHP-1 agonists were also employed to study the molecular mechanisms underlying the regulation of SHP-1 by oxidised lipids. RESULTS: Mice fed a high-fat diet (HFD) exhibited increased femoral artery intimal hyperplasia...... and decreased arterial SHP-1 expression compared with mice fed a regular diet. Arterial SHP-1 expression was also decreased in Zucker fatty rats, Zucker diabetic fatty rats and in patients with type 2 diabetes. In primary cultured VSMCs, oxidised LDL suppressed SHP-1 expression by activating Mek-1 (also known...

  19. User requirements for innovative nuclear reactors and fuel cycle technologies in the area of economics, environment, safety, waste management, proliferation resistance and cross cutting issues, and methodology for innovative technologies assessment

    International Nuclear Information System (INIS)

    Kupitz, Juergen; Depisch, Frank; Allan, Colin

    2003-01-01

    The IAEA General Conference in 2000 has invited ''all interested Member States to combine their efforts under the aegis of the Agency in considering the issues of the nuclear fuel cycle, in particular by examining innovative and proliferation-resistant nuclear technology''. In response to this invitation, the IAEA initiated an ''International Project on Innovative Nuclear Reactors and Fuel Cycles'', INPRO. The overall objectives of INPRO is to help to ensure that nuclear energy is available to contribute in fulfilling in a sustainable manner energy needs in the 21st century, and to bring together all interested Member States, 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 that use sound and economically competitive technology. Phase-I of INPRO was initiated in May 2001. During Phase-I, work was subdivided in two sub phase: Phase 1A (finished in June 2003) and Phase 1B (started in June 2003). Phase 1A dealt with the definition of Basic Principles, User Requirements and Criteria, and the development of a methodology for the evaluation of innovative nuclear technologies. In Phase 1A, task groups for several areas were established: (a) Prospects and Potentials of Nuclear Power, (b) Economics; (c) Sustainability and Environment, (d) Safety of Nuclear Installations, (e) Waste Management, (f) Proliferation Resistance, (g) Crosscutting issues and (h) for the Methodology for Assessment. In Phase-IB evaluations of innovative nuclear energy technologies will be performed by Member States against the INPRO Basic Principles, User Requirements and Criteria. This paper summarizes the results achieved in the Phase 1A of INPRO and is a cooperative effort of the INPRO team, consisting of all INPRO cost free experts and task managers. (author)

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

    International Nuclear Information System (INIS)

    Kostic, Lj.

    1983-01-01

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

    International Nuclear Information System (INIS)

    Donato, A.; Ricci, G.; Pace, A.

    1985-01-01

    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

  2. Analysis of multi-dimensional and countercurrent effects in a BWR loss-of-coolant accident

    International Nuclear Information System (INIS)

    Shiralkar, B.S.; Dix, G.E.; Alamgir, M.

    1989-01-01

    The presence of parallel enclosed channels in a BWR provides opportunities for multiple flow regimes in co-current and countercurrent flow under Loss-of-Coolant Accident (LOCA) conditions. To address and understand these phenomena, an integrated experimental and analytical study has been conducted. The primary experimental facility was the Steam Sector Test Facility (SSTF) which simulated a full scale 30deg sector of a BWR/6 reactor vessel. Both steady-state separate effects tests and integral transients with vessel blowdown and refill were performed. The present of multi-dimensional and parallel channel effects was found to be very beneficial to BWR LOCA performance. The best estimate TRAC-BWR computer code was extended as part of this study by incorporation of a phenomenological upper plenum mixing model. TRAC-BWR was applied to the analysis of these full scale experiments. Excellent predictions of phenomena and experimental trends were achieved. (orig.)

  3. Analysis of multidimensional and countercurrent effects in a BWR loss-of-coolant accident

    International Nuclear Information System (INIS)

    Shiralkar, B.S.; Dix, G.E.; Alamgir, M.

    1991-01-01

    The presence of parallel enclosed channels in a boiling water reactor (BWR) provides opportunities for multiple flow regimes in cocurrent and countercurrent flow under loss-of-coolant accident (LOCA) conditions. To address and understand these phenomena, an integrated experimental and analytical study has been conducted. The primary experimental facility was the steam sector test facility (SSFT), which simulated a full scale 30deg sector of a BWR/6 reactor vessel. Both steady-state separate effects tests an integral transients with vessel vlowdown and refill were performed. The presence of multidimensional and parallel-channel effects was found to be very beneficial to BWR LOCA performance. The best estimate TRAC-BWR computer code was extended as part of this study by incorporation of a phenomenological upper plenum mixing model. TRAC-BWR was applied to the analysis of these full scale experiments. Excellent predictions of phenomena and experimental trends were achieved. (orig.)

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

    International Nuclear Information System (INIS)

    Espinosa-Paredes, Gilberto; Nunez-Carrera, Alejandro; Vazquez-Rodriguez, Alejandro

    2006-01-01

    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

  5. Emergency condensator for BWR type reactor

    International Nuclear Information System (INIS)

    Ubakai, Yoichi; Narumi, Yuichi; Sakata, Yuji.

    1992-01-01

    An emergency condensator is constituted with heat transfer pipes, a steam chamber, an upper pipe plate, a lower pipe plate and a condensate chamber. The upper pipe plate is secured by supports, and a steam pipe is connected to the upper pipe plate. A condensate pipeline and a incondensible gas vent pipe are disposed to the condensate chamber. Taking thermal expansion of the steam pipes and thermal expansion of the heat transfer pipes into consideration, the heat transfer pipe is made as an L-shaped pipe having a vertical portion and a horizontal portion so as to absorb each of the thermal expansion smoothly. The L-shaped heat transfer pipes are constituted as a bundle of pipes having the end portions thereof secured to the upper pipe plate and the lower pipe plate. The emergency condensator is disposed in a emergency condensator pool chamber. Cooling water in contact with the outer side of the L-shaped heat transfer pipes is the pool water in the pool chamber, and the condensator chamber is disposed in concrete walls of the pool chamber. With such a constitution, stress due to thermal expansion of the heat transfer pipes is mitigated, and heat transfer performance, earth quake resistance and maintenancability are improved. (I.N.)

  6. 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.

  7. 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.

  8. Developing and modeling of the 'Laguna Verde' BWR CRDA benchmark

    International Nuclear Information System (INIS)

    Solis-Rodarte, J.; Fu, H.; Ivanov, K.N.; Matsui, Y.; Hotta, A.

    2002-01-01

    Reactivity initiated accidents (RIA) and design basis transients are one of the most important aspects related to nuclear power reactor safety. These events are re-evaluated whenever core alterations (modifications) are made as part of the nuclear safety analysis performed to a new design. These modifications usually include, but are not limited to, power upgrades, longer cycles, new fuel assembly and control rod designs, etc. The results obtained are compared with pre-established bounding analysis values to see if the new core design fulfills the requirements of safety constraints imposed on the design. The control rod drop accident (CRDA) is the design basis transient for the reactivity events of BWR technology. The CRDA is a very localized event depending on the control rod insertion position and the fuel assemblies surrounding the control rod falling from the core. A numerical benchmark was developed based on the CRDA RIA design basis accident to further asses the performance of coupled 3D neutron kinetics/thermal-hydraulics codes. The CRDA in a BWR is a mostly neutronic driven event. This benchmark is based on a real operating nuclear power plant - unit 1 of the Laguna Verde (LV1) nuclear power plant (NPP). The definition of the benchmark is presented briefly together with the benchmark specifications. Some of the cross-sections were modified in order to make the maximum control rod worth greater than one dollar. The transient is initiated at steady-state by dropping the control rod with maximum worth at full speed. The 'Laguna Verde' (LV1) BWR CRDA transient benchmark is calculated using two coupled codes: TRAC-BF1/NEM and TRAC-BF1/ENTREE. Neutron kinetics and thermal hydraulics models were developed for both codes. Comparison of the obtained results is presented along with some discussion of the sensitivity of results to some modeling assumptions

  9. Prony's method application for BWR instabilities characterization

    International Nuclear Information System (INIS)

    Castillo, Rogelio; Ramírez, J. Ramón; Alonso, Gustavo; Ortiz-Villafuerte, Javier

    2015-01-01

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

  10. BWR and PWR chemistry operating experience and perspectives

    International Nuclear Information System (INIS)

    Fruzzetti, K.; Garcia, S.; Lynch, N.; Reid, R.

    2014-01-01

    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)

  11. 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

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

    Science.gov (United States)

    Boehme, Karen A.; Schleicher, Sabine B.; Rolauffs, Bernd

    2018-01-01

    Unlike other malignant bone tumors including osteosarcomas and Ewing sarcomas with a peak incidence in adolescents and young adults, conventional and dedifferentiated chondrosarcomas mainly affect people in the 4th to 7th decade of life. To date, the cell type of chondrosarcoma origin is not clearly defined. However, it seems that mesenchymal stem and progenitor cells (MSPC) in the bone marrow facing a pro-proliferative as well as predominantly chondrogenic differentiation milieu, as is implicated in early stage osteoarthritis (OA) at that age, are the source of chondrosarcoma genesis. But how can MSPC become malignant? Indeed, only one person in 1,000,000 will develop a chondrosarcoma, whereas the incidence of OA is a thousandfold higher. This means a rare coincidence of factors allowing escape from senescence and apoptosis together with induction of angiogenesis and migration is needed to generate a chondrosarcoma. At early stages, chondrosarcomas are still assumed to be an intermediate type of tumor which rarely metastasizes. Unfortunately, advanced stages show a pronounced resistance both against chemo- and radiation-therapy and frequently metastasize. In this review, we elucidate signaling pathways involved in the genesis and therapeutic resistance of chondrosarcomas with a focus on MSPC compared to signaling in articular cartilage (AC). PMID:29361725

  13. Rare Mutations of Peroxisome Proliferator-Activated Receptor Gamma: Frequencies and Relationship with Insulin Resistance and Diabetes Risk in the Mixed Ancestry Population from South Africa

    Directory of Open Access Journals (Sweden)

    Z. Vergotine

    2014-01-01

    Full Text Available Background. Genetic variants in the nuclear transcription receptor, PPARG, are associated with cardiometabolic traits, but reports remain conflicting. We determined the frequency and the clinical relevance of PPARG SNPs in an African mixed ancestry population. Methods. In a cross-sectional study, 820 participants were genotyped for rs1800571, rs72551362, rs72551363, rs72551364, and rs3856806, using allele-specific TaqMan technology. The homeostatic model assessment of insulin (HOMA-IR, β-cells function (HOMA-B%, fasting insulin resistance index (FIRI, and the quantitative insulin-sensitivity check index (QUICKI were calculated. Results. No sequence variants were found except for the rs3856806. The frequency of the PPARG-His447His variant was 23.8% in the overall population group, with no difference by diabetes status (P=0.215. The His447His allele T was associated with none of the markers of insulin resistance overall and by diabetes status. In models adjusted for 2-hour insulin, the T allele was associated with lower prevalent diabetes risk (odds ratio 0.56 (95% CI 0.31–0.95. Conclusion. Our study confirms the almost zero occurrences of known rare PPARG SNPs and has shown for the first time in an African population that one of the common SNPs, His447His, may be protective against type 2 diabetes.

  14. Void effects on BWR Doppler and void reactivity feedback

    International Nuclear Information System (INIS)

    Hsiang-Shou Cheng; Diamond, D.J.

    1978-01-01

    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)

  15. Measurements and analysis of neutron and gamma noise in BWR's

    International Nuclear Information System (INIS)

    Dam, H. van; Kleiss, E.B.J.

    1985-01-01

    Neutron and gamma sensitive collectrons (self-powered detectors) have been designed for incore noise measurements in BWRs. A so-called twin-type has been developed for measurements of two-phase flow characteristics and detailed axial velocity distributions. Construction aspects of the twin detectors are discussed. An analysis is presented of the response of both detector types to incore parametric fluctuations. This analysis is based on detector response functions which provide an insight into the 'field of view' of the two types. The results are supported by experimental verifications; it is shown that incore gamma detectors provide useful additional information about two-phase flow in a BWR. (author)

  16. RETRAN experience with BWR transients at Yankee Atomic Electric Company

    International Nuclear Information System (INIS)

    Ansari, A.A.F.; Cronin, J.T.; Slifer, B.C.

    1981-01-01

    Yankee Atomic Electric Company is actively involved in the development of licensing methods for BWR's. The computer code chosen for analyzing system response under transient conditions is RETRAN. This paper describes the RETRAN model developed for Vermont Yankee, and the results of the RETRAN checkout and qualification that has been achieved at YAEC through comparison of RETRAN predictions to the startup test results performed at the plant as part of the 100% power startup test program. In addition, abnormal operational transients typically analyzed for licensing are also presented

  17. A study on the feasibility of minor actinides in BWR

    International Nuclear Information System (INIS)

    Abdul Waris; Budiono

    2008-01-01

    Preliminary study on the feasibility of actinides minor (MA) recycling without mixing them with plutonium in boiling water reactor (BWR) has been carried out. The results show that increasing of fissile MA content in mixed oxide fuel (MOX) and/or reducing void fraction can enlarge the effective multiplication factor at the beginning of cycle, but the reactor still can not obtain its criticality condition. Furthermore, dropping the void fraction results in higher reactivity swing and therefore plummeting the safety factor of the reactor. (author)

  18. Power ramp tests of BWR-MOX fuels

    International Nuclear Information System (INIS)

    Asahi, K.; Oguma, M.; Higuchi, S.; Kamimua, K.; Shirai, Y.; Bodart, S.; Mertens, L.

    1996-01-01

    Power ramp test of BWR-MOX and UO 2 fuel rods base irradiated up to about 60 GWd/t in Dodewaard reactor have been conducted in BR2 reactor in the framework of the international DOMO programme. The MOX pellets were provided by BN (MIMAS process) and PNC (MH method). The MOX fuel rods with Zr-liner and non-liner cladding and the UO 2 fuel rods with Zr-liner cladding remained intact during the stepwise power ramp tests to about 600 W/cm, even at about 60 GWd/t

  19. Corrosion failure of a BWR embedded reactor containment liner

    International Nuclear Information System (INIS)

    Wegemar, B.

    2006-01-01

    Following sixteen fuel cycles, leakage through a BWR embedded reactor containment liner (carbon steel) was discovered. Leakage was located at a penetration for electrical conductors as a result of penetrating corrosion attack. During construction, porous cement structures and air pockets/cavities were formed due to erroneous injection of grout. Corrosion attacks on the CS steel liner were located at the relatively small, active surfaces in contact with the porous cement structure. The corrosion mechanism was supposed to be anodic dissolution of the steel liner in areas with insufficient passivation. The penetrations were restored according to original design requirements. (author)

  20. Time-dependent coolant velocity measurements in an operating BWR

    International Nuclear Information System (INIS)

    Luebbesmeyer, D.; Crowe, R.D.

    1980-01-01

    A method to measure time-dependent fluid velocities in BWR-bundle elements by noise analysis of the incore-neutron-detector signals is shown. Two application examples of the new method are given. The time behaviour of the fluid velocity in the bundle element during a scheduled power excursion of the plant. The change of power was performed by changing the coolant flow through the core The apparent change of the fluid velocity due to thermal elongation of the helix-drive of the TIP-system. A simplified mathematical model was derived for this elongation to use as a reference to check the validity of the new method. (author)

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

    International Nuclear Information System (INIS)

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

    1991-01-01

    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)

  2. The mechanical structure of the SVEA BWR fuel

    International Nuclear Information System (INIS)

    Nylund, O.; Johansson, A.; Junkrans, S.

    1985-01-01

    The SVEA BWR fuel assembly design is characterized by a double-wall cruciform internal structure forming an internal water gap and dividing the assembly into 4 subbundles. The effect is a favourable distribution of fuel and moderator, a minimum amount of structural material in active core, a combination of structural stability and flexibility for minimum control rod friction in reduced gaps and a reduced creep deformation of the fuel assembly. The results of a laboratory test program confirm the much lower friction force obtained with the SVEA fuel assemblies while withdrawing and inserting the control rod. (RF)

  3. In-situ testing of BWR closure head studs

    International Nuclear Information System (INIS)

    deRaad, J.A.; Wolters, J.T.

    1988-01-01

    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

  4. Decontamination and materials corrosion concerns in the BWR

    International Nuclear Information System (INIS)

    Gordon, B.M.; Gordon, G.M.

    1988-01-01

    The qualification of chemical decontamination processes to decontaminate complete systems or individual components in essential if effective inspection, maintenance, repair or replacement of plant components is to be achieved with minimum exposure of workers to ionizing radiation. However, it is critical that the benefits of decontamination processes are not overshadowed by deleterious materials/ corrosion side effects during the application of the process or during subsequent operation. This paper discusses such potential corrosion/materials problems in the BWR and presents relevant available corrosion data for the various commercial decontamination processes. (author)

  5. Characteristics of axial splits in failed BWR fuel rods

    International Nuclear Information System (INIS)

    Lysell, G.; Grigoriev, V.

    2000-01-01

    Secondary cladding defects in BWR fuel sometimes have the shape of long axial cracks or ''splits''. Due to the large open UO 2 surfaces exposed to the water, fission product and UO 2 release to the coolant can reach excessive levels leading to forced shut downs to remove the failed fuel rods. A number of such fuel rods have been examined in Studsvik over the last 10 years. The paper describes observations from the PIE of long cracks and discusses the driving force of the cracks. Details such as starting cracks, macroscopic and microscopic fracture surface appearance, cross sections of cracks, hydride precipitates, location and degree of plastic deformation are given. (author)

  6. Development of water chemistry diagnosis system for BWR primary loop

    International Nuclear Information System (INIS)

    Nagase, Makoto; Asakura, Yamato; Sakagami, Masaharu; Uchida, Shunsuke; Ohsumi, Katsumi.

    1988-01-01

    The prototype of a water chemistry diagnosis system for BWR primary loop has been developed. Its purposes are improvement of water chemistry control and reduction of the work burden on plant chemistry personnel. It has three main features as follows. (1) Intensifying the observation of water chemistry conditions by variable sampling intervals based on the on-line measured data. (2) Early detection of water chemistry data trends using a second order regression curve which is calculated from the measured data, and then searching the cause of anomaly if anything (3) Diagnosis of Fe concentration in feedwater using model simulations, in order to lower the radiation level in the primary system. (author)

  7. The noise analysis and the BWR operation map

    International Nuclear Information System (INIS)

    Blazquez, J.; Ballestrin, J.

    1996-01-01

    An analytical expression for the Decay Ratio is obtained: DR = exp(-bW / P 1/2 ). The physics behind is also explained. It applies to a commercial BWR Operation Map, on the vicinity of the power instability. This functional form seems fitting to the structure of the Operation map. The power P and the coolant flow are measured straightforward; the Decay Ratio is obtained by neutron noise analysis techniques. The parameter b, depending on the void reactivity coefficient, is then calculated on line during the Reactor Operation. New DR value is now predicted for each new displacement on the Map, so unexpected instability events are more likely avoided. (authors)

  8. 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 ...

  9. 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)

  10. BWR power oscillation evaluation methodologies in core design

    International Nuclear Information System (INIS)

    Hotta, Akitoshi

    1995-01-01

    At the initial stage of BWR development, the power oscillation due to the nuclear-thermal interaction originated in random boiling phenomena and nuclear void feedback was feared. But it was shown that under the high pressure condition in the normal operation of recent commercial BWRs, the core is in very stable state. However, power oscillation events have been observed in actual machines, and it is necessary to do the stability evaluation that sufficiently reflects the detailed operation conditions of actual plants. As the cause of power oscillation events, the instability of control system and nuclear-thermal coupling instability are important, and their mechanisms are explained. As the model for analyzing the stability of BWR core, the nuclear-thermal coupling model in frequency domain is the central existence. As the information for the design, the parameters of fuel assemblies, and the nuclear parameters and the thermohydraulic parameters of cores are enumerated. LAPUR-TSI is a nuclear-thermal coupling model. The analysis system in the software of Tokyo Electric Power Co. is outlined, and the analysis model was verified. (K.I.)

  11. Examination of minor actinide annihilation by BWR core

    International Nuclear Information System (INIS)

    Hida, Kazuki

    1995-01-01

    From the viewpoint of reducing burden for disposing high level waste generated from spent fuel, the examination of recycling minor actinide (MA) to reactors and reducing its accumulation has been advanced. In this study, the possibility of annihilation in the case of recycling it to a BWR was examined. The main MAs are 237 Np, 241 Am, 243 Am, 242 Cm, and 244 Cm. However, as for Cm isotopes, the half life is short, the amount of generation is small, and the rate of neutron emission is high, therefore, those are disposed as waste, and 237 Np, 241 Am and 243 Am were taken as the objects of recycling. In order to grasp the basic characteristics in the case of recycling MAs to a BWR, MAs were added to UO 2 fuel, MOX fuel and HCR fuel and burned, and the nuclear conversion characteristics were examined. As the result, it was found that they were converted to short half life nuclides, and as the neutron spectra were softer, the rate of annihilation was higher. In the case of recycling MAs by concentrating to a specific reactor, reactivity loss, the degree of uranium enrichment required for compensating reactivity, and the rate of MA annihilation were calculated. Based on these data, the MA recycling system was set up, and the rate of MA annihilation was evaluated. This is reported. (K.I.)

  12. BWR containments license renewal industry report; revision 1. Final report

    International Nuclear Information System (INIS)

    Smith, S.; Gregor, F.

    1994-07-01

    The U.S. nuclear power industry, through coordination by the Nuclear Management and Resources Council (NUMARC), and sponsorship by the U.S. Department of Energy (DOE) and the Electric Power Research Institute (EPRI), has evaluated age-related degradation effects for a number of major plant systems, structures, and components, in the license renewal technical Industry Reports (IR's). License renewal applicants may choose to reference these IR's in support of their plant-specific license renewal applications as an equivalent to the integrated plant assessment provisions of the license renewal rule (IOCFR54). The scope of the IR provides the technical basis for license renewal for U.S. Boiling Water Reactor (BWR) containments. The scope of the report includes containments constructed of reinforced or prestressed concrete with steel liners and freestanding stell containments. Those domestic BWR containments designated as Mark I, Mark II or Mark III are covered, but no containments are addressed before these designs. The report includes those items within the jurisdictional boundaries for metal and concrete containments defined by Section III of the ASME Boiler and Pressure Vessel Code, Division 1, Subsection NE (Class MC) and Division 2 (Class CC) and their supports, but excluding snubbers

  13. Appraisal of BWR plutonium burners for energy centers

    International Nuclear Information System (INIS)

    Williamson, H.E.

    1976-01-01

    The design of BWR cores with plutonium loadings beyond the self-generation recycle (SGR) level is investigated with regard to their possible role as plutonium burners in a nuclear energy center. Alternative plutonium burner approaches are also examined including the substitution of thorium for uranium as fertile material in the BWR and the use of a high-temperature gas reactor (HTGR) as a plutonium burner. Effects on core design, fuel cycle facility requirements, economics, and actinide residues are considered. Differences in net fissile material consumption among the various plutonium-burning systems examined were small in comparison to uncertainties in HTGR, thorium cycle, and high plutonium-loaded LWR technology. Variation in the actinide content of high-level wastes is not likely to be a significant factor in determining the feasibility of alternate systems of plutonium utilization. It was found that after 10,000 years the toxicity of actinide high-level wastes from the plutonium-burning fuel cycles was less than would have existed if the processed natural ores had not been used for nuclear fuel. The implications of plutonium burning and possible future fuel cycle options on uranium resource conservation are examined in the framework of current ERDA estimates of minable uranium resources

  14. 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.

  15. Managing the aging of BWR control rod drive systems

    International Nuclear Information System (INIS)

    Greene, R.H.; Farmer, W.S.

    1992-01-01

    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

  16. Aging assessment of BWR control rod drive systems

    International Nuclear Information System (INIS)

    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

  17. Sophistication of operator training using BWR plant simulator

    International Nuclear Information System (INIS)

    Ohshiro, Nobuo; Endou, Hideaki; Fujita, Eimitsu; Miyakita, Kouji

    1986-01-01

    In Japanese nuclear power stations, owing to the improvement of fuel management, thorough maintenance and inspection, and the improvement of facilities, high capacity ratio has been attained. The thorough training of operators in nuclear power stations also contributes to it sufficiently. The BWR operator training center was established in 1971, and started the training of operators in April, 1974. As of the end of March, 1986, more than 1800 trainees completed training. At present, in the BWR operator training center, No.1 simulator of 800 MW class and No.2 simulator of 1100 MW class are operated for training. In this report, the method, by newly adopting it, good result was obtained, is described, that is, the method of introducing the feeling of being present on the spot into the place of training, and the new testing method introduced in retraining course. In the simulator training which is apt to place emphasis on a central control room, the method of stimulating trainees by playing the part of correspondence on the spot and heightening the training effect of multiple monitoring was tried, and the result was confirmed. The test of confirmation on the control board was added. (Kako, I.)

  18. Radiation buildup and control in BWR recirculation piping

    International Nuclear Information System (INIS)

    Meyer, W.; Wood, R.M.; Rao, T.V.; Vook, R.W.

    1987-01-01

    Boiling water nuclear reactors (BWRs) employ stainless steel (Types 304 or 316 NG) pipes in which high-purity water at temperatures of ∼ 275 0 C are circulated. Various components of the system, such as valves and bearings, often contain hard facing metal alloys such as Stellite-6. These components, along with the stainless steel tubing and feedwater, serve as sources of 59 Co. This cobalt, along with other soluble and insoluble impurities, is carried along with the circulating water to the reactor core where it is converted to radioactive 60 Co. After reentering the circulating water, the 60 Co can be incorporated into a complex corrosion layer in the form of CoCr 2 O 4 and/or CoFe 2 O 4 . The presence of even small amounts of 60 Co on the walls of BWR cooling systems is the dominant contributor to inplant radiation levels. Thus BWR owners and their agents are expending significant time and resources in efforts to reduce both the rate and amount of 60 Co buildup. The object of this research is twofold: (a) to form a thin diffusion barrier against the outward migration of cobalt from a cobalt-containing surface and (b) to prevent the growth of a 60 Co-containing corrosion film. The latter goal was the more important since most of the radioactive cobalt will originate from sources other than the stainless steel piping itself

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

    Energy Technology Data Exchange (ETDEWEB)

    Perry, D.R.

    1978-09-01

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

  20. Aging assessment of BWR control rod drive systems

    International Nuclear Information System (INIS)

    Greene, R.H.

    1992-01-01

    This Phase 1 Nuclear Plant Aging Research (NPAR) study examines the aging phenomena associated with boiling water reactor (BWR) control rod drive mechanisms (CRDMs) and assesses 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 NPRDS failure cases attributed to the CRD system, and (4) personal information exchange. As part of this study, nearly 3,500 NPRDS failure reports have been analyzed to examine the prevailing failure trends for CRD system components. An investigation has been conducted that summarizes 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 regarding specific actions that utilities can implement to mitigate CRDM aging. An evaluation has also been made of certain practices and tooling which have enabled some utilities to reduce ALARA exposures received from routine CRDM replacement and rebuilding activities

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

    International Nuclear Information System (INIS)

    1992-09-01

    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

  2. A study of heat capacity temperature limit of BWR

    International Nuclear Information System (INIS)

    Wang, Shih-Jen; Chen, Jyh-Jun; Chien, Chun-Sheng; Teng, Jyh-Tong

    2012-01-01

    Highlights: ► The purpose of this study is to verify the HCTL. ► MAAP4 was used as code to generate a realistic and convenient HCTL. ► The current HCTL curve causes confusing in reading data. ► The revised HCTL curves developed in this study. ► Users can obtain important parameters from the revised HCTL without confusion and interpolation. - Abstract: Heat capacity temperature limit (HCTL) is an important parameter for operation of BWR. Current version of the HCTL was derived, based on simple model of computation aids (CA) of BWR owners’ group (BWROG). However, some parts of the current HCTL are confusing to the users in reading data. The purpose of this study is to verify the HCTL by applying the MAAP4 code to the field of emergency operating procedure (EOP). The trends of HCTL generated by MAAP4 code are consistent with those obtained from CA. A series of revised HCTL evaluated at various times after scram are provided and the confusing part is eliminated.

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

    International Nuclear Information System (INIS)

    1992-09-01

    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

  4. System control model of a turbine for a BWR

    International Nuclear Information System (INIS)

    Vargas O, Y.; Amador G, R.; Ortiz V, J.; Castillo D, R.; Delfin L, A.

    2009-10-01

    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)

  5. Transmutation of minor actinide using BWR fueled mixed oxide

    International Nuclear Information System (INIS)

    Susilo, Jati

    2000-01-01

    Nuclear spent fuel recycle has a strategic importance in the aspect of nuclear fuel economy and prevention of its spread-out. One among other application of recycle is to produce mixed oxide fuel (Mo) namely mixed Plutonium and uranium oxide. As for decreasing the burden of nuclear high level waste (HLW) treatment, transmutation of minor actinide (MA) that has very long half life will be carried out by conversion technique in nuclear reactor. The purpose of this study was to know influence of transition fuel cell regarding the percent weight of transmutation MA in the BWR fueled MOX. Calculation of cell BWR was used SRAC computer code, with assume that the reactor in equilibrium. The percent weight of transmutation MA to be optimum by increasing the discharge burn-up of nuclear fuel, raising ratio of moderator to fuel volume (Vm/Vf), and loading MA with percent weight about 3%-6% and also reducing amount of percent weight Pu in MOX fuel. For mixed fuel standard reactor, reactivity value were obtained between about -50pcm ∼ -230pcm for void coefficient and -1.8pcm ∼ -2.6pcm for fuel temperature coefficient

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

    International Nuclear Information System (INIS)

    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

  7. Evaluation of thermal margin during BWR neutron flux oscillation

    International Nuclear Information System (INIS)

    Takeuchi, Yutaka; Takigawa, Yukio; Chuman, Kazuto; Ebata, Shigeo

    1992-01-01

    Fuel integrity is very important, from the view point of nuclear power plant safety. Recently, neutron flux oscillations were observed at several BWR plants. The present paper describes the evaluations of the thermal margin during BWR neutron flux oscillations, using a three-dimensional transient code. The thermal margin is evaluated as MCPR (minimum critical power ratio). The LaSalle-2 event was simulated and the MCPR during the event was evaluated. It was a core-wide oscillation, at which a large neutron flux oscillation amplitude was observed. The results indicate that the MCPR had a sufficient margin with regard to the design limit. A regional oscillation mode, which is different from a core-wide oscillation, was simulated and the MCPR response was compared with that for the LaSalle-2 event. The MCPR decrement is greater in the regional oscillation, than in the core wide -oscillation, because of the sensitivity difference in a flow-to-power gain. A study was carried out about regional oscillation detectability, from the MCPR response view point. Even in a hypothetically severe case, the regional oscillation is detectable by LPRM signals. (author)

  8. Safety evaluation of BWR off-gas treatment systems

    International Nuclear Information System (INIS)

    Schultz, R.J.; Schmitt, R.C.

    1975-01-01

    Some of the results of a safety evaluation performed on current generic types of BWR off-gas treatment systems including cooled and ambient temperature adsorber beds and cryogenics are presented. The evaluation covered the four generic types of off-gas systems and the systems of five major vendors. This study was part of original work performed under AEC contract for the Directorate of Regulatory Standards. The analysis techniques employed for the safety evaluation of these systems include: Fault Tree Analysis; FMECA (Failure Mode Effects and Criticality Analysis); general system comparisons, contaminant, system control, and design adequacy evaluations; and resultant Off-Site Dose Calculations. The salient areas presented are some of the potential problem areas, the approach that industry has taken to mitigate or design against potential upset conditions, and areas where possible deficiencies still exist. Potential problem areas discussed include hydrogen detonation, hydrogen release to equipment areas, operator/automatic control interface, and needed engineering evaluation to insure safe system operation. Of the systems reviewed, most were in the category of advanced or improved over that commonly in use today, and a conclusion from the study was that these systems offer excellent potential for noble gas control for BWR power plants where more stringent controls may be specified -- now or in the future. (U.S.)

  9. Aging assessment of BWR control rod drive systems

    International Nuclear Information System (INIS)

    Greene, R.H.

    1991-01-01

    This Phase 1 Nuclear Plant Aging Research (NPAR) study examines the aging phenomena associated with boiling water reactor (BWR) control rod drive mechanisms (CRDMs) and assesses 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 NPRDS failure cases attributed to the CRD system, and (4) personal information exchange. As part of this study, nearly 3,500 NPRDS failure reports have been analyzed to examine the prevailing failure trends for CRD system components. An investigation has been conducted that summarizes 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 regarding specific actions that utilities can implement to mitigate CRDM aging. An evaluation has also been made of certain practices and tooling which have enabled some utilities to reduce ALARA exposures received from routine CRDM replacement and rebuilding activities

  10. An optimized BWR fuel lattice for improved fuel utilization

    International Nuclear Information System (INIS)

    Bernander, O.; Helmersson, S.; Schoen, C.G.

    1984-01-01

    Optimization of the BWR fuel lattice has evolved into the water cross concept, termed ''SVEA'', whereby the improved moderation within bundles augments reactivity and thus improves fuel cycle economy. The novel design introduces into the assembly a cruciform and double-walled partition containing nonboiling water, thus forming four subchannels, each of which holds a 4x4 fuel rod bundle. In Scandinavian BWRs - for which commercial SVEA reloads are now scheduled - the reactivity gain is well exploited without adverse impact in other respects. In effect, the water cross design improves both mechanical and thermal-hydraulic performance. Increased average burnup is also promoted through achieving flatter local power distributions. The fuel utilization savings are in the order of 10%, depending on the basis of comparison, e.g. choice of discharge burnup and lattice type. This paper reviews the design considerations and the fuel utilization benefits of the water cross fuel for non-Scandinavian BWRs which have somewhat different core design parameters relative to ASEA-ATOM reactors. For one design proposal, comparisons are made with current standard 8x8 fuel rod bundles as well as with 9x9 type fuel in reactors with symmetric or asymmetric inter-assembly water gaps. The effect on reactivity coefficients and shutdown margin are estimated and an assessment is made of thermal-hydraulic properties. Consideration is also given to a novel and advantageous way of including mixed-oxide fuel in BWR reloads. (author)

  11. A study of heat capacity temperature limit of BWR

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Shih-Jen, E-mail: sjenwang@iner.gov.tw [Institute of Nuclear Energy Research (INER), 1000, Wunhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China); Chen, Jyh-Jun [Department of Mechanical Engineering, Chung Yuan Christian University, 200, Chung Pei Rd., Chung Li City, Taoyuan County 32023, Taiwan (China); Chien, Chun-Sheng [Institute of Nuclear Energy Research (INER), 1000, Wunhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China); Teng, Jyh-Tong [Department of Mechanical Engineering, Chung Yuan Christian University, 200, Chung Pei Rd., Chung Li City, Taoyuan County 32023, Taiwan (China)

    2012-02-15

    Highlights: Black-Right-Pointing-Pointer The purpose of this study is to verify the HCTL. Black-Right-Pointing-Pointer MAAP4 was used as code to generate a realistic and convenient HCTL. Black-Right-Pointing-Pointer The current HCTL curve causes confusing in reading data. Black-Right-Pointing-Pointer The revised HCTL curves developed in this study. Black-Right-Pointing-Pointer Users can obtain important parameters from the revised HCTL without confusion and interpolation. - Abstract: Heat capacity temperature limit (HCTL) is an important parameter for operation of BWR. Current version of the HCTL was derived, based on simple model of computation aids (CA) of BWR owners' group (BWROG). However, some parts of the current HCTL are confusing to the users in reading data. The purpose of this study is to verify the HCTL by applying the MAAP4 code to the field of emergency operating procedure (EOP). The trends of HCTL generated by MAAP4 code are consistent with those obtained from CA. A series of revised HCTL evaluated at various times after scram are provided and the confusing part is eliminated.

  12. A detailed BWR recirculation loop model for RELAP

    Energy Technology Data Exchange (ETDEWEB)

    Araiza-Martínez, Enrique, E-mail: enrique.araiza@inin.gob.mx; Ortiz-Villafuerte, Javier, E-mail: javier.ortiz@inin.gob.mx; Castillo-Durán, Rogelio, E-mail: rogelio.castillo@inin.gob.mx

    2017-01-15

    Highlights: • A new detailed BWR recirculation loop model was developed for RELAP. • All jet pumps, risers, manifold, suction and control valves, and recirculation pump are modeled. • Model is tested against data from partial blockage of two jet pumps. • For practical applications, simulation results showed good agreement with available data. - Abstract: A new detailed geometric model of the whole recirculation loop of a BWR has been developed for the code RELAP. This detailed model includes the 10 jet pumps, 5 risers, manifold, suction and control valves, and the recirculation pump, per recirculation loop. The model is tested against data from an event of partial blockage at the entrance nozzle of one jet pump in both recirculation loops. For practical applications, simulation results showed good agreement with data. Then, values of parameters considered as figure of merit (reactor power, dome pressure, core flow, among others) for this event are compared against those from the common 1 jet pump per loop model. The results show that new detailed model led to a closer prediction of the reported power change. The detailed recirculation loop model can provide more reliable boundary condition data to a CFD models for studies of, for example, flow induced vibration, wear, and crack initiation.

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  15. Operational experiences with on line BWR condenser tube leak verification

    International Nuclear Information System (INIS)

    Bryant, R.A.; Duvall, W.E.; Kirkley, W.B.; Zavadoski, R.W.

    1988-01-01

    Verifying condenser tube leaks at a boiling water reactor is, at best, a difficult task carried out in hot steamy water boxes with concurrent radiation exposure. For small apparent leaks with slight chemical changes there is always uncertainty of whether the problem is a condenser tube leak or a feedback from radwaste. Most conventional methods (e.g soap tests, Saran wrap suction, and helium tests) usually involve a load reduction to isolate the water boxes one at a time and hours of drain down on each box. The sensitivity of the most sensitive test (helium) is of the order of 7500 l per day per box. Sulfur hexafluoride has been successfully used at a BWR to identify one leaking water box out of four while the unit was at 100 % power. The actual tubes leakig in the water box were identified by injecting helium during drain down of the box and subsequent manifold testing. Additional tests with sulfur hexafluoride on the second BWR unit indicated tight water boxes to within the sensitivity of the measurement, i.e. less than 19 l per day for all four boxes. Problems encountered in both tests included sulfur hexafluoride carry over from the plume of the cooling towers and off gas considerations. In brief sulfur hexafluoride can be used to quickly identify which particular water box has a condenser tube leak or, just as quickly, establish the integrity of all the water boxes to a level not previously attainable. (author)

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

    International Nuclear Information System (INIS)

    Rajamaeki, Markku.

    1980-03-01

    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)

  17. Logical model for the control of a BWR turbine

    International Nuclear Information System (INIS)

    Vargas O, Y.; Amador G, R.; Ortiz V, J.; Castillo D, R.

    2009-01-01

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

  18. Exosomal MicroRNA MiR-1246 Promotes Cell Proliferation, Invasion and Drug Resistance by Targeting CCNG2 in Breast Cancer

    Directory of Open Access Journals (Sweden)

    Xiu Juan Li

    2017-12-01

    Full Text Available Background/Aims: Treatment of breast cancer remains a clinical challenge. This study aims to validate exosomal microRNA-1246 (miR-1246 as a serum biomarker for breast cancer and understand the underlying mechanism in breast cancer progression. Methods: The expression levels of endogenous and exosomal miRNAs were examined by real time PCR, and the expression level of the target protein was detected by western blot. Scanning electron and confocal microscopy were used to characterize exosomes and to study their uptake and transfer. Luciferase reporter plasmids and its mutant were used to confirm direct targeting. Furthermore, the functional significance of exosomal miR-1246 was estimated by invasion assay and cell viability assay. Results: In this study, we demonstrate that exosomes carrying microRNA can be transferred among different cell lines through direct uptake. miR-1246 is highly expressed in metastatic breast cancer MDA-MB-231 cells compared to non-metastatic breast cancer cells or non-malignant breast cells. Moreover, miR-1246 can suppress the expression level of its target gene, Cyclin-G2 (CCNG2, indicating its functional significance. Finally, treatment with exosomes derived from MDA-MB-231 cells could enhance the viability, migration and chemotherapy resistance of non-malignant HMLE cells. Conclusions: Together, our results support an important role of exosomes and exosomal miRNAs in regulating breast tumor progression, which highlights their potential for applications in miRNA-based therapeutics.

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

    International Nuclear Information System (INIS)

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

    1980-06-01

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

  20. Strengthening the non proliferation regime: French views

    International Nuclear Information System (INIS)

    Delaune, P.

    2013-01-01

    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.)

  1. Exosomal MicroRNA MiR-1246 Promotes Cell Proliferation, Invasion and Drug Resistance by Targeting CCNG2 in Breast Cancer.

    Science.gov (United States)

    Li, Xiu Juan; Ren, Zhao Jun; Tang, Jin Hai; Yu, Qiao

    2017-01-01

    Treatment of breast cancer remains a clinical challenge. This study aims to validate exosomal microRNA-1246 (miR-1246) as a serum biomarker for breast cancer and understand the underlying mechanism in breast cancer progression. The expression levels of endogenous and exosomal miRNAs were examined by real time PCR, and the expression level of the target protein was detected by western blot. Scanning electron and confocal microscopy were used to characterize exosomes and to study their uptake and transfer. Luciferase reporter plasmids and its mutant were used to confirm direct targeting. Furthermore, the functional significance of exosomal miR-1246 was estimated by invasion assay and cell viability assay. In this study, we demonstrate that exosomes carrying microRNA can be transferred among different cell lines through direct uptake. miR-1246 is highly expressed in metastatic breast cancer MDA-MB-231 cells compared to non-metastatic breast cancer cells or non-malignant breast cells. Moreover, miR-1246 can suppress the expression level of its target gene, Cyclin-G2 (CCNG2), indicating its functional significance. Finally, treatment with exosomes derived from MDA-MB-231 cells could enhance the viability, migration and chemotherapy resistance of non-malignant HMLE cells. Together, our results support an important role of exosomes and exosomal miRNAs in regulating breast tumor progression, which highlights their potential for applications in miRNA-based therapeutics. © 2017 The Author(s). Published by S. Karger AG, Basel.

  2. Energy efficiency and proliferation assessment factors

    International Nuclear Information System (INIS)

    1979-02-01

    The objective of INFCE is to evaluate the nuclear fuel cycles from the point of view of their ability to satisfy the worldwide nuclear energy needs, while minimizing the proliferation risks. Accordingly, the different working groups have to take into consideration as well the energy-efficiency and the proliferation-resistance of these nuclear fuel cycles. The present working paper is aimed at suggesting the main assessment factors which should be taken into consideration

  3. 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)

  4. 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)

  5. FIST small break accident analysis with BWR TRACBO2-pretest predictions

    International Nuclear Information System (INIS)

    Alamgir, M.; Sutherland, W.A.

    1984-01-01

    The BWR Full Integral Simulation Test (FIST) program includes experimental simulation and analytical evaluation of BWR thermal-hydraulic phenomena during transient events. One such event is a small size break in the suction line of one of the recirculation pumps. The results from a test simulating this transient in the FIST facility are compared with a system analysis using the Transient Reactor Analysis Code (TRACB02). This comparison demonstrates BWR-TRAC capability for small break analyses and provides detailed understanding of the phenomena

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

    International Nuclear Information System (INIS)

    Stellwag, B.; Laendner, A.; Weiss, S.; Huettner, F.

    2010-01-01

    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)

  7. Application of TRAC-BD1/MOD1 to a BWR/4 feedwater control failure ATWS

    International Nuclear Information System (INIS)

    Rouhani, S.Z.; Giles, M.M.; Mohr, C.M. Jr.; Weaver, W.L. III.

    1984-01-01

    This paper begins with a short description of the Transient Reactor Analysis Code for Boiling Water Reactors (TRAC-BWR), briefly mentioning some of its main features such as specific BWR models and input structure. Next, an input model of a BWR/4 is described, and, the assumptions used in performing an analysis of the loss of a feedwater controller without scram are listed. The important features of the calculated trends in flows, pressure, reactivity, and power are shown graphically and commented in the text. A comparison of some of the main predicted trends with the calculated results from a similar study by General Electric is also presented

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

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki

    1992-03-01

    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)

  9. The threat of proliferation

    International Nuclear Information System (INIS)

    Palme, Olof.

    1986-01-01

    The paper on the threat of proliferation, is a keynote speech delivered to the Colloquium on Nuclear War, Nuclear Proliferation and their Consequences, Geneva, 1985. Topics discussed in the address include: nuclear weapons, nuclear war, terrorists, Non-Proliferation Treaty, nuclear disarmament, and leadership in world affairs. (UK)

  10. Modeling of SCC initiation and propagation mechanisms in BWR environments

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmeister, Hans, E-mail: Hans.Hoffmeister@hsu-hh.de [Institute for Failure Analysis and Failure Prevention ISSV e.V., c/o Helmut Schmidt University of the Federal Armed Forces, D-22039 Hamburg (Germany); Klein, Oliver [Institute for Failure Analysis and Failure Prevention ISSV e.V., c/o Helmut Schmidt University of the Federal Armed Forces, D-22039 Hamburg (Germany)

    2011-12-15

    Highlights: Black-Right-Pointing-Pointer We show that SSC in BWR environments includes anodic crack propagation and hydrogen assisted cracking. Black-Right-Pointing-Pointer Hydrogen cracking is triggered by crack tip acidification following local impurity accumulations and subsequent phase precipitations. Black-Right-Pointing-Pointer We calculate effects of pH, chlorides, potentials and stress on crack SCC growth rates at 288 Degree-Sign C. - Abstract: During operation of mainly BWRs' (Boiling Water Reactors) excursions from recommended water chemistries may provide favorite conditions for stress corrosion cracking (SCC). Maximum levels for chloride and sulfate ion contents for avoiding local corrosion are therefore given in respective water specifications. In a previously published deterministic 288 Degree-Sign C - corrosion model for Nickel as a main alloying element of BWR components it was demonstrated that, as a theoretically worst case, bulk water chloride levels as low as 30 ppb provide local chloride ion accumulation, dissolution of passivating nickel oxide and precipitation of nickel chlorides followed by subsequent local acidification. In an extension of the above model to SCC the following work shows that, in a first step, local anodic path corrosion with subsequent oxide breakdown, chloride salt formation and acidification at 288 Degree-Sign C would establish local cathodic reduction of accumulated hydrogen ions inside the crack tip fluid. In a second step, local hydrogen reduction charges and increasing local crack tip strains from increasing crack lengths at given global stresses are time stepwise calculated and related to experimentally determined crack critical cathodic hydrogen charges and fracture strains taken from small scale SSRT tensile tests pieces. As a result, at local hydrogen equilibrium potentials higher than those of nickel in the crack tip solution, hydrogen ion reduction initiates hydrogen crack propagation that is enhanced with

  11. Future non-proliferation challenges

    International Nuclear Information System (INIS)

    Yelchenko, Volodymyr

    2008-01-01

    activities and illicit trafficking of nuclear material, equipment and technology. The strengthening of the physical protection of nuclear materials and facilities as an element of non-proliferation regime was highlighted. The proliferation risks associated with the growing global energy demand were also mentioned. In this regard, attention was drawn to the significance of developing proliferation resistant nuclear technologies, including through the international project on Innovative Nuclear Reactors and Fuel Cycles (INPRO). It is important to mention the discussion concerning the promotion of multilateralism in the nuclear fuel cycle and the supply of nuclear fuel, which was considered as a significant contribution to confidence-building in the field of non proliferation, to peaceful uses of nuclear energy and to the overall strength of the non-proliferation regime. The ongoing discussions in IAEA on fuel supply assurances mechanism were welcomed. The NPT relative success in slowing proliferation has been attributed to the combination of the political commitment by most states to the objective of non-proliferation, and a technical mechanism - IAEA safeguards - for verifying that this commitment is being honoured. Indeed, the IAEA will remain an indispensable part of the multilateral nuclear non proliferation regime and global security system and its role should be significantly strengthened. He hopes that the meeting will help to identify the ways of enhancing the important work of IAEA

  12. Application of a simplified seismic risk methodology to the La Salle County Station Unit 2 BWR

    International Nuclear Information System (INIS)

    Lappa, D.A.; Wells, J.E.

    1986-01-01

    It is important to bear in mind that no risk assessment of any U.S. nuclear power plant can be interpreted to be generally representative of more than a handful of other U.S. plants. Variations in factors ranging from plant age and operating experience to NRC licensing requirements and design guidelines have led to a wide diversity of power plants in the United States. Except for a few combinations of plants of comparable design and vintage, the extension of plant-specific results to other nuclear power plants should only be done with considerable trepidation. This situation is worsened for a seismic PRA because of the variability in the seismic hazard from site to site. In the case of this study, it would be a mistake to infer that all BWRs are sufficiently resistant to earthquakes because of the generally low seismic failure probabilities at La Salle. Unless those BWRs had similar site characteristics and were of a similar design and vintage as La Salle, no immediate extension of this study's results would be appropriate. With these thoughts in mind, we turn our attention to one of the questions which the La Salle seismic PRA is supposed to address, namely, the comparable seismic vulnerability of BWRs and PWRs. The La Salle study has provided us with some insight to the seismic risk at a particular BWR. This information may or may not be useful to understanding the seismic vulnerability of other BWRs

  13. 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

  14. 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.

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

    International Nuclear Information System (INIS)

    Vallejo Q, J. A.; Martin del Campo M, C.; Fuentes M, L.; Francois L, J. L.

    2015-09-01

    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)

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

    International Nuclear Information System (INIS)

    Kouji, Hiraiwa; Noriyuki, Yoshida; Mikihide, Nakamaru; Hideaki, Heki; Masanori, Aritomi

    2002-01-01

    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. Cobra-TF simulation of BWR bundle dry out experiments

    Energy Technology Data Exchange (ETDEWEB)

    Frepoli, C.; Ireland, A.; Hochreiter, L.; Ivanov, K. [Penn State Univ., Dept. of Mechanical and Nuclear Engineering, University Park, PA (United States); Velten, R. [Siemens Nuclear Power GmbH, Erlangen (Germany)

    2001-07-01

    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)

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

    International Nuclear Information System (INIS)

    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

  19. Fault tree analysis on BWR core spray system

    International Nuclear Information System (INIS)

    Watanabe, Norio

    1982-06-01

    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. Development of methodology for early detection of BWR instabilities

    International Nuclear Information System (INIS)

    Alessandro Petruzzi; Shin Chin; Kostadin Ivanov; Asok Ray; Fan-Bill Cheung

    2005-01-01

    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

  1. Investigation of detector tube impacting in the Ringhals-1 BWR

    Energy Technology Data Exchange (ETDEWEB)

    Sunde, C.; Pazsit, I. [Department of Nuclear Engineering, Chalmers University of Technology, SE-412 96 Goteborg (Sweden)]. E-mail: kalle@nephy.chalmers.se; imre@nephy.chalmers.se

    2006-07-01

    Neutron noise measurements were made in two consecutive fuel cycles in the Swedish BWR Ringhals-1 with the purpose of diagnostics of vibrations and impacting of detector strings. Two diagnostic tools were used, first a traditional spectral analysis and second a wavelet-based method. Both types of methods have been used in the past, but not simultaneously during one fuel cycle. In addition, a new method, wavelet-based coherence, was tested with success. Based on the results of the analysis, with emphasis on the traditional method, the detector tubes were divided into three groups with respect to the severity and likelihood of impacting. For the first series of measurements, these conclusions could be checked against visual inspection of the fuel assemblies during refuelling after the cycle, in order to find impacting damage. A good correlation between the prediction of the analysis and the inspection results was found. (author)

  2. Investigation of detector tube impacting in the Ringhals-1 BWR

    International Nuclear Information System (INIS)

    Sunde, C.; Pazsit, I.

    2006-01-01

    Neutron noise measurements were made in two consecutive fuel cycles in the Swedish BWR Ringhals-1 with the purpose of diagnostics of vibrations and impacting of detector strings. Two diagnostic tools were used, first a traditional spectral analysis and second a wavelet-based method. Both types of methods have been used in the past, but not simultaneously during one fuel cycle. In addition, a new method, wavelet-based coherence, was tested with success. Based on the results of the analysis, with emphasis on the traditional method, the detector tubes were divided into three groups with respect to the severity and likelihood of impacting. For the first series of measurements, these conclusions could be checked against visual inspection of the fuel assemblies during refuelling after the cycle, in order to find impacting damage. A good correlation between the prediction of the analysis and the inspection results was found. (author)

  3. BWR Mark III pressure suppression containment response to hydrogen deflagration

    International Nuclear Information System (INIS)

    Fuls, G.M.; Gunter, A.D.

    1982-01-01

    The CLASIX-3 computer program has been used to evaluate the temperature and pressure response of the BWR Mark III Suppression Containment System to hydrogen deflagration resulting from a degraded core condition. The CLASIX-3 computer program is an extension of the CLASIX program which was originally developed to analyze ice condenser containments. A brief description is given of the modifications made to CLASIX to increase its flexibility and versatility to include the capability of analyzing the Mark III Containment. Analytical results are presented for the two base case transients. The two base cases are the stuck open steam relief valve and the small break LOCA, both of which are assumed to lead to a degraded core condition and the release of hydrogen to the containment. Results include pressure and temperature response, gas concentrations and suppression pool response

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

    International Nuclear Information System (INIS)

    Raety, H.; Rajamaeki, M.

    1991-05-01

    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

  5. Seismic response analysis of BWR buildings with embedded foundation

    International Nuclear Information System (INIS)

    Fukuzawa, R.; Chiba, O.; Tohdo, M.

    1985-01-01

    An investigation on the effect of various types of foundations embedded into soil deposit to the responses of BWR buildings is made. The frequency characteristics of massless rigid foundations such as impedance functions and input motion are first obtained, depending on the embedment depth, contact condition of side wall to soil and the thickness of surface layer in a two-layered soil. From the results it is found that the embedment depth and the thickness of surface layer have a distinct effect to the responses contributed by the impedance functions and the response values characterized by the input motions are obviously influenced by the embedment depth and the contact length of side wall to soil deposit. (orig.)

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

    International Nuclear Information System (INIS)

    Reparaz, A.; Smith, M.H.; Stephens, L.G.

    1992-01-01

    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

  7. BWR Mark I pressure suppression study: bench mark experiments

    International Nuclear Information System (INIS)

    Lai, W.; McCauley, E.W.

    1977-01-01

    Computer simulations representative of the wetwell of Mark I BWR's have predicted pressures and related phenomena. However, calculational predictions for purposes of engineering decision will be possible only if the code can be verified, i.e., shown to compute in accord with measured values. Described in the report is a set of single downcomer spherical flask bench mark experiments designed to produce quantitative data to validate various air-water dynamic computations; the experiments were performed since relevant bench mark data were not available from outside sources. Secondary purposes of the study were to provide a test bed for the instrumentation and post-experiment data processing techniques to be used in the Laboratory's reactor safety research program and to provide additional masurements for the air-water scaling study

  8. An ecological interface design for BWR nuclear power plants

    International Nuclear Information System (INIS)

    Monta, K.; Itoh, J.

    1992-01-01

    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)

  9. Flow rate control systems for coolants for BWR type reactors

    International Nuclear Information System (INIS)

    Igarashi, Yoko; Kato, Naoyoshi.

    1981-01-01

    Purpose: To increase spontaneous recycling flow rate of coolants in BWR type reactors when the water level in the reactor decreases, by communicating a downcomer with a lower plenum. Constitution: An opening is provided to the back plate disposed at the lower end of a reactor core shroud for communicating a downcomer with a lower plenum, and an ON-OFF valve actuated by an operation rod is provided to the opening. When abnormal water level or pressure in the reactor is detected by a level metal or pressure meter, the operation rod is driven to open the ON-OFF valve, whereby coolants fed from a jet pump partially flows through the opening to increase the spontaneous recycling flow rate of the coolants. This can increase the spontaneous recycling flow rate of the coolants upon spontaneous recycling operation, thereby maintaining the reactor safety and the fuel soundness. (Moriyama, K.)

  10. Availability analysis of United States BWR IV electrical generation plants

    International Nuclear Information System (INIS)

    Renick, D.H.; Li, F.; Todreas, N.E.

    1998-01-01

    Availability, as quantified by power output levels, from all active U.S. BWR IV plants were analyzed over a seven and a half year period to determine the operational characteristics of these plants throughout an operating cycle. The operational data were examined for infant mortality, end of cycle decreased availability, and seasonal availability variations. Scheduled outages were also examined to determine the industry's current approach to planning maintenance outages. The results of this study show that nuclear power plants do suffer significant infant mortality following a refueling outage. And while they do not suffer an end of cycle decrease in availability, a mid-cycle period of decreased availability is evident. This period of decreased availability is due to a combination of increased forced unavailability and seasonally scheduled maintenance and refueling outages. These findings form the start of a rational approach to increasing plant availability. (author)

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

    International Nuclear Information System (INIS)

    Castillo, A.; Ortiz, J.J.; Alonso, G.; Morales, L.B.; Valle, E. del

    2004-01-01

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

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

    International Nuclear Information System (INIS)

    Samson, P.; Neider, T.

    1999-01-01

    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 (