WorldWideScience

Sample records for bwr radiation buildup

  1. 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

  2. Optimum Water Chemistry in radiation field buildup control

    Energy Technology Data Exchange (ETDEWEB)

    Lin, Chien, C. [Vallecitos Nuclear Center, Pleasanton, CA (United States)

    1995-03-01

    Nuclear utilities continue to face the challenGE of reducing exposure of plant maintenance personnel. GE Nuclear Energy has developed the concept of Optimum Water Chemistry (OWC) to reduce the radiation field buildup and minimize the radioactive waste production. It is believed that reduction of radioactive sources and improvement of the water chemistry quality should significantly reduce both the radiation exposure and radwaste production. The most important source of radioactivity is cobalt and replacement of cobalt containing alloy in the core region as well as in the entire primary system is considered the first priority to achieve the goal of low exposure and minimized waste production. A plant specific computerized cobalt transport model has been developed to evaluate various options in a BWR system under specific conditions. Reduction of iron input and maintaining low ionic impurities in the coolant have been identified as two major tasks for operators. Addition of depleted zinc is a proven technique to reduce Co-60 in reactor water and on out-of-core piping surfaces. The effect of HWC on Co-60 transport in the primary system will also be discussed.

  3. Review of GEM Radiation Belt Dropout and Buildup Challenges

    Science.gov (United States)

    Tu, Weichao; Li, Wen; Morley, Steve; Albert, Jay

    2017-04-01

    In Summer 2015 the US NSF GEM (Geospace Environment Modeling) focus group named "Quantitative Assessment of Radiation Belt Modeling" started the "RB dropout" and "RB buildup" challenges, focused on quantitative modeling of the radiation belt buildups and dropouts. This is a community effort which includes selecting challenge events, gathering model inputs that are required to model the radiation belt dynamics during these events (e.g., various magnetospheric waves, plasmapause and density models, electron phase space density data), simulating the challenge events using different types of radiation belt models, and validating the model results by comparison to in situ observations of radiation belt electrons (from Van Allen Probes, THEMIS, GOES, LANL/GEO, etc). The goal is to quantitatively assess the relative importance of various acceleration, transport, and loss processes in the observed radiation belt dropouts and buildups. Since 2015, the community has selected four "challenge" events under four different categories: "storm-time enhancements", "non-storm enhancements", "storm-time dropouts", and "non-storm dropouts". Model inputs and data for each selected event have been coordinated and shared within the community to establish a common basis for simulations and testing. Modelers within and outside US with different types of radiation belt models (diffusion-type, diffusion-convection-type, test particle codes, etc.) have participated in our challenge and shared their simulation results and comparison with spacecraft measurements. Significant progress has been made in quantitative modeling of the radiation belt buildups and dropouts as well as accessing the modeling with new measures of model performance. In this presentation, I will review the activities from our "RB dropout" and "RB buildup" challenges and the progresses achieved in understanding radiation belt physics and improving model validation and verification.

  4. Metal-ion passivation and control of radiation field build-up

    International Nuclear Information System (INIS)

    Venkateswaran, G.

    2000-01-01

    Minimizing the activity transport due to corrosion products is an important goal for many current and future water cooled nuclear power plants (NPPs) to ensure lower occupational doses. Adding trace metals to the primary coolant is one of the preventive measures emerging as the way to control the deposition of radionuclides such as 60 Co on system surfaces. In this respect, the beneficial role of zinc has been experimentally studied in simulated coolants of boiling water reactor (BWR), pressurized light water reactor (PWR) and pressurized heavy water reactor (PHWR). Zinc addition is being practised in a number of operating BWRs and these reactors have experienced a lower radiation field buildup when compared with the fields which existed during pre-zinc operation. PWRs and PHWRs are yet to implement zinc addition. Alternatives to Zn addition, like Mg injection is studied. The control of [Ni] / [Fe] concentration in feed water to ≤ 0.2 has accomplished a great success in controlling the radiation fields in the initial operational stages of newly commissioned BWRs. The concentration Ni and Co species diminished in reactor water with such control. Noble metal passivation of stainless steel surfaces is a very recent development with respect to hydrogen water chemistry followed in BWRs. This seems to result in lowering the required hydrogen levels in feed water for controlling the dissolved oxygen in reactor water. This has a bearing on the fence 16 N activity levels experienced in BWR plants. (author)

  5. 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.

  6. 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

  7. Efforts to control radiation build-up in Ringhals

    Energy Technology Data Exchange (ETDEWEB)

    Egner, K.; Aronsson, P.O.; Erixon, O. [Vattenfall AB, Vaeroebacka (Sweden)

    1995-03-01

    It is well known that good control of the primary chemistry in a PWR is essential in order to minimize material problems and fuel damages. It has also been well established that the water chemistry has a great influence on accumulation of corrosion products on the fuel and the radiation build-up on primary system surfaces. Ringhals was one of the pioneers to increase operating pH in order to reduce radiation build-up and has now been operating for ten years with pH at 7.4 or (in later years) 7.2. Our experience is favourable and includes low radiation levels in the new (1989) steam generators of Ringhals 2. Ringhals 4 has operated almost its whole life at pH 7.2 or higher and it remains one of the cleanest PWRs of its vintage. In addition to strict adherence to a stable operating chemistry, Ringhals is now working on a program with the aim to find optimum shut-down and start-up chemistry to reduce activity levels in the primary systems. A particular goal is to use the shut-down and start-up chemistry at the 1994 outage in Ringhals 3 in order to reduce doserates in preparation for the planned steam generator replacement in 1995. The paper summarizes the experience to date of the established operating chemistry, on-going tests with modified shut-down and start-up chemistry and other measures to limit or reduce the activity build-up.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-03-01

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

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

    International Nuclear Information System (INIS)

    Aldrich, L.R.

    1995-01-01

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

  10. 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)

  11. Radioactivity build-up and decontamination, 3

    International Nuclear Information System (INIS)

    Kanbe, Hiroki; Miyashiro, Hajime; Tomizawa, Toshi

    1983-01-01

    To establish radiation control technology, radioactivity build-up on 17 types of steels were examined under a typical BWR condition up to 400 hours, using Radioactivity Build-up Test Loop (RBTL). Ionic 58 Co was used as tracer in the tests. The results were as follows: (1) Radioactivity build-up levels on steels (SUS 304, SUS XM15J1, SUS 430) and special steels (15Cr-9Ni, 20Cr-9Ni) were ten times larger than those on carbon steels (NC 41, STS 38, STPT 42), anti-weathering steel (SMA 41), low alloy steel (STPA 24) and special steels (5Cr-9Ni, 10Cr-9Ni). (2) A stepwise increase in radioactivity build-up was observed when chromium content of steel was raised from 10 to 15 %. (3) Radioactivity build-up was proportional to chromium concentration of deposits, which suggested that 58 Co radioactivity build-up was responsible to chromium in deposits. (4) Corrosion rate of steels decreased with increasing chromium content of steels. It would be concluded that the most appropriate amount of chromium in a steel is about 10 %, considering both radioactivity build- up and corrosion. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-03-01

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

  13. Exposure Buildup Factors for Heavy Metal Oxide Glass: A Radiation Shield

    DEFF Research Database (Denmark)

    Manonara, S. R.; Hanagodimath, S. M.; Gerward, Leif

    2011-01-01

    Gamma ray exposure buildup factors for three Heavy Metal Oxide (HMO) glass systems, viz. PbO-Bi2O3-B2O3, PbO-B2O3, and Bi2O3-B2O3 glasses are presented. The computations were done by interpolation method using the Geometric Progression fitting formula and ANSI/ANS-6.4.3 library for the energy range...

  14. Development of the radiation models of a BWR type reactor and it facility in the SUN-RAH

    International Nuclear Information System (INIS)

    Barron A, I.

    2005-01-01

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

  15. 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

  16. A study of build-up effects in high-energy radiation fields using a TEPC

    International Nuclear Information System (INIS)

    Hoefert, M.; Stevenson, G.R.; Aroua, A.; Sannikov, A.V.

    1995-01-01

    A dose of 2 mSv close to the body surface of a pregnant woman is considered by ICRP to assure a dose limit of 1 mSv to the foetus. Such an assumption depends on the energy spectrum and composition of the external radiation field and it was tested in radiation fields containing high-energy particles similar to those found around high-energy particle accelerators and in air-craft. Measurements of dose and dose equivalent were performed as a function of wall thickness using a tissue-equivalent proportional counter (TEPC) in radiation fields at the CERN-EU Reference Radiation Facility. Results are presented both with respect to integral quantities and event size spectra. The decrease in dose and dose equivalent at a depth equivalent to that of the foetus was typically 10% in a high-energy stray radiation field and in the case of PuBe source neutrons amounted to only 30%. It is concluded that it would be prudent under such exposure conditions to limit the dose of a pregnant woman to 1 mSv in order to assure that the dose to the foetus remains below the same limit. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

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

  18. 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.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-07-01

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

  20. Klein-Nishina electronic cross-section, Compton cross sections, and buildup factor of wax for radiation shielding and protection.

    Science.gov (United States)

    Alenezi, Manar; Stinson, Kayla R; Maqbool, Muhammad; Bolus, Norman

    2018-01-05

    Klein-Nishina scattering cross-sections, Compton scattering, mass attenuation and energy transfer cross-sections, linear attenuation coefficient and buildup factor of paraffin wax are calculated using 0.662 MeV, 0.835 MeV, 1.17 MeV and 1.33 MeV γ-rays. The mentioned γ-rays were obtained from Cs137, Mn54 and Co60 radioisotopes. Gamma rays obtained from these radioisotopes were passed through circular shaped wax slices and allowed to fall on a NaI detector. The thickness of wax slices were 0.33 cm to 2.9 cm with 6 cm diameter. Lead collimator of 1 cm diameter hole in the middle was used to obtain a collimated beam for narrow beam geometry. Broad beam geometry was used by removing collimator, to investigate buildup factor. Results show that Klein-Nishina electronic cross-section, Compton mass attenuation coefficient and Compton energy transfer coefficient all decrease with increasing photon energy. Slope of the thickness vs counts graph for narrow beam geometry gives linear attenuation coefficient μ = 0.0532 cm-1 for 1.17 MeV beam and μ = 0.0419 cm-1 for 1.33 MeV γ-rays. Variations in buildup factors are observed with increasing thickness of wax for 1.17 MeV and 1.33 MeV beams. © 2018 IOP Publishing Ltd.

  1. Applied measures and further subjects for radiation exposure reduction at BWR in Japan

    International Nuclear Information System (INIS)

    Otoha, K.

    1994-01-01

    The dose equivalent for workers at the Tokyo Electric Power Company's nuclear power plant operation, which started some 20 years ago, has lowered to about one fifth of the highest level record during the period. Apparently this has been achieved by the effects of the measures actively carried out to reduce the radiation exposure. This Report summarizes these radiation exposure reduction effects, followed by discussions on the current state and future developments of the company's activities in this field with careful consideration given to the in individual dose equivalent for plant workers. 1 tab., 4 figs

  2. Definition of loss-of-coolant accident radiation source. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    1978-02-01

    Meaningful qualification testing of nuclear reactor components requires a knowledge of the radiation fields expected in a loss-of-coolant accident (LOCA). The overall objective of this program is to define the LOCA source terms and compare these with the output of various simulators employed for radiation qualification testing. The basis for comparison will be the energy deposition in a model reactor component. The results of the calculations are presented and some interpretation of the results given. The energy release rates and spectra were validated by comparison with other calculations using different codes since experimental data appropriate to these calculations do not exist.

  3. BWR operation for life extension

    International Nuclear Information System (INIS)

    Stancavage, P.

    1987-01-01

    Most nuclear power plant life extension studies conducted to date focus on the technology, licensing, and economics of extended service life. The most significant factor in plant longevity, however, is proper operation and maintenance. This paper highlights the benefits of boiling water reactor (BWR) operation for life extension and discusses specific recommendations that will enhance the prospects for safe, reliable, and economic power production in the long term. The benefits of BWR operation for life extension include a lower cost of electric energy production, increased capacity and availability factors, a lower forced outage rate, and reduced occupational exposure to radiation. Operating experience and advanced in technology have provided a wealth of knowledge that can be used to develop specific recommendations for adding years to the expected life of BWR plants. This paper discusses key factors in operation for life extension

  4. Condensate treatment in BWR circuits by filter demineralizer units using powdered ion exchange resin at medium and high temperature

    International Nuclear Information System (INIS)

    De Martino, R.

    1983-01-01

    Considering the radiation build-up in some BWR reactors, we make a correlation between this phenomenon and the condensate purification system applied and the point of its utilization into the circuits. The application temperature of such a plant seems to have a very important role on the equilibria of metals contained in the reactor water and on the oxide composition. The efficiency of the condensate polishing system and the corrosion control are the most interesting objectives to achieve and to maintain, to control and regulate the physical and chemical process in the feedwater and in the reactor water. Up to date the technology owns major knowledge and a consistent know-how on using chemical products in order to increase the condensate polishing system efficiency. It is also considered a typical parallel case of a conventional power station and a secondary system of BWR units. (author)

  5. 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)

  6. Correlation between Co-60 and X-ray exposures on radiation-induced charge buildup in silicon-on-insulator buried oxides

    International Nuclear Information System (INIS)

    Schwank, James R.; Shaneyfelt, Marty R.; Loemker, Rhonda Ann; Draper, Bruce L.; Dodd, Paul E.; Witczak, StevenN C.; Riewe, Leonard Charles; Ferlet-Cavrois, V.; Paillet, P.; Leray, J.-L.; Fleetwood, D.M.

    2000-01-01

    Large differences in charge buildup in SOI buried oxides can result between x-ray and Co-60 irradiations. The effects of bias configuration and substrate type on charge buildup and hardness assurance issues are explored

  7. 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

  8. Determination of effective atomic numbers, effective electrons numbers, total atomic cross-sections and buildup factor of some compounds for different radiation sources

    Science.gov (United States)

    Levet, A.; Özdemir, Y.

    2017-01-01

    The photon interaction parameters such as mass attenuation coefficient, effective atomic number, effective electron density, buildup factor have been measured for Fe(NO3)3, V4O2, NaCO3·H2O, C6H5FeO7·H2O and CuCI compounds using 137Ba, 157Gd and 241Am γ-rays sources in stable geometry. The mass attenuation coefficients have been determined experimentally via Energy Dispersive X-ray Fluorescence Spectroscopy (EDXRF) system and theoretically by using WinXCom computer program. Then, effective atomic numbers, Zeff, and electron densities, Neff, have been calculated by using the mass attenuation coefficients. The obtained values of effective atomic numbers have been compared with the ones calculated according to a different approach proposed by Hine and the calculated ones from theory. Also, photon buildup factors were obtained by changing collimator diameters in the different photon energies. We observed that the buildup factor increased as the collimator diameter increased for all sources used.

  9. 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

  10. Estimate of radiation-induced steel embrittlement in the BWR core shroud and vessel wall from reactor-grade MOX/UOX fuel for the nuclear power plant at Laguna Verde, Veracruz, Mexico

    Science.gov (United States)

    Vickers, Lisa Rene

    The government of Mexico has expressed interest to utilize the Laguna Verde boiling water reactor (BWR) nuclear power plant for the disposition of reprocessed spent uranium oxide (UOX) fuel in the form of reactor-grade mixed-oxide (MOX) fuel. MOX fuel would replace spent UOX fuel as a fraction in the core from 18--30% depending on the fuel loading cycle. MOX fuel is expected to increase the neutron fluence, flux, fuel centerline temperature, reactor core pressure, and yield higher energy neutrons. There is concern that a core with a fraction of MOX fuel (i.e., increased 239Pu wt%) would increase the radiation-induced steel embrittlement within the core shroud and vessel wall as compared to only conventional, enriched UOX fuel in the core. The evaluation of radiation-induced steel embrittlement within the core shroud and vessel wall is a concern because of the potentially adverse affect to plant and public safety, environment, and operating life of the reactor. This dissertation provides computational results of the neutron fluence, flux, energy spectrum, and radiation damage displacements per atom per second (dpa-s-1) in steel within the core shroud and vessel wall of the Laguna Verde Unit 1 BWR. The results were computed using the nuclear data processing code NJOY99 and the continuous energy Monte Carlo Neutral Particle transport code MCNP4B. The MCNP4B model of the reactor core was for maximum core loading fractions of ⅓ MOX and ⅔ UOX reactor-grade fuel in an equilibrium core. The primary conclusion of this dissertation was that the addition of the maximum fraction of ⅓ MOX fuel to the LV1 BWR core did significantly accelerate the radiation-induced steel embrittlement such that without mitigation of steel embrittlement by periodic thermal annealing or reduction in operating parameters such as, neutron fluence, core temperature and pressure, it posed a potentially adverse affect to the plant and public safety, environment, and operating life of the reactor.

  11. 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

  12. 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

  13. Recent developments in BWR water chemistry

    International Nuclear Information System (INIS)

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

  14. BWR Stability Issues in Japan

    Directory of Open Access Journals (Sweden)

    Hideaki Ikeda

    2008-01-01

    Full Text Available The present paper reviews activates relevant to the boiling water reactor (BWR stability phenomenon, which has a coupled neutronic and thermal-hydraulic nature, from the viewpoint of model and code developments and their applications to the BWR stability solution methodology in Japan.

  15. A boiling-water reactor concept for low radiation exposure based on operating experience

    International Nuclear Information System (INIS)

    Koine, Y.; Uchida, S.; Izumiya, M.; Miki, M.

    1983-01-01

    A review of boiling-water reactor (BWR) operating experience indicates the significant role of water chemistry in determining the radiation dose rate contributing to occupational exposure. The major contributor among the radioactive species involved is identified as 60 Co, produced by neutron activation of 59 Co originating from structural materials. Iron crud, a fine solid form of corrosion product in the reactor water, is also shown to enhance the radiation dose rate. A theoretical study, supported by the operating experience and an extensive confirmatory test, led to the computerized analytical model called DR CRUD which is capable of predicting long-term radiation dose buildup. It accounts for the mechanism of radiation buildup through corrosion products such as irons, cobalts and other radioactive elements; their generation, transport, activation, interaction and deposition in the reactor coolant system are simulated. A scoping analysis, using this model as a tool, establishes the base line of the BWR concept for low occupational exposure. The base line consists of a set of target values for an annual exposure of 200 man.rem in an 1100 MW(e) BWR unit. They are the parameters that will be built into the design such as iron and cobalt inputs to the reactor water, and the capability of the reactor and the condensate purification system. Applicable means of technology are identified to meet the targets, ranging from improved water chemistry to the purification technique, optimized material selection and the recommended operational procedure. Extensive test programmes provide specifications of these means for use in BWRs. Combinations of their application are reviewed to define the concept of reduced exposure. Analytical study verifies the effectiveness of the proposed BWR concept in achieving a low radiation dose rate; occupational exposure is reduced to 200 man.rem/a. (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. The distribution of the activity in the primary system of a BWR

    International Nuclear Information System (INIS)

    Westerberg, S.; Hermansson, H.-P.

    1980-12-01

    A model for the distribution of activity is presented. The model is based upon an earlier mathematical model which has been modified and applied to the calculation of activity of the primary water system and to the buildup of activity in ion exchange materials. Calculations have been performed for Co-60. A certain agreement of the calculations and the data from Swedish BWR plants has been obtained. (G.B.)

  18. BWR type nuclear plant

    International Nuclear Information System (INIS)

    Fujita, Yoshio; Okano, Kimifumi; Sasaki, Hiroshi; Okura, Minoru

    1998-01-01

    The present invention provides a BWR plant capable of reducing the size of the reactor building while maintaining reliability of a pool water cooling and cleaning facility even when two fuel storage pools are disposed in the reactor building. Namely, in the reactor building, two fuel storage pools, a temporality storing pool for temporary storing incore structures and a suppression pool are disposed. A primary cleaning facility for cooling and cleaning pool water for each of fuel storage pools comprises a serge tank, a pump, a heat exchanger and a filtration desalting device. A secondary cleaning facility for cleaning pool water in the suppression pool comprises a pump and a filtration desalting device. The first cleaning facility can be switched to be used for the secondary cleaning facility. Specifically, upstream and downstream of the pump of the primary cleaning facility and those of the pump of the secondary cleaning facility are connected by communication pipelines. (I.S.)

  19. Study of the corrosion products in the primary system of PWR plants as the source of radiation fields build-up

    International Nuclear Information System (INIS)

    Brabant, R. van; Regge, P. de.

    1982-01-01

    In the first part the behaviour of the corrosion products in the primary system of PWR plants is depicted on the basis of a literature review of the field. Water chemistry, corrosion processes and activation of corrosion products are the main topics. In the second part the results of the characterization of corrosion particles in the primary coolant circuit of the Doel 1 and 2 reactors are described, during steady state operation and transient phases. In the third part the possibilities for radiation control at nuclear power plants are outlined. The filtration possibilities for the reactor coolant are explored in detail. (author)

  20. 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.

  1. 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.)

  2. BWR crack control

    International Nuclear Information System (INIS)

    Immel, R.

    1981-01-01

    Cooperative research by the Electric Power Research Institute (EPRI) and the Boiling Water Reactor Owners Group (BWROG) may be finding a solution to the problem of stress corrosion cracking in reactor piping, which had caused 254 cracks in BWRs around the world by 1981. Welded joints of type-304 stainless steel in BWR recirculation systems are particularly vulnerable to microscopic stress corrosion cracking, which differs from fatigue cracking because it occurs when electrochemical processes break down the protective film in the zones affected by welding heat. The cracking is not hazardous to the public because it takes place within the containment building and causes leaks that are quickly detected. About one percent of welds crack, causing downtime and personnel exposure. The EPRI-BWROG project used a model of stress, water condition, and steel sensitization to find out the causes, develop remedies, and get the remedies into the field. Among the remedies are the adaptive learning network, an automated, microprocessor-controlled ultrasonic testing system that is scheduled for field testing and utility application in 1982. 2 figures

  3. Energy absorption and exposure build-up factors in teeth

    International Nuclear Information System (INIS)

    Manjunatha, H.C.; Rudraswamy, B.

    2010-01-01

    Full text: Gamma and X-radiation are widely used in medical imaging and radiation therapy. The user of radioisotopes must have knowledge about how radiation interacts with matter, especially with the human body, because when photons enter the medium/body, they degrade their energy and build up in the medium, giving rise to secondary radiation which can be estimated by a factor which is called the 'build-up factor'. It is essential to study the exposure build up factor in radiation dosimetry. G.P. fitting method has been used to compute energy absorption and exposure build-up factor of teeth (enamel outer surface (EOS), enamel middle (EM), enamel dentin junction towards enamel (EDJE), enamel dentin junction towards dentin (EDJD), dentin middle (DM) and dentin inner surface (DIS)) for wide energy range (0.015 MeV-15 MeV) up to the penetration depth of 40 mean free path. The dependence of energy absorption and exposure build up factor on incident photon energy, Penetration depth and effective atomic number has also been assessed. The relative dose distribution at a distance r from the point source is also estimated. The computed exposure and absorption build-up factors are useful to estimate the gamma and Bremsstrahlung radiation dose distribution teeth which is useful in clinical dosimetry

  4. Composition and Distribution of Tramp Uranium Contamination on BWR and PWR Fuel Rods

    International Nuclear Information System (INIS)

    Schienbein, Marcel; Zeh, Peter; Hurtado, Antonio; Rosskamp, Matthias; Mailand, Irene; Bolz, Michael

    2012-09-01

    In a joint research project of VGB and AREVA NP GmbH the behaviour of alpha nuclides in nuclear power plants with light water reactors has been investigated. Understanding the source and the behaviour of alpha nuclides is of big importance for planning radiation protection measures for outages and upcoming dismantling projects. Previous publications have shown the correlation between plant specific alpha contamination of the core and the so called 'tramp fuel' or 'tramp uranium' level which is linked to the defect history of fuel assemblies and accordingly the amount of previously washed out fuel from defective fuel rods. The methodology of tramp fuel estimation is based on fission product concentrations in reactor coolant but also needs a good knowledge of tramp fuel composition and in-core distribution on the outer surface of fuel rods itself. Sampling campaigns of CRUD deposits of irradiated fuel assemblies in different NPPs were performed. CRUD analyses including nuclide specific alpha analysis have shown systematic differences between BWR and PWR plants. Those data combined with literature results of fuel pellet investigations led to model improvements showing that a main part of fission products is caused by fission of Pu-239 an activation product of U-238. CRUD investigations also gave a better picture of the in-core composition and distribution of the tramp uranium contamination. It was shown that the tramp uranium distribution in PWR plants is time dependent. Even new fuel assemblies will be notably contaminated after only one cycle of operation. For PWR applies the following logic: the higher the local power the higher the contamination. With increasing burnup the local rod power usually decreases leading to decreasing tramp uranium contamination on the fuel rod surface. This is not applicable for tramp uranium contamination in BWR. CRUD contamination (including the tramp fuel deposits) is much more fixed and is constantly increasing

  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. Range of the radiation monitor for the rigid vent of primary containment during normal and emergency operation for a BWR-5 in Laguna Verde

    International Nuclear Information System (INIS)

    Tijerina S, F.; Pozos S, A. M.; Cabrera U, S.; Mata A, J. A.; Sandoval V, S.; Ovando C, R.; Vargas A, A.; Gallardo R, I.; Cruz G, M.; Amador C, C.

    2014-10-01

    The earthquake followed by a tsunami, happened in March, 2011 in the coasts of oriental Japan, caused damages in the nuclear power plants 1 at 4 of Fukushima Daiichi leading to damage of the fuel in three of the reactors and to the radiation liberation to the exterior. As consequence of those events, the regulations requires that the power plants with Primary Containment type Mark I and II evaluate to have a system of rigid vent with a monitoring equipment of radiation effluents. The present work covers the rigid vent of diameter 12 of the Primary Containment, type Mark-II, of nuclear power plant of Laguna Verde in conditions of severe accident and normal operation, low regime of Extended Power Up rate (EPU - 2317 MWt), using the codes MAAP3B, MICROSHILED 5.05 and the Bardach Black Boxes methodology. As a result the measurement range of the radiation monitor that is required for monitoring the gassy liberation to the atmosphere was determined. The conclusion is that the superior limit of the range of the radiation meter during a Severe Accident is of 8.55 E + 05 R/h (8.55 E + 08 m R/h) and the superior limit in normal operation of 1.412 E-11 at 2.540 E-7 R/h (1.412 E-14 at 2.540 E-10 m R/h). (Author)

  7. 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)

  8. 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

  9. 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.)

  10. 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.).

  11. Improved SVR Model for Multi-Layer Buildup Factor Calculation

    International Nuclear Information System (INIS)

    Trontl, K.; Pevec, D.; Smuc, T.

    2006-01-01

    The accuracy of point kernel method applied in gamma ray dose rate calculations in shielding design and radiation safety analysis is limited by the accuracy of buildup factors used in calculations. Although buildup factors for single-layer shields are well defined and understood, buildup factors for stratified shields represent a complex physical problem that is hard to express in mathematical terms. The traditional approach for expressing buildup factors of multi-layer shields is through semi-empirical formulas obtained by fitting the results of transport theory or Monte Carlo calculations. Such an approach requires an ad-hoc definition of the fitting function and often results with numerous and usually inadequately explained and defined correction factors added to the final empirical formula. Even more, finally obtained formulas are generally limited to a small number of predefined combinations of materials within relatively small range of gamma ray energies and shield thicknesses. Recently, a new approach has been suggested by the authors involving one of machine learning techniques called Support Vector Machines, i.e., Support Vector Regression (SVR). Preliminary investigations performed for double-layer shields revealed great potential of the method, but also pointed out some drawbacks of the developed model, mostly related to the selection of one of the parameters describing the problem (material atomic number), and the method in which the model was designed to evolve during the learning process. It is the aim of this paper to introduce a new parameter (single material buildup factor) that is to replace the existing material atomic number as an input parameter. The comparison of two models generated by different input parameters has been performed. The second goal is to improve the evolution process of learning, i.e., the experimental computational procedure that provides a framework for automated construction of complex regression models of predefined

  12. 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).

  13. From detached to attached buildup complexes

    DEFF Research Database (Denmark)

    Rafaelsen, B.; Elvebakk, G.; Andreassen, K.

    2008-01-01

    , up to 2.5 km wide and 300 m thick, and interpreted to consist of vertically stacked complexes of sub-seismic scale carbonate buildups. Evaporites were deposited and later subject to karstification, possibly during a period of sub-aerial exposure, before a transgression and the subsequent carbonate......Carbonate buildups were abundant during the Palaeozoic. Three-dimensional seismic data from the Finnmark Platform, Barents Sea, has been used to reconstruct the evolution of laterally extensive carbonate buildup complexes in space and time. The results suggest that the location of Upper Palaeozoic...

  14. BWR type reactor system

    International Nuclear Information System (INIS)

    Morooka, Shin-ichi.

    1980-01-01

    Purpose: To reduce the internal structure in a reactor by rapidly and efficiently transferring heat generated in a reactor core out of the reactor and eliminating the danger of radiation exposure. Constitution: Steam generated in a pressure vessel is introduced into heat pipe group by inserting the heat pipe group into the steam dome of the pressure vessel. The introduced steam is condensed in the heat pipes to transfer the heat of the steam to the heat pipe group. The transferred heat is transmitted to a heat exchanger provided out of a containment vessel to generate steam to operate a turbine. Thus, it is not necessary to introduce the steam including radioactive substance externally and can remove only the heat so as to carry out the desired purpose. (Kamimura, M.)

  15. Momentum management strategy during Space Station buildup

    Science.gov (United States)

    Bishop, Lynda; Malchow, Harvey; Hattis, Philip

    1988-01-01

    The use of momentum storage devices to control effectors for Space Station attitude control throughout the buildup sequence is discussed. Particular attention is given to the problem of providing satisfactory management of momentum storage effectors throughout buildup while experiencing variable torque loading. Continuous and discrete control strategies are compared and the effects of alternative control moment gyro strategies on peak momentum storage requirements and on commanded maneuver characteristics are described.

  16. 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

  17. Automatic refueling platform and CRD remote handling device for BWR plant

    International Nuclear Information System (INIS)

    Kato, Hiroaki; Takagi, Kaoru

    1978-01-01

    In BWR plants, machines for replacing fuel assemblies and control rod drives are usually operated directly by personnel. An automatic refueling platform and a CRD remote handling device aiming at radiation exposure reduction and handling perfectness are described, which are already used in BWR plants. Automation of the former is achieved in transporting fuel assemblies between a reactor pressure vessel and a fuel storage pool, shuffling fuel assemblies in a reactor core and moving fuel assemblies in a fuel storage pool. In the latter, replacement of CRDs is nearly all performed remotely. (Mori, K.)

  18. 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)

  19. 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)

  20. Electron contamination and build-up doses in conformal radiotherapy fields.

    Science.gov (United States)

    Hounsell, A R; Wilkinson, J M

    1999-01-01

    The dose in the build-up region depends upon the primary photon beam, backscattered radiation from the patient and contamination radiation from outside the patient. In this paper, a model based on measured data is proposed which allows the build-up dose for arbitrarily shaped treatment fields to be determined. The dose in the build-up region is assumed to comprise a primary photon component and a contamination component that is a function of the field size and shape. This contamination component, for modelling purposes, is subdivided into contributions that correspond to elements of 1 cm by 1 cm cross-sectional area at the plane of the isocentre. The magnitude of these components has been obtained by fitting measured data to an exponential function. The exponent was found to vary linearly with depth for energies between 4 MV and 20 MV. The coefficient decreased linearly with depth at 4, 6 and 8 MV, but exhibited a broad build-up region at 20 MV. The primary component, in the build-up region, could be approximated by a 100 - (100 - PSD) e(-mu d) function, where PSD is the primary surface dose. The values obtained during the fitting procedure were used to calculate dose in the build-up region for arbitrarily shaped fields. Good agreement was found in each case.

  1. BWR Source Term Generation and Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    J.C. Ryman

    2003-07-31

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

  2. BWR Assembly Optimization for Minor Actinide Recycling

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-03-22

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

  3. 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)

  4. The Calculation Of Titanium Buildup Factor Based On Monte Carlo Method

    International Nuclear Information System (INIS)

    Has, Hengky Istianto; Achmad, Balza; Harto, Andang Widi

    2001-01-01

    The objective of radioactive-waste container is to reduce radiation emission to the environment. For that purpose, we need material with ability to shield that radiation and last for 10.000 years. Titanium is one of the materials that can be used to make containers. Unfortunately, its buildup factor, which is an importance factor in setting up radiation shielding, has not been calculated. Therefore, the calculations of titanium buildup factor as a function of other parameters is needed. Buildup factor can be determined either experimentally or by simulation. The purpose of this study is to determine titanium buildup factor using simulation program based on Monte Carlo method. Monte Carlo is a stochastic method, therefore is proper to calculate nuclear radiation which naturally has random characteristic. Simulation program also able to give result while experiments can not be performed, because of their limitations.The result of the simulation is, that by increasing titanium thickness the buildup factor number and dosage increase. In contrary If photon energy is higher, then buildup factor number and dosage are lower. The photon energy used in the simulation was ranged from 0.2 MeV to 2.0 MeV with 0.2 MeV step size, while the thickness was ranged from 0.2 cm to 3.0 cm with step size of 0.2 cm. The highest buildup factor number is β = 1.4540 ± 0.047229 at 0.2 MeV photon energy with titanium thickness of 3.0 cm. The lowest is β = 1.0123 ± 0.000650 at 2.0 MeV photon energy with 0.2 cm thickness of titanium. For the dosage buildup factor, the highest dose is β D = 1.3991 ± 0.013999 at 0.2 MeV of the photon energy with a titanium thickness of 3.0 cm and the lowest is β D = 1.0042 ± 0.000597 at 2.0 MeV with titanium thickness of 0.2 cm. For the photon energy and the thickness of titanium used in simulation, buildup factor and dosage buildup factor as a function of photon energy and titanium thickness can be formulated as follow β = 1.1264 e - 0.0855 E e 0 .0584 T

  5. 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)

  6. Am-241 buildup in nematode organisms

    International Nuclear Information System (INIS)

    Martyushov, V.Z.; Tarasov, O.V.

    1990-01-01

    The process of Am-241 intake into earthworm organisms from chernozem leached in their presence in soil contaminated with this radionuclide is studied. The data on Am-241 buildup values during long-time radionuclide intake into earthworm organisms from soil are given. It s shown that Am-241 buildup in earthworm organisms do not exceed its concentration in soil for the whole observation period (as Am-241 presents in soil in state unavailable for animals). Intensive extraction of the radionuclide from the organisms is observed when earthworm contacts with soil are stopped

  7. Data library of gamma-ray buildup factors for point isotropic source

    International Nuclear Information System (INIS)

    Sakamoto, Yukio; Tanaka, Shun-ichi; Harima, Yoshiko.

    1988-01-01

    Gamma-ray buildup factors for a point isotropic source have been calculated as a function of atomic number of heavy elements and source energies over an energy range from 0.015 MeV to 15 MeV, for penetration depths up to 40 mfp, bu the PALLAS-PL,SP-Br code. These data include the contribution of bremsstrahlung, annihilation radiation and fluorescence X-ray. The calculated absorbed-dose, exposure and dose-equivalent buildup factors are tabulated for molybdenum, tin, tungsten, lead and uranium, which are practical interest shield materials, lanthanum and gadolinium which are important materials for obtaining buildup factors by interpolation with the atomic number. In the case of high atomic number materials, inclusion of brems-strahlung source has great influence on the buildup factors for high source energies and that of fluorescence X-ray gives spectracular effects on those for low energies close to the K edge of attenuation cross section. Furthermore, the geometrical-progression (G-P) parameters have been determined for these buildup factors in order to obtain the values of buildup factors at arbitrary distances and energies. (author)

  8. 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)

  9. 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

  10. Development and recent trend of disign of BWR nuclear power plants

    International Nuclear Information System (INIS)

    Kani, Jiro

    1977-01-01

    Many improvements have been carried out in BWR nuclear power plants from BWR-1, represented by Dresden No. 1 plant, to the present BWR-6 as the capacity has increased. In Japan, the plants up to BWR-5 have been constructed. In addition, further fine design improvements are being performed in the complete domestic manufacturing of BRWs based on the operational experiences to date. A variety of investigations on the standardization of nuclear power facilities have been progressing under the leadership of Japanese Ministry of International Trade and Industry since 1975. In this standardization, it is intended to forward the plant design taking eight concrete items into consideration, mainly aiming at carrying cut unerringly the maintenance and inspection, reduction of exposure of employees to radiation, and improvements of the rate of operation of plants and equipment reliability. The containment vessel has been developed in three forms, from Mark 1 through 3, adopting the pressure control system consistently since BWR-2. Mark 1 and 2 were constructed in Japan. However, these designs sacrificed the workability and increased radiation exposure during maintenance as a result of placing emphasis on the safety facilities, therefore Toshiba Electric has investigated the advanced Mark 1 type. Its features are the design for improving the work efficiency in a containment vessel, reducing the radiation exposure of workers, shortening plant construction period, and considering the aseismatic capability. In addition, the following themes are being planned as future standardization: (1) electrically driven control rod driving system, (2) improved design of reactor core, and (3) internal pump system as compared with external re-circulation. (Wakatsuki, Y.)

  11. 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)

  12. Monte Carlo simulation of photon buildup factors for shielding materials in radiotherapy x-ray facilities

    Energy Technology Data Exchange (ETDEWEB)

    Karim Karoui, Mohamed [Faculte des Sciences de Monastir, Avenue de l' environnement 5019 Monastir -Tunisia (Tunisia); Kharrati, Hedi [Ecole Superieure des Sciences et Techniques de la Sante de Monastir, Avenue Avicenne 5000 Monastir (Tunisia)

    2013-07-15

    Purpose: This paper presents the results of a series of calculations to determine buildup factors for ordinary concrete, baryte concrete, lead, steel, and iron in broad beam geometry for photons energies from 0.125 to 25.125 MeV at 0.250 MeV intervals.Methods: Monte Carlo N-particle radiation transport computer code has been used to determine the buildup factors for the studied shielding materials.Results: The computation of the primary broad beams using buildup factors data was done for nine published megavoltage photon beam spectra ranging from 4 to 25 MV in nominal energies, representing linacs made by the three major manufacturers. The first tenth value layer and the equilibrium tenth value layer are calculated from the broad beam transmission for these nine primary megavoltage photon beam spectra.Conclusions: The results, compared with published data, show the ability of these buildup factor data to predict shielding transmission curves for the primary radiation beam. Therefore, the buildup factor data can be combined with primary, scatter, and leakage x-ray spectra to perform computation of broad beam transmission for barriers in radiotherapy shielding x-ray facilities.

  13. 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.)

  14. Natural convection type BWR reactor

    International Nuclear Information System (INIS)

    Tobimatsu, Toshimi.

    1990-01-01

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

  15. 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.

  16. A BWR fuel channel tracking system

    International Nuclear Information System (INIS)

    Reynolds, R.S.

    1987-01-01

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

  17. 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.)

  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. 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

  20. 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

  1. 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)

  2. A study of the energy absorption and exposure buildup factors of some anti-inflammatory drugs

    International Nuclear Information System (INIS)

    Ekinci, Neslihan; Kavaz, Esra; Özdemir, Yüksel

    2014-01-01

    Human radiation exposure is increasing due to radiation development in science and technology. The development of radioprotective agents is important for protecting patients from the side effects of radiotherapy and for protecting the public from unwanted irradiation. Radioprotective agents are used to reduce the damage caused by radiation in healthy tissues. There are several classes of radioprotective compounds that are under investigation. Analgesics and anti-inflammatory compounds are being considered for treating or preventing the effects of damage due to radiation exposure, or for increasing the chance of survival after exposure to a high dose of radiation. In this study, we investigated the radioprotective effects of some analgesic and anti-inflammatory compounds by evaluating buildup factors. The gamma ray energy absorption (EABF) and exposure buildup factors (EBF) were calculated to select compounds in a 0.015–15 MeV energy region up to a penetration depth of 40 mfp (mean free path). Variations of EABF and EBF with incident photon energy and penetration depth elements were also investigated. Significant variations in both EABF and EBF values were observed for several compounds at the moderate energy region. At energies below 0.15 MeV, EABF and EBF values increased with decreasing equivalent atomic number (Z eq ) of the samples. In addition, EABF and EBF were the largest for ibuprofen, aspirin, paracetamol, naproxen and ketoprofen at 0.05 and 0.06 MeV, respectively, and the EABF value was 0.1 MeV for aceclofenac. From these results, we concluded that the buildup of photons is less for aceclofenac compared to other materials. - Highlights: • Buildup factors of anti-inflammatory drugs have been calculated by a G-P fitting method. • Z eff of diclofenac was observed higher than other compounds. • It was found that buildup of photons is less for aceclofenac and diclofenac. • It would be appealing to use aceclofenac and diclofenac as radioprotective

  3. Innovative nuclear technologies based on radiation induced surface activation. (5) Development of high performance BWR by the radiation induced surface activation visualization study on the boiling enhancement with irradiation

    International Nuclear Information System (INIS)

    Imai, Yasuyuki; Okamoto, Koji; Madarame, Haruki; Takamasa, Tomoji

    2004-01-01

    Improvement of CHF requires that the cooling liquid can contact the heating surface, or a high-wettability, highly hydrophilic heating surface, even if a vapor bubble layer is generated on the surface. In our previous study, we confirmed that the surface wettability changed significantly or that highly hydrophilic conditions were achieved, after irradiation of 60 Co gamma ray, by the Radiation Induced Surface Activation (RISA) phenomenon. To delineate the effect of RISA on boiling phenomena, CHF of metal oxides irradiated by gamma rays were investigated. The heating test section made of titanium was 0.5 mm in diameter. Oxidation of the surface was carried out by plasma jetting. The test section was irradiated by 60 Co gamma ray with predetermined radiation intensity and period. A test piece had been hold horizontally on the electrode after 5400 kGy irradiation. Then, the whole CHF test apparatus with test piece was set on the table in the gamma ray irradiation room. The test piece was irradiated in the water at least 30 minutes. A CHF experiment in the pool boiling condition was carried out under atmospheric pressure under irradiation. The results of on-site experiment were compared with that of off-site one. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-07-01

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

  5. 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

  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 consolidation system final design report

    International Nuclear Information System (INIS)

    Garner, G.L.; Kelly, M.J.; Larsen, W.R.

    1993-05-01

    Because of delays in the opening of a permanent geologic repository to accept spent fuel from nuclear reactor plants, several utilities are seeking additional off-site storage to avert premature shutdown. Fuel rod consolidation is a proven, viable option for pressurized water reactor (PWR) plants, but until now, no consolidation system addressed boiling water reactor (BWR) spent-fuel assemblies.The purpose of this project, jointly funded by the Empire State Electric Energy Research Company (ESEERCO) and the Electric Power Research Institute (EPRI), is to develop a system for consolidating BWR spent fuel assemblies. This design will provide more efficient storage in reactor spent-fuel pools. The design goal is a 2:1 consolidation of the fuel rods and a minimum 10:1 compaction of the non-fuel bearing components. in addition, the consolidation system must be operationally compatible with BWR reactor plants and be economically viable with other forms of supplementary on-site storage. The work began in Lynchburg, Virginia on May 6, 1991 and concluded on September 30, 1992 with the delivery of the final report. The design achieves all of the project goals. Furthermore, consolidation of BWR spent-fuel assemblies is feasible, compatible with reactor plant operations and potentially the lowest cost option for a utility seeking to add oil-site storage capacity

  8. 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)

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

    International Nuclear Information System (INIS)

    Tarvainen, M.; Paakkunainen, M.; Tiitta, A.; Sarparanta, K.

    1994-04-01

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

  10. Experimental investigation of control absorber blade effects in a modern 10x10 BWR assembly

    International Nuclear Information System (INIS)

    Jatuff, F.; Grimm, P.; Murphy, M.; Luethi, A.; Seiler, R.; Joneja, O.; Meister, A.; Geemert, R. van; Brogli, R.; Chawla, R.; Williams, T.; Helmersson, S.

    2001-01-01

    The accurate estimation of reactor physics parameters related to the presence of cruciform absorber blades. In Boiling Water Reactors (BWR) is important for safety assessment, and for achieving a flexible operation during the cycle. Characteristics which are affected strongly include the power distribution for controlled core regions and its impact on linear heat generation rate margins, as well as the build-up of plutonium, and its influence on core excess reactivity and the reactivity worth of the shutdown system. PSI and the Swiss Nuclear Utilities (UAK) are conducting an experimental reactor physics programme related to modern Light Water Reactor (LWR) fuel assemblies, as employed in the Swiss nuclear power plants: the so-called. LWR-PROTEUS Phase I project. A significant part of this project has been devoted to the characterization of highly heterogeneous BWR fuel elements in the presence of absorber blades. The paper presents typical results for the performance of modern lattice codes in the estimation of controlled assembly reaction rate distributions, the sensitivity to the geometrical and material characterization, and a preliminary comparison of reflected-test-zone calculations with experimental reaction rate distributions measured in a Westinghouse SVEA-96+ assembly under full-density water moderation conditions in the presence of Westinghouse boron-carbide absorber blades. (author)

  11. 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.)

  12. The HAMBO BWR simulator of HAMMLAB

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-02-15

    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)

  13. BWR water chemistry guidelines in Japan

    International Nuclear Information System (INIS)

    Hirano, Hideo; Usui, Naoshi; Kitajima, Hideaki

    2014-01-01

    Through Nuclear Power Plant (NPP) operations for more than 40 years in Japan, sustainable developments of water chemistry technologies, including state-of-the-art and unique R and D results, have been developed to achieve the highest performance of maintaining NPP reactor and coolant system component integrity and fuel reliability and excellent dose rate reduction in the world. Based on this long-term operational experience, Boiling Water Reactor (BWR) utilities have established effective and reasonable water chemistry parameters with reference to safety regulations and municipal laws. However, only a few parameters are regulated by control target to ensure the long-term reliability of the reactor and coolant system. After the Fukushima Daiichi Nuclear Power Station accident, it is even more necessary to improve the safety levels of plant operation and increase the transparency of plant management in order to recover the public's confidence forward nuclear energy. Based on these situations, the BWR water chemistry guidelines task group has been working to establish the guidelines under the Standards Committee of Atomic Energy Society of Japan (AESJ). The draft will be deliberated on and approved by the Standards Committee after it receives public comments. The Japanese BWR water chemistry guidelines should provide the world-class strategies for maintaining the long-term integrity of materials and fuel. Concretely, the guidelines should provide reasonable 'control values', 'diagnostic values' and 'action levels' for multiple parameters and suggest the necessary actions to exceed the action levels. This paper provides a framework for the preliminary guidelines for BWR primary water chemistry. (author)

  14. Applied methods for mitigation of damage by stress corrosion in BWR type reactors

    International Nuclear Information System (INIS)

    Hernandez C, R.; Diaz S, A.; Gachuz M, M.; Arganis J, C.

    1998-01-01

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

  15. Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

    International Nuclear Information System (INIS)

    Greneche, D.

    2014-01-01

    This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

  16. Range of the radiation monitor for the rigid vent of primary containment during normal and emergency operation for a BWR-5 in Laguna Verde; Rango del monitor de radiacion para el venteo rigido de la contencion primaria durante operacion normal y emergencia para un reactor BWR-5 en Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Tijerina S, F.; Pozos S, A. M.; Cabrera U, S.; Mata A, J. A.; Sandoval V, S.; Ovando C, R.; Vargas A, A.; Gallardo R, I.; Cruz G, M.; Amador C, C., E-mail: francisco.tijerina@cfe.gob.mx [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Subgerencia de Ingenieria, Km 44.5 Carretera Cardel-Nautla, 91476 Laguna Verde, Alto Lucero, Veracruz (Mexico)

    2014-10-15

    The earthquake followed by a tsunami, happened in March, 2011 in the coasts of oriental Japan, caused damages in the nuclear power plants 1 at 4 of Fukushima Daiichi leading to damage of the fuel in three of the reactors and to the radiation liberation to the exterior. As consequence of those events, the regulations requires that the power plants with Primary Containment type Mark I and II evaluate to have a system of rigid vent with a monitoring equipment of radiation effluents. The present work covers the rigid vent of diameter 12 of the Primary Containment, type Mark-II, of nuclear power plant of Laguna Verde in conditions of severe accident and normal operation, low regime of Extended Power Up rate (EPU - 2317 MWt), using the codes MAAP3B, MICROSHILED 5.05 and the Bardach Black Boxes methodology. As a result the measurement range of the radiation monitor that is required for monitoring the gassy liberation to the atmosphere was determined. The conclusion is that the superior limit of the range of the radiation meter during a Severe Accident is of 8.55 E + 05 R/h (8.55 E + 08 m R/h) and the superior limit in normal operation of 1.412 E-11 at 2.540 E-7 R/h (1.412 E-14 at 2.540 E-10 m R/h). (Author)

  17. 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)

  18. Decontamination techniques for BWR power generation plant

    International Nuclear Information System (INIS)

    Yoshikawa, Susumu

    1990-01-01

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

  19. Development of alternative materials for BWR fuel springs

    International Nuclear Information System (INIS)

    Uruma, Y.; Osato, T.; Yamazaki, K.

    2002-01-01

    Major sources of radioactivity introduced into reactor water of BWR were estimated fuel crud and in-core materials (especially, fuel springs). Fuel springs are used for fixation of fuel cladding tubes with spacer grid. Those are small parts (total length is only within 25 mm) and so many numbers are loaded simultaneously and then total surfaces area are calculated up to about 200 m 2 . Fuel springs are located under high radiation field and high oxidative environment. Conventional fuel spring is made of alloy-X750 which is one of nickel-based alloy and is reported to show relatively higher corrosion release rate. 58 Co and 60 Co will be released directly into reactor water from intensely radio-activated fuel springs surface and increase radioactivity concentrations in primary coolant. Corrosion release control from fuel springs is an important technical item and a development of alternative material instead of alloy-X750 for fuel spring is a key subject to achieve ultra low man-rem exposure BWR plant. In present work, alloy-X718 which started usage for PWR fuel springs and stainless steel type 316L which has many mechanical property data are picked up for alternative materials and compared their corrosion behaviors with conventional material. Corrosion experiment was conducted under vapor-water two phases flow which is simulated fuel cladding surface boiling condition. After exposure, corrosion film formed under corrosion test was analyzed in detail and corrosion film amount and corrosion release amount are estimated among three materials. (authors)

  20. Calculation of point isotropic buildup factors of gamma rays for water and lead

    Directory of Open Access Journals (Sweden)

    A. S. H.

    2001-12-01

    Full Text Available   Exposure buildup factors for water and lead have been calculated by the Monte-Carlo method for an isotropic point source in an infinite homogeneous medium, using the latest cross secions available on the Internet. The types of interactions considered are ,photoelectric effect, incoherent (or bound-electron Compton. Scattering, coherent (or Rayleigh scattering and pair production. Fluorescence radiations have also been taken into acount for lead. For each material, calculations were made at 10 gamma ray energies in the 40 keV to 10 MeV range and up to penetration depths of 10 mean free paths at each energy point. The results presented in this paper can be considered as modified gamma ray exposure buildup factors and be used in radiation shielding designs.

  1. Effect of tungsten-187 in primary coolant on dose rate build-up in Vandellos 2

    International Nuclear Information System (INIS)

    Fernandez Lillo, E.; Llovet, R.; Boronat, M.

    1994-01-01

    The present work proposes a relationship between the Cobalt-60 piping deposited activity and the relatively high levels of Tungsten-187 in the coolant of Vandellos 2. The conclusions of this work can be applicable to other plants, since it proposes a tool to estimate and quantify the contribution of stellite to the generation of Cobalt-60 and the radiation dose build-up. (authors). 7 figs., 6 refs

  2. Monte Carlo simulation of photon buildup factors for shielding materials in diagnostic x-ray facilities

    International Nuclear Information System (INIS)

    Kharrati, Hedi; Agrebi, Amel; Karoui, Mohamed Karim

    2012-01-01

    Purpose: A simulation of buildup factors for ordinary concrete, steel, lead, plate glass, lead glass, and gypsum wallboard in broad beam geometry for photons energies from 10 keV to 150 keV at 5 keV intervals is presented. Methods: Monte Carlo N-particle radiation transport computer code has been used to determine the buildup factors for the studied shielding materials. Results: An example concretizing the use of the obtained buildup factors data in computing the broad beam transmission for tube potentials at 70, 100, 120, and 140 kVp is given. The half value layer, the tenth value layer, and the equilibrium tenth value layer are calculated from the broad beam transmission for these tube potentials. Conclusions: The obtained values compared with those calculated from the published data show the ability of these data to predict shielding transmission curves. Therefore, the buildup factors data can be combined with primary, scatter, and leakage x-ray spectra to provide a computationally based solution to broad beam transmission for barriers in shielding x-ray facilities.

  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. Dose rate reduction method for NMCA applied BWR plants

    International Nuclear Information System (INIS)

    Nagase, Makoto; Aizawa, Motohiro; Ito, Tsuyoshi; Hosokawa, Hideyuki; Varela, Juan; Caine, Thomas

    2012-09-01

    BRAC (BWR Radiation Assessment and Control) dose rate is used as an indicator of the incorporation of activated corrosion by products into BWR recirculation piping, which is known to be a significant contributor to dose rate received by workers during refueling outages. In order to reduce radiation exposure of the workers during the outage, it is desirable to keep BRAC dose rates as low as possible. After HWC was adopted to reduce IGSCC, a BRAC dose rate increase was observed in many plants. As a countermeasure to these rapid dose rate increases under HWC conditions, Zn injection was widely adopted in United States and Europe resulting in a reduction of BRAC dose rates. However, BRAC dose rates in several plants remain high, prompting the industry to continue to investigate methods to achieve further reductions. In recent years a large portion of the BWR fleet has adopted NMCA (NobleChem TM ) to enhance the hydrogen injection effect to suppress SCC. After NMCA, especially OLNC (On-Line NobleChem TM ), BRAC dose rates were observed to decrease. In some OLNC applied BWR plants this reduction was observed year after year to reach a new reduced equilibrium level. This dose rate reduction trends suggest the potential dose reduction might be obtained by the combination of Pt and Zn injection. So, laboratory experiments and in-plant tests were carried out to evaluate the effect of Pt and Zn on Co-60 deposition behaviour. Firstly, laboratory experiments were conducted to study the effect of noble metal deposition on Co deposition on stainless steel surfaces. Polished type 316 stainless steel coupons were prepared and some of them were OLNC treated in the test loop before the Co deposition test. Water chemistry conditions to simulate HWC were as follows: Dissolved oxygen, hydrogen and hydrogen peroxide were below 5 ppb, 100 ppb and 0 ppb (no addition), respectively. Zn was injected to target a concentration of 5 ppb. The test was conducted up to 1500 hours at 553 K. Test

  6. 1100 MW BWR power station

    International Nuclear Information System (INIS)

    Anon.

    1983-01-01

    Now, the start-up test of No. 2 plant in Fukushima No. 2 Nuclear Power Station is smoothly in progress, and the start of its commercial operation is scheduled at the beginning of 1984. Here, the main features of No. 2 plant including piping design are explained. For No. 2 plant, many improving measures were adopted as the base plant of the improvement and standardization project of the Ministry of International Trade and Industry, such as the adoption of Mark-2 improved PCV, the adoption of an intermediate loop in the auxiliary cooling system, one-body forging of the lower end cover of the reactor pressure vessel, the adoption of many curved pipes, the adoption of large one-body structural components in reactor recirculation system piping and so on, which are related to the reduction of radiation exposure and the improvement of plant reliability in operation and regular inspection. Also, in order to do general adjustment in the arrangement of equipment and piping and in route design, and to establish the rational construction work plan, model engineering was adopted. In No. 2 plant, a remote-controlled automatic and semiautomatic ultrasonic flaw detection system was adopted to reduce radiation exposure in in-service inspection. Automatic welding was adopted to improve the quality. (Kako, I.)

  7. 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

  8. 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

  9. Fuel cycle cost projections. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Clark, L.L.; Chockie, A.D.

    1979-12-01

    This report estimates current and future costs associated with the light water reactor nuclear fuel cycle for both once-through and thermal recycle cases. Using a range of future nuclear power generating scenarios, process flows are developed for each segment of the nuclear fuel cycle. Capital and operating costs are estimated and are combined with the process flows to generate unit cost projections for each fuel cycle segment. The unit costs and process flows are combined in the NUCOST program to estimate fuel cycle power costs through the year 2020. The unit costs are also used to estimate the fuel costs of an individual model PWR and BWR.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

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

  11. Exposure buildup factors for a cobalt-60 point isotropic source for single and two layer slabs

    International Nuclear Information System (INIS)

    Chakarova, R.

    1992-01-01

    Exposure buildup factors for point isotropic cobalt-60 sources are calculated by the Monte Carlo method with statistical errors ranging from 1.5 to 7% for 1-5 mean free paths (mfp) thick water and iron single slabs and for 1 and 2 mfp iron layers followed by water layers 1-5 mfp thick. The computations take into account Compton scattering. The Monte Carlo data for single slab geometries are approximated by Geometric Progression formula. Kalos's formula using the calculated single slab buildup factors may be applied to reproduce the data for two-layered slabs. The presented results and discussion may help when choosing the manner in which the radiation field gamma irradiation units will be described. (author)

  12. Metal transfer and build-up in friction and cutting

    CERN Document Server

    Kuznetsov, V D

    1956-01-01

    Metal Transfer and Build-up in Friction and Cutting aims to systematize our knowledge of the metal build-up, to describe some of the investigations past and present carried out in SFTI (Tomsk), and to make an effort to explain a number of the phenomena in cutting, scratching, and sliding from the point of view of metal transfer theory. The book opens with a chapter on the temperature of the rubbing interface of two solids. This temperature is needed in order to elucidate the nature of the formation of a build-up in scratching, cutting, and sliding. Separate chapters follow on the seizure phen

  13. Mobile crud and transportation of radioactivity in BWR

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-07-01

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

  14. 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)

  15. 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.)

  16. 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)

  17. 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)

  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. LBB application in Swedish BWR design

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

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

  20. 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.)

  1. 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.

  2. 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)

  3. Evaluation of BWR emergency procedure guidelines for BWR ATWS using RAMONA-3B code

    International Nuclear Information System (INIS)

    Neymotin, L.; Slovik, G.; Cazzoli, E.; Saha, P.

    1985-01-01

    An MSIV Closure ATWS calculation for a typical BWR/4 (Browns Ferry, Unit 1) was performed using the RAMONA-3B code which is a BWR systems transient code combining three-dimensional neutronic core representation with multi-channel one-dimensional thermal hydraulics. The main objective of the study was to perform a best-estimate evaluation of the recently proposed Emergency Procedure Guidelines for Anticipated Transients Without Scram (ATWS). Emphasis was placed on evaluating the effects of lowering the downcomer water level to the Top of Active Fuel (TAF) and vessel depressurization. The calculation was run up to approximately 1200 seconds. Both actions, namely, lowering the water level and vessel depressurization, lowered the reactor power to some extent. However, the pressure suppression pool water temperature still reached approximately 90 0 C (potential High Pressure Coolant Injection (HPCI) pump seal failure temperature) in twenty minutes. Thus, other actions such as boron injection and/or manual control rod insertion are necessary to mitigate a BWR/4 Main Steam Isolation Valve (MSIV) closure ATWS. 19 refs., 14 figs., 3 tabs

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

    International Nuclear Information System (INIS)

    King, D.B.

    1989-07-01

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

  5. Application of TRAB to BWR transient and accident analyses

    International Nuclear Information System (INIS)

    Raety, H.

    1995-01-01

    TRAB is a one-dimensional reactor dynamics code developed at VTT (Technical Research Centre of Finland) for BWRs. It models the core and the main circulation system inside the reactor vessel, including the steam dome with related systems, steam lines, recirculation pumps, incoming and outgoing flows as well as control and protection systems. The core model includes a one-dimensional description of the geometry, neutronics, rod heat transfer, and thermal hydraulics, using presently at most three parallel axial hydraulic channels. A synthesis model in the neutronics allows to include radially asymmetric phenomena in the modelling. TRAB has been extensively used for analyses of the TVO ABB Atom type BWRs in Olkiluoto, including analyses of pump trip, overpressurization, rod drop, ATWS, and stability. Analyses by VTT have been carried out both for the TVO Power Company and the Finnish Centre for Radiation and Nuclear Safety (STUK). The code is also used for BWR analyses at STUK. (orig.) (12 refs., 5 figs., 1 tab.)

  6. 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

  7. 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)

  8. 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.)

  9. Fast spectrum transmutation in a BWR

    International Nuclear Information System (INIS)

    Wallenius, J.; Westlen, D.

    2007-01-01

    We propose an innovative fuel design for boiling water reactors, where the use of metallic alloy fuel and hafnium clad results in a fast neutron spectrum, suitable for transmutation of minor actinides. Monte Carlo calculations made with the JEFF3.1 data library show that in the top of an up-rated BWR, it is possible to achieve fission probabilities for even neutron number nuclides similar to those of sodium-cooled reactors. Thus, from a strict neutronic perspective, multi-recycling of americium and curium could be performed in the top of BWRs. Fuel and coolant temperature feedbacks remain within acceptable ranges, but control rod worths are reduced to the extent that enriched boron might have to be used to ensure sufficient shutdown margins. (authors)

  10. 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

  11. 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)

  12. 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

  13. Agitation device in BWR type reactor

    International Nuclear Information System (INIS)

    Nakamizo, Hiroshi.

    1988-01-01

    Purpose: To secure the integrity of a pressure vessel upon reactor accident by maintaining the water temperature in the pressure suppression pool at a low temperature. Constitution: The pressure vessel and the pressure control pool in a BWR type reactor are connected with a vent pipe by way of a safety valve. Steams generated in the pressure vessel are introduced through the vent pipe into pool water. A turbine is disposed at the midway of the vent pipe and rotated with steams flowing through the vent pipe. A vane is disposed in the pressure suppression pool for agitating water and rotated by the turbine. The temperature distribution for the water in the pool is made uniform by the stirring. In this way, by rotating the vane using steams, since water can automatically be agitated as required without utilizing any power source of other system such as an electric power supply, pool water can be cooled at a high reliability. (Ikeda, J.)

  14. 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

  15. 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)

  16. Carbonate Buildups in the Pernambuco Basin, NE Brazil.

    Science.gov (United States)

    Buarque, Bruno V; Barbosa, José A; Oliveira, Jefferson T C; Magalhães, José R G; Correia, Osvaldo J

    2017-01-01

    The Pernambuco Basin represents one of the most prominent regions for deep water oil and gas exploration off the Brazilian coast. This study aims to identify and describe the occurrence of carbonate buildups in the offshore regions of the basin. The study was based on an analysis of a set of 143 2D time-migrated seismic sections that cover the offshore region of the Pernambuco Basin. An interpretation of the seismic dataset was used to define the main seismic sequences related to the main regional pulses of deposition and to identify three main groups of carbonate buildups: 1) shelf margin reefs, 2) patch reefs, and 3) isolated carbonate buildups. The carbonate buildups formed in two main intervals during post-rift sequences of the Santonian-Maastrichtian and Paleocene-Middle Miocene, which extend the known periods of carbonate deposit formation in the basin. The formation of carbonate buildups was controlled by the tectonic evolution of the Pernambuco Plateau, which created a series of paleotopographic highs that enabled the establishment of oceanographically and climatically favorable conditions for carbonate formation and deposition. These findings are important for understanding the evolution of the basin and for future evaluations of its petroleum potential.

  17. Carbonate Buildups in the Pernambuco Basin, NE Brazil

    Directory of Open Access Journals (Sweden)

    BRUNO V. BUARQUE

    Full Text Available ABSTRACT The Pernambuco Basin represents one of the most prominent regions for deep water oil and gas exploration off the Brazilian coast. This study aims to identify and describe the occurrence of carbonate buildups in the offshore regions of the basin. The study was based on an analysis of a set of 143 2D time-migrated seismic sections that cover the offshore region of the Pernambuco Basin. An interpretation of the seismic dataset was used to define the main seismic sequences related to the main regional pulses of deposition and to identify three main groups of carbonate buildups: 1 shelf margin reefs, 2 patch reefs, and 3 isolated carbonate buildups. The carbonate buildups formed in two main intervals during post-rift sequences of the Santonian-Maastrichtian and Paleocene-Middle Miocene, which extend the known periods of carbonate deposit formation in the basin. The formation of carbonate buildups was controlled by the tectonic evolution of the Pernambuco Plateau, which created a series of paleotopographic highs that enabled the establishment of oceanographically and climatically favorable conditions for carbonate formation and deposition. These findings are important for understanding the evolution of the basin and for future evaluations of its petroleum potential.

  18. BWR spent fuel storage cask performance test. Volume 1. Cask handling experience and decay heat, heat transfer, and shielding data

    International Nuclear Information System (INIS)

    McKinnon, M.A.; Doman, J.W.; Tanner, J.E.; Guenther, R.J.; Creer, J.M.; King, C.E.

    1986-02-01

    This report documents a heat transfer and shielding performance test conducted on a Ridihalgh, Eggers and Associates REA 2023 boiling water reactor (BWR) spent fuel storage cask. The testing effort consisted of three parts: pretest preparations, performance testing, and post-test activities. Pretest preparations included conducting cask handling dry runs and characterizing BWR spent fuel assemblies from Nebraska Public Power District's Cooper Nuclear Station. The performance test matrix included 14 runs consisting of two loadings, two cask orientations, and three backfill environments. Post-test activities included calorimetry and axial radiation scans of selected fuel assemblies, in-basin sipping of each assembly, crud collection, video and photographic scans, and decontamination of the cask interior and exterior

  19. An economic analysis of BWR control rod blade management strategies. Final report

    International Nuclear Information System (INIS)

    Welsh, J.

    1995-12-01

    Nuclear power plants have available a number of alternative courses of action that can contribute to the reduction of personnel exposure to radiation. Possible actions at boiling water reactor (BWR) plants include accelerating the replacement of high-cobalt control rod blades (CRB) or the blades' high-cobalt pins and rollers with low or non-cobalt substitutes. To help utilities understand the exposure reduction and the economic costs and benefits associated with management alternatives, such as accelerated replacement of blades, pins and rollers, EPRI has initiated a project called Cost/Benefit Software for Analyses of Radiation Control Measures (RP1935-32). Through this project EPRI will incorporate engineering-economic techniques into a series of analytical tools that will provide useful insights about alternative exposure reduction options. Prototype software has been developed in an Excel worksheet to analyze issues associated with BWR control rod blade management options. The CRB replacement problem framework and analysis methodology incorporated into the software tool will help plant managers consider explicitly key engineering and economic issues that are relevant to exposure reduction decisions. This tool generates results that can help plant managers make decisions that are fiscally wise by showing all the cost and benefit implications associated with a management action under consideration. This report describes the general analytical approach for evaluating exposure reduction alternatives. The methodology used to analyze blade and pin and roller replacement alternatives, and the results of a case study application of the methodology and the software prototype at Commonwealth Edison

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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.)

  5. 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.)

  6. 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.)

  7. 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.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Cancino P, G.

    2016-10-01

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

  9. 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

  10. 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)

  11. 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)

  12. 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

  13. 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)

  14. 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)

  15. Fuel relocation effects in BWR LOCA conditions

    International Nuclear Information System (INIS)

    Raul Orive Moreno; Ines Gallego Cabezon; Pablo Julio Garcia Sedano; Yolanda Tofino Gomez; Pedro Mata Alonso

    2005-01-01

    One of the objectives of the EPRI's Fuel Reliability Program is to establish the bases for the licensing of nuclear fuel to burnup levels beyond the current licensed value of 62 GWd/MT rod average burnup. One of the licensing points of concern is the behavior of the high burnup fuel in LOCA conditions. To respond to this concern a series of LOCA experiments are being performed at Argonne National Laboratory using fuel rods from Limerick NPP at 57 GWd/MT and H.B. Robinson at 67 GWd/MT. ANL LOCA tests indicate potential fuel relocation during LOCA. This could result in an increase of LHGR during a real plant LOCA. This report presents the LOCA analyses performed by IBERDROLA (Spanish utility), using results from the Cofrentes NPP (BWR-6) LOCA evaluations with GE-14 fuel design for the whole exposure range, quantifying fuel relocation impact. This effect has been modeled and implemented in FRAP-T6/APK (vendor independent IBERDROLA licensing thermomechanical code), as well as the wall-to-fluid heat transfer area increase in the ballooned region. Separate and combined impacts on PCT and ECR values can be evaluated with this modified code version. A new hoop strain versus rupture temperature curve is implemented in code, starting from NUREG-0630 model data base, but with a more best-estimate fit, in order to reproduce expected experimental values. The increase of heat transfer in the ballooned region has been validated with Halden LOCA tests. Preliminary results indicate that the effect of fuel relocation is expected to be compensated by the increased heat transfer area. This effect is to be confirmed with the Halden LOCA tests in progress. (author)

  16. 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)

  17. Decay heat measurements and predictions of BWR spent fuel

    International Nuclear Information System (INIS)

    McKinnon, M.A.; Heeb, C.M.; Creer, A.M.

    1986-06-01

    Pre-calorimetry decay heat predictions obtained with the ORIGEN2 computer code were compared to calorimeter data obtained for eleven boiling water reactor (BWR) spent fuel assemblies using General Electric, Morris Operation's in-pool calorimeter. Ten of the 7 x 7 BWR spent fuel assemblies were obtained from Nebraska Public Power District's Cooper Nuclear Station. The remaining BWR assembly was from Commonwealth Edison's Dresden Nuclear Power Plant. The assemblies had burnups ranging from 5.3 to 27.6 GWD/MTU and had been removed from their respective reactors for 2 or more years. The majority of the assemblies had burnups of between 20 and 28 GWD/MTU and had been out of the reactor 2 to 4 years. The assemblies represent spent fuel that has been continuously burned and fuel that has been reinserted. Comparisons of ORIGEN2 pre-calorimetry decay heat predictions with calorimeter data showed that predictions agreed with data within the precision/repeatibility of the experimental data (+-15 Watts or 5% for a 300 Watt BWR assembly). Comparisons of predicted axial gamma profiles based on core-averaged axial burnups with measured profiles showed difference. These differences may be explained by reactor operation with partially inserted control rods

  18. 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)

  19. 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.)

  20. Humidity Buildup in Electronic Enclosures Exposed to Constant Conditions

    DEFF Research Database (Denmark)

    Conseil, Helene; Staliulionis, Zygimantas; Jellesen, Morten Stendahl

    2017-01-01

    Electronic components and devices are exposed to a wide variety of climatic conditions, therefore the protection of electronic devices from humidity is becoming a critical factor in the system design. The ingress of moisture into typical electronic enclosures has been studied with defined paramet....... The moisture buildup inside the enclosure has been simulated using an equivalent RC circuit consisting of variables like controlled resistors and capacitors to describe the diffusivity, permeability, and storage in polymers....

  1. 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

  2. Quantitative assessment of the effect of corrosion product buildup on occupational exposure

    International Nuclear Information System (INIS)

    Divine, J.R.

    1982-10-01

    The program was developed to provide a method for predicting occupational exposures caused by the deposition of radioactive corrosion products outside the core of the primary system of an operating power reactor. This predictive capability will be useful in forecasting total occupational doses during maintenance, inspection, decontamination, waste treatment, and disposal. In developing a reliable predictive model, a better understanding of the parameters important to corrosion product film formation, corrosion product transport, and corrosion product film removal will be developed. This understanding can lead to new concepts in reactor design to minimize the buildup and transport of radioactive corrosion products or to improve methods of operation. To achieve this goal, three objectives were established to provide: (1) criteria for acceptable coolant sampling procedures and sampling equipment that will provide data which will be used in the model development; (2) a quantitative assessment of the effect of corrosion product deposits on occupational exposure; and (3) a model which describes the influence of flow, temperature, coolant chemistry, construction materials, radiation, and other operating parameters on the transport and buildup of corrosion products

  3. 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)

  4. Feasibility studies of computed tomography in partial defect detection of spent BWR fuel

    International Nuclear Information System (INIS)

    Levai, F.; Tikkinen, J.; Tarvainen, M.; Arlt, R.

    1990-10-01

    Feasibility studies were made for tomographic reconstruction of a cross-sectional activity distribution of a spent nuclear fuel assembly. The purpose was to determine the number of fuel rods (pins) and localize the positisons where pins are missing. The activity distribution map showing the locations of fuel rods in the assembly was reconstructed. The theoretical part of this work consists of simulation of image reconstruction based on theoretically calculated data from a reference assembly model. Evaluation of different image reconstruction techniques was made. Measurements were made in real facility conditions. Gamma radiation from an irradiated 8 x 8 - 1 BWR fuel assembly was measured through a narrow custom made collimator from different angles and positions. The measured data set was used as projections for reconstructing the activity profile of the assembly in cross-sectional plane

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

    International Nuclear Information System (INIS)

    Thakre, S.; Ma, W.

    2013-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-08-15

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

  7. Monte Carlo simulation of x-ray buildup factors of lead and its applications in shielding of diagnostic x-ray facilities

    International Nuclear Information System (INIS)

    Kharrati, Hedi; Agrebi, Amel; Karaoui, Mohamed-Karim

    2007-01-01

    X-ray buildup factors of lead in broad beam geometry for energies from 15 to 150 keV are determined using the general purpose Monte Carlo N-particle radiation transport computer code (MCNP4C). The obtained buildup factors data are fitted to a modified three parameter Archer et al. model for ease in calculating the broad beam transmission with computer at any tube potentials/filters combinations in diagnostic energies range. An example for their use to compute the broad beam transmission at 70, 100, 120, and 140 kVp is given. The calculated broad beam transmission is compared to data derived from literature, presenting good agreement. Therefore, the combination of the buildup factors data as determined and a mathematical model to generate x-ray spectra provide a computationally based solution to broad beam transmission for lead barriers in shielding x-ray facilities

  8. Estimation of dose rate around the spent control rods of a BWR

    International Nuclear Information System (INIS)

    Cancino P, G.

    2016-01-01

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

  9. Buildup of 236U in the gaseous diffusion plant product

    International Nuclear Information System (INIS)

    Ford, J.S.

    1975-01-01

    A generalized projection of the average annual 236 U concentration that can be expected in future enriched uranium product from the US-ERDA gaseous diffusion plants when reprocessed fuels become available for cascade feeding is given. It is concluded that the buildup of 236 U is not an ever-increasing function, but approaches a limiting value. Projected concentrations result in only slight separative work losses and present no operational problem to ERDA in supplying light water reactor requirements. The use of recycle uranium from power reactor spent fuels will result in significant savings in natural uranium feed

  10. A comprehensive study of the energy absorption and exposure buildup factors of different bricks for gamma-rays shielding

    Directory of Open Access Journals (Sweden)

    M.I. Sayyed

    Full Text Available The present investigation has been performed on different bricks for the purpose of gamma-ray shielding. The values of the mass attenuation coefficient (µ/ρ, energy absorption buildup factor (EABF and exposure buildup factor (EBF were determined and utilized to assess the shielding effectiveness of the bricks under investigation. The mass attenuation coefficients of the selected bricks were calculated theoretically using WinXcom program and compared with MCNPX code. Good agreement between WinXcom and MCNPX results was observed. Furthermore, the EABF and EBF have been discussed as functions of the incident photon energy and penetration depth. It has been found that the EABF and EBF values are very large in the intermediate energy region. The steel slag showed good shielding properties, consequently, this brick is eco-friendly and feasible compared with other types of bricks used for construction. The results in this work should be useful in the construction of effectual shielding against hazardous gamma-rays. Keywords: Brick, Mass attenuation coefficient, Buildup factor, G-P fitting, Radiation shielding

  11. Radiation

    International Nuclear Information System (INIS)

    2013-01-01

    The chapter one presents the composition of matter and atomic theory; matter structure; transitions; origin of radiation; radioactivity; nuclear radiation; interactions in decay processes; radiation produced by the interaction of radiation with matter

  12. 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

  13. Recent developments in ultrasonic inspection of BWR pipes in USA

    International Nuclear Information System (INIS)

    Behravesh, M.; Dau, G.J.

    1985-01-01

    Following the 1982 discovery of widespread intergranular stress corrosion cracking (IGSCC) in the recirculation piping system at the Nine Mile Point unit 1 plant, the U.S. Nuclear Regulatory Commission (NRC) made the requirement that the operating BWR plants demonstrate the effectiveness of their ultrasonic testing procedures. Since that time, the detection, identification, and sizing of IGSCC in BWR pipes has received an even greater amount of attention. Initially EPRI through its NDE Center in Charlotte took an active role in providing a mechanism whereby the affected utilities could demonstrate the required capabilities. This effort was soon followed with the development and regular offering of a one week course on IGSCC detection whose final exam was patterned after the requirements of the US NRC Bulletin IE 83-02 on performance capability demonstration

  14. A closed-form formulation for the build-up factor and absorbed energy for photons and electrons in the Compton energy range in Cartesian geometry

    International Nuclear Information System (INIS)

    Borges, Volnei; Vilhena, Marco Tullio; Fernandes, Julio Cesar Lombaldo

    2011-01-01

    In this work, we report on a closed-form formulation for the build-up factor and absorbed energy, in one and two dimensional Cartesian geometry for photons and electrons, in the Compton energy range. For the one-dimensional case we use the LTS N method, assuming the Klein-Nishina scattering kernel for the determination of the angular radiation intensity for photons. We apply the two-dimensional LTS N nodal solution for the averaged angular radiation evaluation for the two-dimensional case, using the Klein-Nishina kernel for photons and the Compton kernel for electrons. From the angular radiation intensity we construct a closed-form solution for the build-up factor and evaluate the absorbed energy. We present numerical simulations and comparisons against results from the literature. (author)

  15. A closed-form formulation for the build-up factor and absorbed energy for photons and electrons in the Compton energy range in Cartesian geometry

    Energy Technology Data Exchange (ETDEWEB)

    Borges, Volnei; Vilhena, Marco Tullio, E-mail: borges@ufrgs.b, E-mail: vilhena@pq.cnpq.b [Universidade Federal do Rio Grande do Sul (PROMEC/UFRGS), Porto Alegre, RS (Brazil). Programa de Pos-Graduacao em Engenharia Mecanica; Fernandes, Julio Cesar Lombaldo, E-mail: julio.lombaldo@ufrgs.b [Universidade Federal do Rio Grande do Sul (DMPA/UFRGS), Porto Alegre, RS (Brazil). Dept. de Matematica Pura e Aplicada. Programa de Pos Graduacao em Matematica Aplicada

    2011-07-01

    In this work, we report on a closed-form formulation for the build-up factor and absorbed energy, in one and two dimensional Cartesian geometry for photons and electrons, in the Compton energy range. For the one-dimensional case we use the LTS{sub N} method, assuming the Klein-Nishina scattering kernel for the determination of the angular radiation intensity for photons. We apply the two-dimensional LTS{sub N} nodal solution for the averaged angular radiation evaluation for the two-dimensional case, using the Klein-Nishina kernel for photons and the Compton kernel for electrons. From the angular radiation intensity we construct a closed-form solution for the build-up factor and evaluate the absorbed energy. We present numerical simulations and comparisons against results from the literature. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-01

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

  17. 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

  18. 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

  19. BWR Mark I pressure suppression pool dynamics studies

    International Nuclear Information System (INIS)

    McCauley, E.W.; Martin, R.W.; Lai, W.; Morrison, F.A.; Sutton, S.B.

    1976-11-01

    This report summarizes the initial effort of the Lawrence Livermore Laboratory involvement with the study of BWR Mark I pressure suppression pool dynamics. Analytical activity is described and calculational results are presented for several simplified geometries. Computer code authentication will be provided by a currently active program in benchmark tests. The experiment and some results are presented. A combined analytical and experimental program to evaluate air scaling hypotheses for hydrodynamic forces and pool motion is presented, along with some conclusions regarding model scaling

  20. Energy buildup in sheared force-free magnetic fields

    Science.gov (United States)

    Wolfson, Richard; Low, Boon C.

    1992-01-01

    Photospheric displacement of the footpoints of solar magnetic field lines results in shearing and twisting of the field, and consequently in the buildup of electric currents and magnetic free energy in the corona. The sudden release of this free energy may be the origin of eruptive events like coronal mass ejections, prominence eruptions, and flares. An important question is whether such an energy release may be accompanied by the opening of magnetic field lines that were previously closed, for such open field lines can provide a route for matter frozen into the field to escape the sun altogether. This paper presents the results of numerical calculations showing that opening of the magnetic field is permitted energetically, in that it is possible to build up more free energy in a sheared, closed, force-free magnetic field than is in a related magnetic configuration having both closed and open field lines. Whether or not the closed force-free field attains enough energy to become partially open depends on the form of the shear profile; the results presented compare the energy buildup for different shear profiles. Implications for solar activity are discussed briefly.

  1. Scintiscanning of arthritis and analysis of build-up curves

    International Nuclear Information System (INIS)

    Yamagishi, Tsuneo; Omori, Shigeo; Miyawaki, Haruo; Maniwa, Masato; Yoshizaki, Kenichi

    1975-01-01

    In the present study 40 knee joints with rheumatoid arthritis, 23 knee joints with osteoarthrosis deformans, 3 knee joints with non-synovitis, one knee joint with pyogenic arthritis and 4 normal knee joints were scanned. By analysis of build-up curves obtained immediately after the intravenous injection of sup(99m)Tc-pertechnetate, the rate of accumulation of radioactivity (t 1/2) in the affected joints was simultaneously estimated in order to compare them with clinical findings. 1. Scintiscanning of arthritis, rheumatoid arthritis, osteoarthrosis deformans of the knee joint, non-specific synovitis, and pyogenic arthritis of the knee joint, yielded a positive scan for all of the joint diseases. 2. In the scintigram of healthy knee joints, there are no areas of RI accumulation or right to left difference. 3. In some instances abnormal uptake of RI was seen on scintigrams of arthritis even after normal clinical and laboratory findings had been achieved with therapy. 4. sup(99m)Tc-pertechnetate, a radionuclide with a short half-life, allows repeated scans and provides a useful radiologic means of evaluating therapeutic course and effectiveness. 5. Analysis of build-up curves revealed that the rate of accumulation of RI was faster in rheumatoid arthritis than in osteoarthrosis deformans. (auth.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Matthew D. Hinds

    2001-10-17

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

  3. 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

  4. Investigation of Burnup Credit Modeling Issues Associated with BWR Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, J.C.

    2000-10-12

    Although significant effort has been dedicated to the study of burnup-credit issues over the past decade, U.S. studies to-date have primarily focused on spent pressurized-water-reactor (PWR) fuel. The current licensing approach taken by the U.S. Department of Energy for burnup credit in transportation seeks approval for PWR fuel only. Burnup credit for boiling-water-reactor (BWR) fuel has not yet been formally sought. Burnup credit for PWR fuel was pursued first because: (1) nearly two-thirds (by mass) of the total discharged commercial spent fuel in the United States is PWR fuel, (2) it can substantially increase the fuel assembly capacity with respect to current designs for PWR storage and transportation casks, and (3) fuel depletion in PWRs is generally less complicated than fuel depletion in BWRs. However, due to international needs, the increased enrichment of modern BWR fuels, and criticality safety issues related to permanent disposal within the United States, more attention has recently focused on spent BWR fuel. Specifically, credit for fuel burnup in the criticality safety analysis for long-term disposal of spent nuclear fuel enables improved design efficiency, which, due to the large mass of fissile material that will be stored in the repository, can have substantial financial benefits. For criticality safety purposes, current PWR storage and transportation canister designs employ flux traps between assemblies. Credit for fuel burnup will eliminate the need for these flux traps, and thus, significantly increase the PWR assembly capacity (for a fixed canister volume). Increases in assembly capacity of approximately one-third are expected. In contrast, current BWR canister designs do not require flux traps for criticality safety, and thus, are already at their maximum capacity in terms of physical storage. Therefore, benefits associated with burnup credit for BWR storage and transportation casks may be limited to increasing the enrichment capacity and

  5. 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)

  6. Integrated Electronic Microscopy Method to Characterize BWR Crud Deposits

    International Nuclear Information System (INIS)

    Pop, Mike G.; Lockamon, Brian; Howe, James M.; Oleshko, Vladimir P.

    2007-01-01

    The primary objective of this paper is to present the best combination of analytical Electron Microscopy (EM) techniques suited for studying Boiling Water Reactor (BWR) fuel crud deposits in their 'as found' condition, for example un-adulterated portions of the deposits located in crevices at the surface of deposit. The secondary objective of the paper is to suggest a strategy to connect the analytical EM results with the Power Diffraction File (PDF-4) crystal database to ultimately explain crystal growth phenomena in crevices. The samples analyzed as part of this work were collected from steam generators in Nuclear Power Plants. These samples were selected as segregates for BWR deposits due to their similar structures and because they were freely releasable for study in the University labs. The samples were analyzed extensively through various EM techniques at magnifications up to 150,000 X. Subsequent evaluation of the analysis results demonstrated that the selected samples exhibited characteristics that were very close to second-burned fuel crud deposits from operating BWR plants. (authors)

  7. 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

  8. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-06-01

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

  10. Simulation of Electron-Cloud Build-Up for the Cold Arcs of the LHC and Comparison with Measured Data

    CERN Document Server

    Maury Cuna, H; Rumolo, G; Tavian, L; Zimmermann, F

    2011-01-01

    The electron cloud generated by synchrotron radiation or residual gas ionization is a concern for LHC operation and performance. We report the results of simulations studies which examine the electron cloud build-up, at injection energy, 3.5 TeV for various operation parameters. In particular, we determine the value of the secondary emission yield corresponding to the multipacting threshold, and investigate the electron density, and heat as a function of bunch intensity for dipoles and field-free regions. We also include a comparison between simulations results and measured heat-load data from the LHC scrubbing runs in 2011.

  11. 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

  12. Electron-Cloud Build-Up: Theory and Data

    International Nuclear Information System (INIS)

    Furman, M.A.

    2010-01-01

    We present a broad-brush survey of the phenomenology, history and importance of the electron-cloud effect (ECE). We briefly discuss the simulation techniques used to quantify the electron-cloud (EC) dynamics. Finally, we present in more detail an effective theory to describe the EC density build-up in terms of a few effective parameters. For further details, the reader is encouraged to refer to the proceedings of many prior workshops, either dedicated to EC or with significant EC contents, including the entire 'ECLOUD' series. In addition, the proceedings of the various flavors of Particle Accelerator Conferences contain a large number of EC-related publications. The ICFA Beam Dynamics Newsletter series contains one dedicated issue, and several occasional articles, on EC. An extensive reference database is the LHC website on EC.

  13. Orbit averaged radial buildup code for tandem mirror geometry

    Energy Technology Data Exchange (ETDEWEB)

    Campbell, M.; Futch, A.H.

    1981-01-22

    The radial Fokker-Planck (RFP) model of A. Futch was modified to treat plasma buildup in the tandem mirror plug and center cell with a self-consistent model (TOARBUC). Two major changes have been made to the original version of this code. First, the center cell is treated as having separate electron and ion confining potentials with the ion potential having the opposite sign of that in a conventional mirror. Second, a two-electron-temperature treatment derived by R. Cohen was included in the present model to allow the plug and center cell to have different T/sub e/'s as observed in the experiment. The following sections explain these changes in greater detail.

  14. Detection of hydrogen buildup in initially pure nonhydrogenous liquids

    International Nuclear Information System (INIS)

    McNeany, S.R.; Jenkins, J.D.

    1978-12-01

    A technique for monitoring hydrogen buildup in initially pure nonhydrogenous liquids is described in this report. The detection method is based upon the neutron-moderating properties of hydrogen. The analysis leading to the selection and design of a hydrogen-monitoring device is described. An experimental mockup of the device was then constructed and tested for hydrogen sensitivity. A hot cell was used for these tests. A device proved capable of measuring hydrogen concentrations in the range of 0 to 13.0 x 10 27 atoms/m 3 , with an accuracy of about 1.0 x 10 27 atoms/m 3 . A typical measurement can be made in 3 to 5 min. The experimental results confirmed the sensitivities predicted by the analysis and demonstrated that such a device would be practical for hydrogen concentration measurements for criticality control in an HTGR fuel refabrication plant

  15. Electron-Cloud Build-Up: Theory and Data

    Energy Technology Data Exchange (ETDEWEB)

    Furman, M. A.

    2010-10-08

    We present a broad-brush survey of the phenomenology, history and importance of the electron-cloud effect (ECE). We briefly discuss the simulation techniques used to quantify the electron-cloud (EC) dynamics. Finally, we present in more detail an effective theory to describe the EC density build-up in terms of a few effective parameters. For further details, the reader is encouraged to refer to the proceedings of many prior workshops, either dedicated to EC or with significant EC contents, including the entire 'ECLOUD' series. In addition, the proceedings of the various flavors of Particle Accelerator Conferences contain a large number of EC-related publications. The ICFA Beam Dynamics Newsletter series contains one dedicated issue, and several occasional articles, on EC. An extensive reference database is the LHC website on EC.

  16. 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)

  17. Contour Crafting Simulation Plan for Lunar Settlement Infrastructure Build-Up

    Data.gov (United States)

    National Aeronautics and Space Administration — Economically viable and reliable building systems and tool sets are being sought, examined, and tested for extraterrestrial habitat and infrastructure buildup. This...

  18. Characterization of 2 MeV, 4 MeV, 6 MeV and 18 MeV buildup caps for use with a 0.6 cubic centimeter thimble ionization chamber

    International Nuclear Information System (INIS)

    Salyer, R.L.; VanDenburg, J.W.; Prinja, A.K.; Kirby, T.; Busch, R.; Hong-Nian Jow

    1996-07-01

    The purpose of this research is to characterize existing 2 MeV, 4 MeV and 6 MeV buildup caps, and to determine if a buildup cap can be made for the 0.6 cm 3 thimble ionization chamber that will accurately measure exposures in a high-energy photon radiation field. Two different radiation transport codes were used to computationally characterize existing 2 MeV, 4 MeV, and 6 MeV buildup caps for a 0.6 cm 3 active volume thimble ionization chamber: ITS, The Integrated TIGER Series of Coupled Electron-Photon Monte Carlo Transport Codes; and CEPXS/ONEDANT, A One-Dimensional Coupled Electron-Photon Discrete Ordinates Code Package. These codes were also used to determine the design characteristics of a buildup cap for use in the 18 MeV photon beam produced by the 14 TW pulsed power HERMES-III electron accelerator. The maximum range of the secondary electron, the depth at which maximum dose occurs, and the point where dose and collision kerma are equal have been determined to establish the validity of electronic equilibrium. The ionization chamber with the appropriate buildup cap was then subjected to a 4 MeV and a 6 MeV bremmstrahlung radiation spectrum to determine the detector response

  19. Cross-section adjustment techniques for BWR adaptive simulation

    Science.gov (United States)

    Jessee, Matthew Anderson

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

  20. 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.

  1. Experimental study of advanced continuous acoustic emission monitoring of BWR components. Final report

    International Nuclear Information System (INIS)

    McElroy, J.W.; Hartman, W.F.

    1980-09-01

    The program consisted of installing, maintaining, and monitoring AE sensors located on primary piping, nozzles, and valves in the BWR system. Analysis of the AE data was correlated to the results of supplementary nondestructive testing techniques used during the in-service inspection, performed at refueling outages. Purpose of the program was to develop the on-line surveillance acoustic emission technique in order to identify areas of possible structural degradation. Result of reducing inspection time was to reduce accumulated radiation exposure to inspecting personnel and to reduce the amount of critical plant outage time by identifying the critical inspection areas during operation. The program demonstrated the capability of acoustic emission instrumentation to endure the nuclear reactor environment. The acoustic emission sensors withstood 12 months of reactor operation at temperatures of 400 0 F and greater in high radiation fields. The preamplifiers, also mounted in the reactor environment, operated for the 12-month period in 100% humidity, 250 0 F conditions. The remaining cable and AE instrumentation were operated in controlled environments

  2. Neutron fluence effect on the IASCC susceptibility of AISI 304 stainless steel under simulated BWR conditions

    International Nuclear Information System (INIS)

    Navas, M.; Castano, M.L; Gomez-Briceno, D.; Karlsen, T.

    2004-01-01

    Full text: Neutron irradiation modifies the stress corrosion cracking (SCC) resistance of AISI 304 stainless steel in high temperature water. The microstructure and the microchemistry of materials change with the increasing of dose, inducing Radiation Hardening and Radiation Induced Segregation (RIS). SCC behaviour of irradiated material is influenced by these changes, leading to a value of threshold neutron fluence which could depend on different variables. Therefore, fully understanding of the IASCC material susceptibility implies the study of the effect of critical parameters like accumulated neutron fluence, material composition and water chemistry. Experimental work is being carried out in collaboration with the Halden Reactor Project and it includes the performance of Constant Extension Rate Tests (CERT) at CIEMAT laboratories in out-of-pile loops that simulate BWR operating conditions. The main objective is to determine the influence of neutron fluence on the SCC susceptibility of austenitic steels. A test matrix was defined to test unirradiated and irradiated specimens fabricated from tubes used previously in the Crack Initiation Test (IFA 618) performed at the Halden Reactor Project. According to the irradiation periods of IFA 618, three materials (Annealed and thermally-sensitised AISI 304 and AISI 316L) with three different accumulated neutron fluences are available. This paper presents the results obtained with annealed and thermally sensitised AISI 304 SS tested in different environments. (Author)

  3. 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)

  4. 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

  5. Improving BWR fuel critical power without increasing bundle pressure drop

    International Nuclear Information System (INIS)

    Matzner, B.; Shiraishi, L.M.; Danielson, D.W.; Congdon, S.P.

    2004-01-01

    It has been almost axiomatic that BWR fuel bundle critical power performance could not be improved without an accompanying increase in bundle pressure drop. It appeared that in order to increase the bundle dryout resistance it was necessary to perturb the bundle coolant flow paths in some fashion. This resulted in an unacceptable bundle pressure drop increase. However, by adding part length rods to decrease bundle pressure drop and by inserting an extra spacer with rearranged spacer pitch and flow trippers on the channel wall at the top of the bundle to increase critical power it was possible to achieve the goal of increased bundle critical power without pressure drop increase. (author)

  6. 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

  7. Water sorption and solubility of core build-up materials.

    Science.gov (United States)

    Zankuli, M A; Devlin, H; Silikas, N

    2014-12-01

    To investigate the variation in water sorption and solubility across a range of different core build-up materials. Five materials were tested, four of which are resin-based materials (Grandio Core, Core.X Flow, Bright Flow Core, Speedee) and one resin-modified glass ionomer (Fuji II LC). All specimens (n=10) were immersed in 10ml distilled water in individual glass containers and weighed at one week, 14 and 28 days. After a total immersion time of 28 days, 7 specimens were dried to a constant mass, in a desiccator for 28 days. Three samples of each material were not dried, but were left in distilled water for 1 year, to determine the long-term water sorption properties. Specimens were weighed at monthly intervals until 6 months and then at the 9th and 12th months. Each specimen was measured using a digital electronic caliper (Mitutoyo Corporation, Japan). After 28 days immersion, the change in water sorption and solubility of the materials ranged from 12.9 to 67.1μg/mm(3) (PGrandio Core had the lowest water sorption and solubility among the tested materials. According to the ISO 4049 standards, all the tested materials showed acceptable water sorption and solubility, apart from the water sorption behavior of Fuji II LC. Copyright © 2014 Academy of Dental Materials. Published by Elsevier Ltd. All rights reserved.

  8. The Mechanism for Energy Buildup in the Solar Corona

    Science.gov (United States)

    Antiochos, Spiro; Knizhnik, Kalman; DeVore, Richard

    2017-10-01

    Magnetic reconnection and helicity conservation are two of the most important basic processes determining the structure and dynamics of laboratory and space plasmas. The most energetic dynamics in the solar system are the giant CMEs/flares that produce the most dangerous space weather at Earth, yet may also have been essential for the origin of life. The origin of these explosions is that the lowest-lying magnetic flux in the Sun's corona undergoes the continual buildup of stress and free energy that can be released only through explosive ejection. We perform MHD simulations of a coronal volume driven by quasi-random boundary flows designed to model the processes by which the solar interior drives the corona. Our simulations are uniquely accurate in preserving magnetic helicity. We show that even though small-scale stress is injected randomly throughout the corona, the net result of magnetic reconnection is a coherent stressing of the lowest-lying field lines. This highly counter-intuitive result - magnetic stress builds up locally rather than spreading out to a minimum energy state - is the fundamental mechanism responsible for the Sun's magnetic explosions. It is likely to be a mechanism that is ubiquitous throughout laboratory and space plasmas. This work was supported by the NASA LWS and SR Programs.

  9. 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.

  10. BWR online monitoring system based on noise analysis

    International Nuclear Information System (INIS)

    Ortiz-Villafuerte, Javier; Castillo-Duran, Rogelio; Alonso, Gustavo; Calleros-Micheland, Gabriel

    2006-01-01

    A monitoring system for during operation early detection of an anomaly and/or faulty behavior of equipment and systems related to the dynamics of a boiling water reactor (BWR) has been developed. The monitoring system is based on the analysis of the 'noise' or fluctuations of a signal from a sensor or measurement device. An efficient prime factor algorithm to compute the fast Fourier transform allows the continuous, real-time comparison of the normalized power spectrum density function of the signal against previously stored reference patterns in a continuously evolving matrix. The monitoring system has been successfully tested offline. Four examples of the application of the monitoring system to the detection and diagnostic of faulty equipment behavior are presented in this work: the detection of two different events of partial blockage at the jet pump inlet nozzle, miss-calibration of a recirculation mass flow sensor, and detection of a faulty data acquisition card. The events occurred at the two BWR Units of the Laguna Verde Nuclear Power Plant. The monitoring system and its possible coupling to the data and processing information system of the Laguna Verde Nuclear Power Plant are described. The signal processing methodology is presented along with the introduction of the application of the evolutionary matrix concept for determining the base signature of reactor equipment or component and the detection of off normal operation conditions

  11. 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.)

  12. 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.

  13. 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)

  14. 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

  15. 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

  16. BWR stability analysis with the BNL Engineering Plant Analyzer

    International Nuclear Information System (INIS)

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

    1992-10-01

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

  17. FASMON monitoring of space dependent oscillations in a BWR

    International Nuclear Information System (INIS)

    Bergdahl, B.G.; Oguma, R.

    1994-01-01

    BWR-instabilities can occur in fundamental mode with in-phase core wide oscillations or in excited modes as out-of-phase oscillations. The fundamental mode can be successfully monitored with DR (Decay Ratio) and amplitude calculations while the out-of-phase oscillations demands phase analysis between the LPRM-signals in different parts of the core. The present report describes how the BWR-stability monitor FASMON detects such excited modes and presents the topology of the flux swing in a horizontal plane in the core. FASMON is the only available method to detect out-of-phase oscillations fast (within 20 seconds) and robust. The FASMON algorithm is based on adaptive recursive filtering technique. All investigations with FASMON presented in this report show that the software is a good monitor tool. The software makes surveillance of the phase differences and amplitudes for a number of LPRM signals. The method to calculate the phase differences is robust with the following characteristics: The response time within 20 seconds for the FASMON algorithm is very short; Robust - works even under transient conditions; Independent of the signal amplitude; A dead sensor in a LPRM-pair does not cause alarm. The report gives results from investigations using FASMON both on theoretical and measured signals. As a conclusion the investigation of the instability shows that FASMON works even under circumstances with in-phase conditions and low amplitude up to unstable conditions with out-of-phase oscillations and high amplitude. 5 refs, 12 figs

  18. 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)

  19. 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)

  20. 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)

  1. 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

  2. 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.)

  3. 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

  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. 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

  6. Utility experience with the BWR power shape monitoring system

    Energy Technology Data Exchange (ETDEWEB)

    Emrich, W.J. Jr.

    1986-01-01

    The Oyster Creek Nuclear Generating Station has used the Electric Power Research Institute (EPRI) boiling water reactor (BWR) power shape monitoring system (PSMS) for approximately one cycle of power operation. During this time, the PSMS was used for both on-line core monitoring and off-line core simulation. The ability to obtain core power distributions in near real time and the ability to simulate anticipated core maneuvers has enabled site personnel to avoid difficult operational situations and to more easily detect problems when they occur. The PSMS is an integrated set of computer codes used to monitor and predict the nuclear and thermal-hydraulic behavior of a BWR code. The PSMS uses a three-dimensional, one-energy-group nodal model similar to the General Electric FLARE code to estimate the core power distribution. The PSMS contains four major modules: (1) data acquisition and storage module: collects plant data; (2) duty cycle approximation module; performs three-dimensional nuclear and thermal-hydraulic calculations; (3) model calibration module: performs all statistical evaluations; and (4) data display module: responsible for displaying PSMS data. The PSMS also has the ability to perform useful predictive calculations. These include: power, flow, or control rod searches; control rod pattern optimizations; and fast approximate predictions.

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Stellwag, Bernhard; Laendner, Alexander; Weiss, Steffen [AREVA NP GmbH, Erlangen (Germany); Huettner, Frank [Vattenfall Europe Nuclear Energy GmbH (Germany)

    2011-03-15

    Nineteen boiling water reactor (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 the 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. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Stellwag, B.; Laendner, A.; Weiss, S. [AREVA NP GmbH, Erlangen (Germany); Huettner, F. [Vattenfall Europe Nuclear Energy GmbH, Hamburg (Germany)

    2010-07-01

    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)

  10. 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)

  11. 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)

  12. Heat and humidity buildup under earmuff-type hearing protectors

    Directory of Open Access Journals (Sweden)

    Rickie R Davis

    2011-01-01

    Full Text Available A major barrier to effective wear of hearing protection is comfort. This study examined several comfort indicators in the earmuff-type hearing protectors. Twenty subjects wore hearing protectors instrumented with two different temperature/humidity measurement systems (Omega and iButton while walking a corridor for about 25 min. The instruments recorded the temperature and humidity every 10 s and their results were compared. In addition, skin surface pH was measured at the ear canal entrance before and after the task. Finally, the subject indicated earmuff comfort at the beginning and end of the session. Earmuff comfort decreased significantly over the course of the walking task. Ear canal pH became slightly less acidic, but the change was not statistically significant. The two temperature/humidity systems provided comparable results. Heat increased at about 0.3°F while humidity built up at about 0.5%/min. However, the study found some limitations on the instrumentation. The complexity of the electrical connections and equipment in the Omega probe system led to loss of three subject′s data. The iButton device was more robust, but provided only 256 gradations of temperature and relative humidity. Even with its limitations, the iButton device would be a valuable tool for field studies. The present study showed that the buildup of heat and humidity can be modeled using linear equations. The present study demonstrates that relatively inexpensive tools and a low-exertion task can provide important information about the under-earmuff environment, which can inform assumptions about comfort during use.

  13. A new approximating formula for calculating gamma-ray buildup factors in multilayer shields

    International Nuclear Information System (INIS)

    Assad, A.; Chiron, M.; Nimal, J.C.; Diop, C.M.; Ridoux, P.

    1999-01-01

    This study proposes a new approximating formula for calculating gamma-ray buildup factors in multilayer shields. The formula combines the buildup factors of single-layer shields with products and quotients. The feasibility of the formula for reproducing the buildup factors was tested by using point isotropic buildup factors calculated with the SN1D discrete ordinates code as reference data. The dose buildup factors of single-, double-, and multilayer shields composed of water, aluminum, iron, and lead were calculated for a spherical geometry in the energy range between 10 MeV and 40 keV and for total thicknesses of up to 30 mean free paths. The calculation of the buildup factors takes into account the bound electron effect of Compton scattering (incoherent scattering), the coherent scattering, the pair production, and the secondary sources of bremsstrahlung and fluorescence. The tests have shown that the approximating formula reproduces the reference data of double-layer shields very well for most cases. With the same parameters and with a new physical consideration that takes into account in a global way the degradation of the gamma-ray energy spectrum, the buildup factors of three- and five-layer shields were also very well reproduced

  14. Experimental study on the best control method for valve stem seals in BWR

    International Nuclear Information System (INIS)

    Tashiro, Hisao

    1982-01-01

    Valves as many as about 10,000 are used for a nuclear power plant, and their maintenance and management require much labor. Especially it is an important problem to find the countermeasures for preventing leak from valve stem seals. For this purpose, the leakless valves of bellows or diaphragm type are used for the important places regarding environmental control, but generally the valves with gland packings are used. The requirements for the gland packings of valves for nuclear power stations are excellent sealing capability, low stem friction, slight deterioration due to thermal cycles, no contamination of coolant, excellent endurance to radiation, no damage or corrosion of valve stems and easy insert and taking-out of packings. The asbestos packings with graphite treatment have been used so far, but recently expanded graphite packings are adopted. Their sealing capability is excellent, but stem friction is large. The accelerated experiment of 10,000 stem actions was carried out at 283 deg C and 72 kgf/cm 2 of BWR conditions, using the setup simulating valve stem sealing, for the purposes of clarifying the satisfactory combination of packings and the optimum managing method of stem sealing. The experimental setup, method and conditions, the packings tested and the results are reported. (Kako, I.)

  15. Recent studies on the welding of austenitic stainless steel piping for BWR service

    International Nuclear Information System (INIS)

    Childs, W.J.

    1986-01-01

    The incidence of intergranular stress corrosion cracking (IGSCC) in stainless steel piping in BWR power plants has led to the development of various countermeasures. Replacement of the susceptible Type 304 stainless steel with Type 316 nuclear grade stainless steel has been done by a number of plants. In order to minimize radiation exposure to welding personnel, automatic GTA welding has been used wherever possible when we make the field welds. Studies have shown that the residual stresses in the welded butt joints are affected by the welding process, weld joint design and welding procedures. A new weld joint design has been developed which minimizes the volume of deposited metal while providing adequate access for welding. It also minimizes axial and radial shrinkage and the resulting residual stresses. Other countermeasures, which have been used, include stress modifications such as induction heating stress improvement (IHSI) and last pass heat sink welding (LPHSW). It has been shown that these remedies must be process adjusted to account for the welding process employed. In some cases where UT cracking indication have been detected or where through wall cracking has occurred, weld surfacing has been used to extend life. A further approach to preventing IGSCC in the weld HAZ has been through improvement of the water chemistry by injecting hydrogen to reduce the oxygen level and by keeping the impurity level low

  16. Increase of Steam Moisture in the BWR-Facility KKP 1

    International Nuclear Information System (INIS)

    Noack, Volker

    2002-01-01

    Main steam moisture in a BWR facility is determined by steam quality at core outlet and efficiency of steam separators and steam dryers. Transport of water with steam is accompanied by transport of radionuclides out of RPV resulting in enhanced radiation level in the main steam system. A remarkable increase of main steam moisture started at KKP 1 in 1997. In the following years increase of steam outlet moisture started at lower and lower core mass flow rates. Dose rate in main steam system increased simultaneously. Core mass flow rate and thus thermal power had to be reduced during stretch out operation to keep the main steam moisture below the specified boundary of 0.2 %. This boundary also guarantees, that radiological exposure remains far below approved values. The increase of main steam moisture corresponds with the application of low leakage core loading. Low leakage core loading results in enhanced steam generation in the center and in reduced steam generation in the outer zones of the core. It can be shown, that the uneven steam generation in the core became stronger over the years. Therefore, steam quality at inlet of the outer steam separators was getting lower. This resulted in higher carry over of water in this steam separators and steam dryers, thus explaining the increasing main steam moisture. KKP 1 started in 2000 with spectral shift operation. As one should expect, this resulted in reduced steam moisture. It remains the question of steam moisture in case of stretch out operation. Countermeasures are briefly discussed. (authors)

  17. Buildup dynamics of dissipative soliton in an ultrafast fiber laser with net-normal dispersion.

    Science.gov (United States)

    Chen, Hong-Jie; Liu, Meng; Yao, Jian; Hu, Song; He, Jian-Bo; Luo, Ai-Ping; Xu, Wen-Cheng; Luo, Zhi-Chao

    2018-02-05

    Taking advantage of technology of spatio-temporal reconstruction and dispersive Fourier transform (DFT), we experimentally observed the buildup dynamics of dissipative soliton in an ultrafast fiber laser in the net-normal dispersion regime. The soliton buildup dynamics were analyzed in both the spectral and temporal domains. We firstly revealed that the appearing of the spectral sharp peaks with oscillation structures during the mode-locking transition is caused by the formation of structural dissipative soliton. The experimental results were explained by the numerical simulations. These findings would give some new insights into the dissipative soliton buildup dynamics in ultrafast fiber lasers.

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

    International Nuclear Information System (INIS)

    Castano, M.L.; Schaaf, B. van der; Roth, A.; Ohms, C.; Gavillet, D.; Dyck, S. van

    2004-01-01

    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 20 n/cm 2 , below the threshold fluence of 5 x 10 20 n/cm 2 (equivalent to ∼ 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 performed. Measurements of residual

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

    Energy Technology Data Exchange (ETDEWEB)

    Arganis J, C. R.

    2010-07-01

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

  20. 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)

  1. 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

  2. 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.)

  3. 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

  4. 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)

  5. Preliminary study on seismic proving test of BWR core internals

    International Nuclear Information System (INIS)

    Watabe, M.; Shibata, H.; Sato, H.; Kawakami, S.; Ohno, T.; Ichiki, T.; Sasaki, Y.

    1989-01-01

    A series of seismic proving tests for PWR and BWR facilities have been programmed and conducted. The seismic design of nuclear plants is based on criteria far more stringent than applied to non-nuclear installations. Seismic design involves the complex, advanced and sophisticated calculations. The validity of the techniques employed for these calculations is not susceptible to easy comprehension and acceptance by the population at large. It has become important to demonstrate the seismic strength and functional reliability of currently designed and existing nuclear power facilities. This paper deals with an experimental research project to demonstrate the seismic reliability proving tests for nuclear power plant facilities. The seismic proving test described in this report was conducted from 1986 to 1988 for the boiling water reactor core internals

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1986-01-01

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

  7. Blowdown heat transfer and transient boiling transition in BWR's

    International Nuclear Information System (INIS)

    Sozzi, G.L.; Burnette, G.W.

    1977-01-01

    Experimental results from the NRC/EPRI/GE BWR Blowdown Heat Transfer Program are evaluated in terms of bundle local heat transfer performance and in terms of cross-sectional average bundle thermal-hydraulic fluid conditions. The bundle heat transfer performance was generally found to be nucleate boiling below the two-phase mixture level interface with highly dispersed film boiling or steam cooling heat transfer above the interface. Comparisons are presented for predictions of boiling transition (BT) and post BT heat transfer performance during the blowdown phase of the LOCA experiments. These predictions utilize a drift flux void fraction model. The comparisons show very good agreement of both the onset of BT and the post BT heat transfer. 12 references

  8. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-11-15

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

  10. Summary report of seismic PSA of BWR model plant

    International Nuclear Information System (INIS)

    1999-05-01

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

  11. PWR and BWR spent fuel assembly gamma spectra measurements

    Science.gov (United States)

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

    2016-10-01

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

  12. Opacity Build-up in Impulsive Relativistic Sources

    Energy Technology Data Exchange (ETDEWEB)

    Granot, Jonathan; Cohen-Tanugi, Johann; Silva, Eduardo do Couto e

    2007-09-28

    Opacity effects in relativistic sources of high-energy gamma-rays, such as gamma-ray bursts (GRBs) or Blazars, can probe the Lorentz factor of the outflow as well as the distance of the emission site from the source, and thus help constrain the composition of the outflow (protons, pairs, magnetic field) and the emission mechanism. Most previous works consider the opacity in steady state. Here we study the effects of the time dependence of the opacity to pair production ({gamma}{gamma} {yields} e{sup +}e{sup -}) in an impulsive relativistic source, which may be relevant for the prompt gamma-ray emission in GRBs or flares in Blazars. We present a simple, yet rich, semi-analytic model for the time and energy dependence of the optical depth, {tau}{gamma}{gamma}, in which a thin spherical shell expands ultra-relativistically and emits isotropically in its own rest frame over a finite range of radii, R{sub 0} {le} R {le} R{sub 0}+{Delta}R. This is particularly relevant for GRB internal shocks. We find that in an impulsive source ({Delta}R {approx}< R{sub 0}), while the instantaneous spectrum (which is typically hard to measure due to poor photon statistics) has an exponential cutoff above the photon energy {var_epsilon}1(T) where t{gamma}{gamma}({var_epsilon}1) = 1, the time integrated spectrum (which is easier to measure) has a power-law high-energy tail above the photon energy {var_epsilon}1* {approx} {var_epsilon}1({Delta}T) where {Delta}T is the duration of the emission episode. Furthermore, photons with energies {var_epsilon} > {var_epsilon}1* are expected to arrive mainly near the onset of the spike in the light curve or flare, which corresponds to the short emission episode. This arises since in such impulsive sources it takes time to build-up the (target) photon field, and thus the optical depth {tau}{gamma}{gamma}({var_epsilon}) initially increases with time and {var_epsilon}1(T) correspondingly decreases with time, so that photons of energy {var_epsilon} > {var

  13. Opacity Build-up in Impulsive Relativistic Sources

    International Nuclear Information System (INIS)

    Granot, Jonathan; Cohen-Tanugi, Johann; Silva, Eduardo do Couto e

    2007-01-01

    Opacity effects in relativistic sources of high-energy gamma-rays, such as gamma-ray bursts (GRBs) or Blazars, can probe the Lorentz factor of the outflow as well as the distance of the emission site from the source, and thus help constrain the composition of the outflow (protons, pairs, magnetic field) and the emission mechanism. Most previous works consider the opacity in steady state. Here we study the effects of the time dependence of the opacity to pair production (γγ → e + e - ) in an impulsive relativistic source, which may be relevant for the prompt gamma-ray emission in GRBs or flares in Blazars. We present a simple, yet rich, semi-analytic model for the time and energy dependence of the optical depth, τγγ, in which a thin spherical shell expands ultra-relativistically and emits isotropically in its own rest frame over a finite range of radii, R 0 (le) R (le) R 0 +ΔR. This is particularly relevant for GRB internal shocks. We find that in an impulsive source (ΔR ∼ 0 ), while the instantaneous spectrum (which is typically hard to measure due to poor photon statistics) has an exponential cutoff above the photon energy (var e psilon)1(T) where tγγ((var e psilon)1) = 1, the time integrated spectrum (which is easier to measure) has a power-law high-energy tail above the photon energy (var e psilon)1* ∼ (var e psilon)1(ΔT) where ΔT is the duration of the emission episode. Furthermore, photons with energies (var e psilon) > (var e psilon)1* are expected to arrive mainly near the onset of the spike in the light curve or flare, which corresponds to the short emission episode. This arises since in such impulsive sources it takes time to build-up the (target) photon field, and thus the optical depth τγγ((var e psilon)) initially increases with time and (var e psilon)1(T) correspondingly decreases with time, so that photons of energy (var e psilon) > (var e psilon)1* are able to escape the source mainly very early on while (var e psilon)1(T) > (var

  14. Electron-cloud build-up in hadron machines

    International Nuclear Information System (INIS)

    Furman, M.A.

    2004-01-01

    ,17]. In this article we focus on the mechanisms of electron-cloud buildup and dissipation for hadronic beams, particularly those with very long, intense, bunches

  15. Buildup of Abiotic Oxygen and Ozone in Atmospheres of Temperate Terrestrial Exoplanets

    Science.gov (United States)

    Kleinboehl, Armin; Willacy, Karen; Friedson, Andrew James; Swain, Mark R.

    2015-12-01

    The last two decades have seen a rapid increase in the detection and characterization of exoplanets. A focus of future missions will be on the subset of transiting, terrestrial, temperate exoplanets as they are the strongest candidates to harbor life as we know it.An important bioindicator for life as we know it is the existence of significant amounts of oxygen, and its photochemical byproduct ozone, in the exoplanet’s atmosphere. However, abiotic processes also produce oxygen and ozone, and the amount of oxygen abiotically produced in an atmosphere will largely depend on other atmospheric parameters. Constraining this parameter space will be essential to avoid ‘false positive’ detections of life, that is the interpretation of oxygen or ozone as a bioindicator despite being produced abiotically.Based on 1D radiative-convective model calculations, Wordsworth and Pierrehumbert (ApJL, 2014) recently pointed out that the formation and buildup of abiotic oxygen on water-rich planets largely depends on the amount of non-condensable gases in the atmosphere. The amount of non-condensable gases determines whether an atmosphere will develop a 'cold-trap' (similar to the tropopause on Earth) that contains most of the water in the lower atmosphere and dries out the upper atmosphere. If water vapor is a major constituent of the atmosphere, this cold-trapping is inhibited, leading to a much moister upper atmosphere. Water vapor in the upper atmosphere is photolyzed due to the availability of hard UV radiation, yielding oxygen.We use a photochemical model coupled to a 1D radiative-convective climate model to self-consistently study this effect in atmospheres with N2, CO2 and H2O as the main constituents. These are typical constituents for secondary, oxidized atmospheres, and they can exist in a wide range of ratios. We calculate the amounts of abiotically produced oxygen and ozone and determine the vertical structure of temperature and constituent mixing ratios for various

  16. History matching of transient pressure build-up in a simulation model using adjoint method

    Energy Technology Data Exchange (ETDEWEB)

    Ajala, I.; Haekal, Rachmat; Ganzer, L. [Technische Univ. Clausthal, Clausthal-Zellerfeld (Germany); Almuallim, H. [Firmsoft Technologies, Inc., Calgary, AB (Canada); Schulze-Riegert, R. [SPT Group GmbH, Hamburg (Germany)

    2013-08-01

    The aim of this work is the efficient and computer-assisted history-matching of pressure build-up and pressure derivatives by small modification to reservoir rock properties on a grid by grid level. (orig.)

  17. The effect of stimulus context on the buildup to stream segregation

    Directory of Open Access Journals (Sweden)

    Jonathan eSussman-Fort

    2014-04-01

    Full Text Available Stream segregation is the process by which the auditory system disentangles the mixture of sound inputs into discrete sources that cohere across time. The length of time required for this to occur is termed the ‘buildup’ period. In the current study, we used the buildup period as an index of how quickly sounds are segregated into constituent parts. Specifically, we tested the hypothesis that stimulus context impacts the timing of the buildup and, therefore, affects when stream segregation is detected. To measure the timing of the buildup we recorded the Mismatch Negativity component (MMN of event-related brain potentials (ERPs, during passive listening, to determine when the streams were neurophysiologically segregated. In each condition, a pattern of repeating low (L and high (H tones (L-L-H was presented in trains of stimuli separated by silence, with the H tones forming a simple intensity oddball paradigm and the L tones serving as distractors. To determine the timing of the buildup, probe tones occurred in two positions of the trains, early (within the buildup period and late (past the buildup period. The context was manipulated by presenting roving versus non-roving frequencies across trains in two conditions. MMNs were elicited by intensity probe tones in the Non-Roving condition (early and late positions and the Roving condition (late position only indicating that neurophysiologic segregation occurred faster in the Non-Roving condition. This suggests a shorter buildup period when frequency was repeated from train to train. Overall, our results demonstrate that the dynamics of the environment influence the way in which the auditory system extracts regularities from the input. The results support the hypothesis that the buildup to segregation is highly dependent upon stimulus context and that the auditory system works to maintain a consistent representation of the environment when no new information suggests that reanalyzing the scene

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

    International Nuclear Information System (INIS)

    Sehgal, B.R.; Dinh, T.N.

    2002-12-01

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

  19. Analysis of the build-up of semi and non volatile organic compounds on urban roads.

    Science.gov (United States)

    Mahbub, Parvez; Ayoko, Godwin A; Goonetilleke, Ashantha; Egodawatta, Prasanna

    2011-04-01

    Vehicular traffic in urban areas may adversely affect urban water quality through the build-up of traffic generated semi and non volatile organic compounds (SVOCs and NVOCs) on road surfaces. The characterisation of the build-up processes is the key to developing mitigation measures for the removal of such pollutants from urban stormwater. An in-depth analysis of the build-up of SVOCs and NVOCs was undertaken in the Gold Coast region in Australia. Principal Component Analysis (PCA) and Multicriteria Decision tools such as PROMETHEE and GAIA were employed to understand the SVOC and NVOC build-up under combined traffic scenarios of low, moderate, and high traffic in different land uses. It was found that congestion in the commercial areas and use of lubricants and motor oils in the industrial areas were the main sources of SVOCs and NVOCs on urban roads, respectively. The contribution from residential areas to the build-up of such pollutants was hardly noticeable. It was also revealed through this investigation that the target SVOCs and NVOCs were mainly attached to particulate fractions of 75-300 μm whilst the redistribution of coarse fractions due to vehicle activity mainly occurred in the >300 μm size range. Lastly, under combined traffic scenario, moderate traffic with average daily traffic ranging from 2300 to 5900 and average congestion of 0.47 were found to dominate SVOC and NVOC build-up on roads. Copyright © 2011 Elsevier Ltd. All rights reserved.

  20. Determination of contamination-free build-up for 60Co

    International Nuclear Information System (INIS)

    Higgins, P.D.; Sibata, CH.; Paliwal, B.R.

    1985-01-01

    Experimental verification of the difference between absorbed dose in tissue and the collision fraction of kerma requires precise knowledge of the absorbed dose curve, particularly in the build-up and build-down regions. A simple method from direct measurement of contamination-free build-up for 60 Co, which should also be applicable for most of the photon energies commonly employed for treatment, is presented. It is shown that the contribution from air-scattered electrons to the surface dose may be removed by extrapolating measurements of build-up to zero field size. The remaining contribution to contamination from the collimators and other source-related hardware may be minimised by measuring these build-up curves sufficiently far from the source. These results were tested by measuring the build-up using a magnet to sweep scattered electrons from the primary photon beam and by measuring the surface dose in the limit of an evacuated beam path. The relative dose at zero depth in polystyrene was found to be approximately 8.9+-0.3% of the dose at the depth of maximum build-up. (author)

  1. Modelling ionising radiation induced defect generation in bipolar oxides with gated diodes

    International Nuclear Information System (INIS)

    Barnaby, H.J.; Cirba, C.; Schrimpf, R.D.; Kosier, St.; Fouillat, P.; Montagner, X.

    1999-01-01

    Radiation-induced oxide defects that degrade electrical characteristics of bipolar junction transistor (BJTs) can be measured with the use of gated diodes. The buildup of defects and their effect on device radiation response are modeled with computer simulation. (authors)

  2. Radiation

    International Nuclear Information System (INIS)

    Davidson, J.H.

    1986-01-01

    The basic facts about radiation are explained, along with some simple and natural ways of combating its ill-effects, based on ancient healing wisdom as well as the latest biochemical and technological research. Details are also given of the diet that saved thousands of lives in Nagasaki after the Atomic bomb attack. Special comment is made on the use of radiation for food processing. (U.K.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-15

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

  5. Lower pressurization to increase BWR electric power under thermal hydraulic criteria

    International Nuclear Information System (INIS)

    Kataoka, Kazuyoshi; Chuman, Kazuto; Mizumachi, Wataru; Yoshioka, Ritsuo; Mori, Michitsugu; Horie, Akira; Machida, Yuzo

    1995-01-01

    Electric power output versus core size is one of the factors that determine the electricity generation costs of BWRs. The power output is roughly calculated from the thermal power of the BWR core and the thermal efficiency of the BWR turbine system. The thermal power is restricted by the reactor's thermal hydraulic criteria such as the maximum linear heat generation rate, the minimum critical power ratio, the pressure drop in the core and the feedwater temperature at the BWR inlet. The combination of a system pressure of approximately 5.5 MPa and a feedwater temperature of approximately 439 K offers the maximum electric power output for a BWR with 9 x 9 fuel bundles. The amount of electric power generated is about 9% more than that generated by a conventional BWR under the thermal hydraulic criteria. The electric power output increases as the system pressure and the feedwater temperature are decreased from the current design of 7.3 MPa and 488 K, respectively, because the increased critical power of the fuel bundles compensates for the lower thermal efficiency. (author)

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

    International Nuclear Information System (INIS)

    Powers, J.; Yonezawa, H.; Aoyagi, Y.; Kataoka, K.

    2015-09-01

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

  7. Characterizing heavy metal build-up on urban road surfaces: Implication for stormwater reuse

    Energy Technology Data Exchange (ETDEWEB)

    Liu, An [Research Centre of Environmental Engineering and Management, Graduate School at Shenzhen, Tsinghua University, 518055 Shenzhen (China); Cooperative Research and Education Centre for Environmental Technology, Kyoto University–Tsinghua University, 518055 Shenzhen (China); Liu, Liang; Li, Dunzhu [Research Centre of Environmental Engineering and Management, Graduate School at Shenzhen, Tsinghua University, 518055 Shenzhen (China); Guan, Yuntao, E-mail: guanyt@tsinghua.edu.cn [Research Centre of Environmental Engineering and Management, Graduate School at Shenzhen, Tsinghua University, 518055 Shenzhen (China); School of Environment, Tsinghua University, Beijing 100084 (China)

    2015-05-15

    Stormwater reuse is increasingly popular in the worldwide. In terms of urban road stormwater, it commonly contains toxic pollutants such as heavy metals, which could undermine the reuse safety. The research study investigated heavy metal build-up characteristics on urban roads in a typical megacity of South China. The research outcomes show the high variability in heavy metal build-up loads among different urban road sites. The degree of traffic congestion and road surface roughness was found to exert a more significant influence on heavy metal build-up rather than traffic volume. Due to relatively higher heavy metal loads, stormwater from roads with more congested traffic conditions or rougher surfaces might be suitable for low-water-quality required activities while the stormwater from by-pass road sections could be appropriate for relatively high-water-quality required purposes since the stormwater could be relatively less polluted. Based on the research outcomes, a decision-making process for heavy metals based urban road stormwater reuse was proposed. The new finding highlights the importance to undertaking a “fit-for-purpose” road stormwater reuse strategy. Additionally, the research results can also contribute to enhancing stormwater reuse safety. - Highlights: • Heavy metal (HM) build-up varies with traffic and road surface conditions. • Traffic congestion and surface roughness exert a higher impact on HM build-up. • A “fit-for-purpose” strategy could suit urban road stormwater reuse.

  8. Characterizing heavy metal build-up on urban road surfaces: Implication for stormwater reuse

    International Nuclear Information System (INIS)

    Liu, An; Liu, Liang; Li, Dunzhu; Guan, Yuntao

    2015-01-01

    Stormwater reuse is increasingly popular in the worldwide. In terms of urban road stormwater, it commonly contains toxic pollutants such as heavy metals, which could undermine the reuse safety. The research study investigated heavy metal build-up characteristics on urban roads in a typical megacity of South China. The research outcomes show the high variability in heavy metal build-up loads among different urban road sites. The degree of traffic congestion and road surface roughness was found to exert a more significant influence on heavy metal build-up rather than traffic volume. Due to relatively higher heavy metal loads, stormwater from roads with more congested traffic conditions or rougher surfaces might be suitable for low-water-quality required activities while the stormwater from by-pass road sections could be appropriate for relatively high-water-quality required purposes since the stormwater could be relatively less polluted. Based on the research outcomes, a decision-making process for heavy metals based urban road stormwater reuse was proposed. The new finding highlights the importance to undertaking a “fit-for-purpose” road stormwater reuse strategy. Additionally, the research results can also contribute to enhancing stormwater reuse safety. - Highlights: • Heavy metal (HM) build-up varies with traffic and road surface conditions. • Traffic congestion and surface roughness exert a higher impact on HM build-up. • A “fit-for-purpose” strategy could suit urban road stormwater reuse

  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/6, and documents the positions of the Nuclear Regulatory Commission based on the B ampersand W Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. This document Volume 3, contains the Bases for Sections 3.4--3.10 of the improved STS

  10. Impact of metals recycling on a Swedish BWR decommissioning project

    International Nuclear Information System (INIS)

    Larsson, Arne; Lidar, Per; Hedin, Gunnar; Bergh, Niklas

    2014-01-01

    Decommissioning of nuclear power plants generates large volumes of radioactive or potentially contaminated metals. By proper management of the waste streams significant amounts can be free released and recycled either directly or after decontamination and melting. A significant part of the required work should be performed early in the process to make the project run smoothly without costly surprises and delays. A large portion of the clearance activities can be performed on-site. This on-site work should focus on the so called low-risk for contamination material. Other material can be decontaminated and released on site if schedule and the available facility areas so allow. It should be noted that the on-site decontamination and clearance activities can be a significant bottle neck for a decommissioning project. The availability of and access to a specialized metals recycling facility is an asset for a decommissioning project. This paper will describe the forecasted positive impact of a well-structured metals characterisation, categorisation and clearance process for a BWR plant decommissioning project. The paper is based on recent studies, performed projects and recent in-house development. (authors)

  11. Incorporation of an BWR evaporator concentrate in polyethylene

    International Nuclear Information System (INIS)

    Moriyama, N.; Dojiro, S.; Matsuzuru, H.

    1982-01-01

    The adaptability of polyethylene solidification method to an evaporator concentrate produced in BWR has been examined. The polyethylene product incorporating sodium sulfate as a main constituent of the concentrate was prepared using a batch-type melter. The product was tested with respect to homogeneity, density, mechanical property, water resistance and leachability, in order to evaluate the safety of the product. The product can incorporate the waste (sodium sulfate) up to 70 wt.% but it may be recommended through tests of water resistance and leachability that the maximum waste content in the product is 50 wt.%. The product with 50 wt.% waste has the density of 1.28 g/cm 3 , the compressive strength of 213 kgf/cm 2 , enough toughness, low leachability (diffusion coefficient of sodium is 10/sup -7/-10/sup ///31 /sup 6/ cm/sup 2//day) and excellent water resistance. This solidification method offers about 5 times larger volume reduction effect than a cement solidification does. 11 refs

  12. Technology developments for Japanese BWR MOX fuel utilization

    International Nuclear Information System (INIS)

    Oguma, M.; Mochida, T.; Nomata, T.; Asahi, K.

    1997-01-01

    The Long-Term Program for Research, Development and Utilization of Nuclear Energy established by the Atomic Energy Commission of Japan asserts that Japan will promote systematic utilization of MOX fuel in LWRs. Based on this Japanese nuclear energy policy, we have been pushing development of MOX fuel technology aimed at future full scale utilization of this fuel in BWRs. In this paper, the main R and D topics are described from three subject areas, MOX core and fuel design, MOX fuel irradiation behaviour, and MOX fuel fabrication technology. For the first area, we explain the compatibility of MOX fuel with UO 2 core, the feasibility of the full MOX core, and the adaptability of MOX design methods based on a mock-up criticality experiment. In the second, we outline the Tsuruga MOX irradiation program and the DOMO program, and suggest that MOX fuel behaviour is comparable to ordinary BWR UO 2 fuel behaviour. In the third, we examine the development of a fully automated MOX bundle assembling apparatus and its features. (author). 14 refs, 11 figs, 3 tabs

  13. Reactor power control device in BWR power plant

    International Nuclear Information System (INIS)

    Kurosawa, Tsuneo.

    1997-01-01

    The present invention provides a device for controlling reactor power based on a start-up/shut down program in a BWR type reactor, as well as for detecting deviation, if occurs, of the power from the start-up/shut down program, to control a recycling flow rate control system or control rod drive mechanisms. Namely, a power instruction section successively executes the start-up/shut down program and controls the coolant recycling system and the control rod driving mechanisms to control the power. A current state monitoring and calculation section receives a process amount, calculates parameters showing the plant state, compares/monitors them with predetermined values, detecting the deviation, if occurs, of the plant state from the start-up/shut down program, and prevents output of a power increase control signal which leads to power increase. A forecasting and monitoring/calculation section forecasts and calculates the plant state when not yet executed steps of the start-up/shut down program are performed, stops the execution of the start-up/shut down program in the next step in a case of forecasting that the results of the calculation will deviate from the start-up/shut down program. (I.S.)

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

    International Nuclear Information System (INIS)

    Hernandez S, A.

    2004-01-01

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

  15. GOTHIC MODEL OF BWR SECONDARY CONTAINMENT DRAWDOWN ANALYSES

    International Nuclear Information System (INIS)

    Hansen, P.N.

    2004-01-01

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

  16. Kuosheng BWR/6 containment safety analysis with gothic code

    International Nuclear Information System (INIS)

    Lin Ansheng; Wang Jongrong; Yuann Rueyyng; Shih Chunkuan

    2011-01-01

    Kuosheng Nuclear Power Plant in Taiwan is a GE-designed twin-unit BWR/6 plant, each unit rated at 2894 MWt. In this study, we presented the calculated results of the containment pressure and temperature responses after the main steam line break accident, which is the design basis for the containment system. During the simulation, a power of SPU range (105.1%) was used and a model of the Mark III type containment was built using the containment thermal-hydraulic program GOTHIC. The simulation consists of short and long-term responses. The drywell pressure and temperature responses which display the maximum values in the early state of the LOCA were investigated in the short-term response; the primary containment pressure and temperature responses in the long-term response. The blowdown flow was provided by FSAR and used as boundary conditions in the short-term model; in the long-term model, the blowdown flow was calculated using a GOTHIC built-in homogeneous equilibrium model. In the long-term analysis, a simplifier RPV model was employed to calculate the blowdown flow. Finally, the calculated results, similar to the FSAR results, indicate the GOTHIC code has the capability to simulate the pressure/temperature response of Mark III containment to the main steam line break LOCA. (author)

  17. Controlling device for BWR type nuclear power plants

    International Nuclear Information System (INIS)

    Sekimizu, Koichi

    1987-01-01

    Purpose: To enable to operate BWR type nuclear power plants while entirely drawing control rods upon stationary operation near the rated power. Method: Upon stationary operation of nuclear power plants near the rated power, an excess reactivity of nuclear fuels is controlled by entirely withdrawing control rods, and varying the feedwater enthalpy thereby changing the void amount. For this purpose, a feedwater heater is additionally disposed between a high pressure feedwater heater of a feedwater pipeway and a nuclear reactor pressure vessel, in which steams used for rising the temperature of the feedwater are introduced to the high temperature side of the high pressure feedwater heater and used again for the heating of the feedwater. In this case, the feedwater enthalpy can be increased approximately to the saturation enthalpy by adjusting such that about 30 % of the main steams are introduced to the high temperature side of the feedwater heater by a steam flow rate control valve, which enables to control the stationary operation without using control rods. (Horiuchi, T.)

  18. Calibration of the TVO spent BWR reference fuel assembly

    International Nuclear Information System (INIS)

    Tarvainen, M.; Baecklin, A.; Haakanson, A.

    1992-02-01

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

  19. Natural heat transfer augmentation in passive advanced BWR plants

    International Nuclear Information System (INIS)

    Gamble, R.E.; Peterson, P.F.; Greif, R.

    2001-01-01

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

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

    Directory of Open Access Journals (Sweden)

    Javier Ortiz-Villafuerte

    2013-01-01

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

  1. Experimental and numerical investigations of BWR fuel bundle inlet flow

    International Nuclear Information System (INIS)

    Hoashi, E; Morooka, S; Ishitori, T; Komita, H; Endo, T; Honda, H; Yamamoto, T; Kato, T; Kawamura, S

    2009-01-01

    We have been studying the mechanism of the flow pattern near the fuel bundle inlet of BWR using both flow visualization test and computational fluid dynamics (CFD) simulation. In the visualization test, both single- and multi-bundle test sections were used. The former test section includes only a corner orifice facing two support beams and the latter simulates 16 bundles surrounded by four beams. An observation window is set on the side of the walls imitating the support beams upstream of the orifices in both test sections. In the CFD simulation, as well as the visualization test, the single-bundle model is composed of one bundle with a corner orifice and the multi-bundle model is a 1/4 cut of the test section that includes 4 bundles with the following four orifices: a corner orifice facing the corner of the two neighboring support beams, a center orifice at the opposite side from the corner orifice, and two side orifices. Twin-vortices were observed just upstream of the corner orifice in the multi-bundle test as well as the single-bundle test. A single-vortex and a vortex filament were observed at the side orifice inlet and no vortex was observed at the center orifice. These flow patterns were also predicted in the CFD simulation using Reynolds Stress Model as a turbulent model and the results were in good agreement with the test results mentioned above. (author)

  2. Calculation device for fuel power history in BWR type reactors

    International Nuclear Information System (INIS)

    Sakagami, Masaharu.

    1980-01-01

    Purpose: To enable calculations for power history and various variants of power change in the power history of fuels in a BWR type reactor or the like. Constitution: The outputs of the process computation for the nuclear reactor by a process computer are stored and the reactor core power distribution is judged from the calculated values for the reactor core power distribution based on the stored data. Data such as for thermal power, core flow rate, control rod position and power distribution are recorded where the changes in the power distribution exceed a predetermined amount, and data such as for thermal power and core flow rate are recorded where the changes are within the level of the predetermined amount, as effective data excluding unnecessary data. Accordingly, the recorded data are taken out as required and the fuel power history and the various variants in the fuel power are calculated and determined in a calculation device for fuel power history and variants for fuel power fluctuation. (Furukawa, Y.)

  3. A New Method for Early Anomaly Detection of BWR Instabilities

    International Nuclear Information System (INIS)

    Ivanov, K.N.

    2005-01-01

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

  4. 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/6, and documents the positions of the Nuclear Regulatory Commission based on the B ampersand W Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. This document Volume 1, contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS

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

    Energy Technology Data Exchange (ETDEWEB)

    Ivanov, K. N.

    2005-11-27

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

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

    International Nuclear Information System (INIS)

    Ortiz S, J.J.

    1998-01-01

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

  7. Some studies for aseismic design of BWR power station

    International Nuclear Information System (INIS)

    Kasai, Hiroaki; Tanaka, Kihachiro; Takayanagi, Masaaki; Moriyama, Takeo; Kashimura, Yoshisada

    1978-01-01

    Nuclear power stations are constructed regarding the safety as the most important matter, and in Japan where earthquakes occur frequently, attention must be paid to the aseismatic characteristics in particular. BWR power generation facilities have high aseismatic performances on the basis of the experiences of construction and the design technology over many years. The tests and researches concerning the aseismatic characteristics are roughly divided into those of establishing reasonable design conditions, of improving the accuracy of the method of analyzing earthquake response, and of proving the maintenance of functions. Artificial earthquake waves are made by the method of correcting recorded waves or the method of sine wave synthesis. For the analysis of building-ground systems, the modeling method using finite element analysis has been studied. As for the researches on the aseismatic characteristics of core equipments, the vibration experiment on fuel assemblies and the proving test on the insertion of control rods were carried out. Especially in the proving test on the insertion of control rods, it was found that the stop of a nuclear reactor was able to be carried out without fail at the time of the earthquake far exceeding the maximum fuel amplitude estimated for the plants in Japan. Such results of researches are accumulated continuously and reflected to the design. (Kako, I.)

  8. New innovative electrocoagulation (EC) treatment technology for BWR colloidal iron utilizing the seeding and filtration electronically (SAFETTM) system

    International Nuclear Information System (INIS)

    Denton, Mark S.; Bostick, William D.

    2007-01-01

    The presence of iron (iron oxide from carbon steel piping) buildup in Boiling Water Reactor (BWR) circuits and wastewaters is decades old. In, perhaps the last decade, the advent of precoatless filters for condensate blow down has compounded this problem due to the lack of a solid substrate (e.g., Powdex resin pre-coat) to help drop the iron out of solution. The presence and buildup of this iron in condensate phase separators (CPS) further confounds the problem when the tank is decanted back to the plant. Iron carryover here is unavoidable without further treatment steps. The form of iron in these tanks, which partially settles and is pumped to a de-waterable high integrity container (HIC), is particularly difficult and time consuming to de-water (low shear strength, high water content). The addition upstream from the condensate phase separator (CPS) of chemicals, such as polymers, to carry out the iron, only produces an iron form even more difficult to filter and de-water (even less shear strength, higher water content, and a gel/slime consistency). Typical, untreated colloidal material contains both sub-micron particles up to, let's say 100 micron. It is believed that the sub-micron particles penetrate filters, or sheet filters, thus plugging the pores for what should have been the successful filtration of the larger micron particles. Like BWR iron wastewaters, fuel pools/storage basins (especially in the decon. phase) often contain colloids which make clarity and the resulting visibility nearly impossible. Likewise, miscellaneous, often high conductivity, waste streams at various plants contain such colloids, iron, salts (sometimes seawater intrusion and referred to as Salt Water Collection Tanks), dirt/clay, surfactants, waxes, chelants, etc. Such waste streams are not ideally suited for standard dead-end (cartridges) or cross-flow filtration (UF/RO) followed even by demineralizers. Filter and bed plugging are almost assured. The key to solving these dilemmas

  9. Evaluation of the cracking by stress corrosion in nuclear reactor environments type BWR

    International Nuclear Information System (INIS)

    Arganis J, C. R.

    2010-01-01

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

  10. Paleochannel and beach-bar palimpsest topography as initial substrate for coralligenous buildups offshore Venice, Italy.

    Science.gov (United States)

    Tosi, Luigi; Zecchin, Massimo; Franchi, Fulvio; Bergamasco, Andrea; Da Lio, Cristina; Baradello, Luca; Mazzoli, Claudio; Montagna, Paolo; Taviani, Marco; Tagliapietra, Davide; Carol, Eleonora; Franceschini, Gianluca; Giovanardi, Otello; Donnici, Sandra

    2017-05-02

    We provide a model for the genesis of Holocene coralligenous buildups occurring in the northwestern Adriatic Sea offshore Venice at 17-24 m depth. High-resolution geophysical surveys and underwater SCUBA diving reconnaissance revealed meandering shaped morphologies underneath bio-concretionned rocky buildups. These morphologies are inferred to have been inherited from Pleistocene fluvial systems reactivated as tidal channels during the post- Last Glacial Maximum transgression, when the study area was a lagoon protected by a sandy barrier. The lithification of the sandy fossil channel-levee systems is estimated to have occurred at ca. 7 cal. ka BP, likely due to the interaction between marine and less saline fluids related to onshore freshwater discharge at sea through a sealed water-table. The carbonate-cemented sandy layers served as nucleus for subsequent coralligenous buildups growth.

  11. Change in surface SP caused by pressure buildup observed at the Nigorikawa geothermal area; Nigorikawa chiiki ni okeru atsuryoku buildup ji no shizen den`i henka

    Energy Technology Data Exchange (ETDEWEB)

    Yasukawa, K.; Yano, Y.; Matsushima, N.; Ishido, T. [Geological Survey of Japan, Tsukuba (Japan); Takahashi, M.; Suzuki, I.; Aoyama, K.; Kuwano, T.

    1996-10-01

    To examine the effect of change of subsurface flow system on the surface SP (self potential), SP measurements were carried out before and after the pressure buildup and drawdown during the periodic inspection at Nigorikawa area. Relation between the SP distribution and the observed data was also examined by 2-D numerical simulation. Tendency was found that the SP increased gradually with the production near the production well, decreased during the pressure buildup, and increased again during the drawdown. There were some points having the reverse tendency in the surrounding area. Behavior during the pressure buildup and drawdown was not clear. The resistivity near the ground surface was low ranging between 2 and 5 ohm/m within the Nigorikawa basin. The variation of SP was not so large when compared with the measuring error. The SP profiles on the secondary section passing in the center of caldera at the production stop and at one week after the production start were well corresponded with the profiles under natural conditions which were reproduces using the 2-D model. It was considered that the SP profile before the production stop was affected by the production. 12 refs., 6 figs., 1 tab.

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

    International Nuclear Information System (INIS)

    Karlberg, G.; Goddard, C.; Fitzpatrick, S.

    1994-02-01

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

  13. Design guideline to prevent the pipe rupture by radiolysis gases in BWR steam piping

    International Nuclear Information System (INIS)

    Inagaki, T.; Miyagawa, M.; Ota, T.; Sato, T.; Sakata, K.

    2009-01-01

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

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

    International Nuclear Information System (INIS)

    Andrade, G.G. de.

    1973-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

  17. A method of the sensitivity analysis of build-up and decay of actinides

    International Nuclear Information System (INIS)

    Mitani, Hiroshi; Koyama, Kinji; Kuroi, Hideo

    1977-07-01

    To make sensitivity analysis of build-up and decay of actinides, mathematical methods related to this problem have been investigated in detail. Application of time-dependent perturbation technique and Bateman method to sensitivity analysis is mainly studied. For the purpose, a basic equation and its adjoint equation for build-up and decay of actinides are systematically solved by introducing Laplace and modified Laplace transforms and their convolution theorems. Then, the mathematical method of sensitivity analyses is formulated by the above technique; its physical significance is also discussed. Finally, application of eigenvalue-method is investigated. Sensitivity coefficients can be directly calculated by this method. (auth.)

  18. The Effect of Sloshing on a Tank Pressure Build-up Unit

    OpenAIRE

    Banne, Håvard Bolstad

    2017-01-01

    This thesis work has aimed to identify how sloshing will affect a liquefied natural gas (LNG) fuel tank. The physical nature of LNG means it needs to be kept cooled and pressurized in order to remain in a liquid state. By implementing a pressure build-up unit (PBU) it is possible to pressurize the tank vaporizing the tank’s contents, for the vapour then to return to tank in a loop, building pressure in the process. A tank pressure build-up unit has been built in the laboratory ...

  19. Basic principle of constant q/sub a/ current build-up in tokamaks

    International Nuclear Information System (INIS)

    Kikuchi, M.

    1985-05-01

    An analytic expression is derived such that the current profile shape is kept constant during the current build-up phase in tokamaks. The required conductivity profile is parametrized by two externally controllable parameters, I/sub p/ and a/sub p/ in the case of the Gaussian current profile. It is shown that a Gaussian current profile can be maintained for a realistically broad conductivity profile by using the constant q/sub a/ current build-up method even under the condition of a high I/sub p/

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

    International Nuclear Information System (INIS)

    Khan, H.J.; Cheng, H.S.; Rohatgi, U.S.

    1992-01-01

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

  1. Behavior to the fracture of an AISI 304 stainless steel sensitized in BWR reactor conditions (288 degrees Centigrade and 80 Kg/cm2)

    International Nuclear Information System (INIS)

    Hernandez C, R.; Diaz S, A.; Garcia R, R.; Aguilar T, A.; Gachuz M, M.; Arganis J, C.; Merino C, J.

    1999-01-01

    It is a knew fact that ductility of a lot of structural alloys can be deteriorated by the environment effect which are exposed, and that their consequent embrittlement can put in doubt the safety of their functioning; such is the case of austenitic stainless steels used in internal components of the BWR type reactors which not only is subjected to the effect combined of the aggressive environment which surround it (pressure, temperature, corrosion potential, conductivity medium, local state of efforts, etc.), but also to the action of present neutron radiation, manifesting microstructural changes which are reflected in the augmentation of its susceptibility to the intergranular cracking, phenomena generally known as IASCC ''Irradiation Assisted Stress Corrosion Cracking''. Once appeared the cracking in the material, the useful life of a component is limited by the rapidity to growth of these cracking, making necessary evaluations which can to predict its behavior, therefore the present work shows the preliminary results for determining the behavior to the fracture of an AISI 304 stainless steel sensitized, in a dynamic recirculation circuit which allows to simulate the operation conditions of a BWR reactor (288 Centigrade and 80 kg/cm 2 ). (Author)

  2. Complete BWR--EM LOCA analysis using the WRAP--EM system

    International Nuclear Information System (INIS)

    Beckmeyer, R.R.; Gregory, M.V.; Buckner, M.R.

    1979-01-01

    The Water Reactor Analysis Package, Evaluation Model (WRAP--EM), provides a complete analysis of postulated loss-of-coolant accidents (LOCA's) in light--water nuclear power reactors. The system is being developed at the Savannah River Laboratory (SRL) for use by the Nuclear Regulatory Commission (NRC) to interpret and evaluate reactor vendor, evaluation model (EM) analyses. The initial version of the WRAP--EM system for analysis of boiling water reactors (BWR's) is operational. To demonstrate the complete capability of the WRAP--BWR--EM system, a LOCA analysis has been performed for the Hope Creek Plant

  3. A linear reactivity model approximation to predict BWR fuel assembly burnup

    Energy Technology Data Exchange (ETDEWEB)

    Delfin L, A.; Alonso V, G. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Valle G, E. del [IPN, ESFM, 07738 Mexico D.F. (Mexico)]. e-mail: adl@nuclear.inin.mx

    2007-07-01

    A second order nodal method was derived for the one and a half group version of neutron diffusion theory in X-Y geometry. Based on the linear reactivity model, two reactivity relations are developed to treat fresh and depleted loading batches as a way to account for the spatial variation of moderator properties due to the substantial fraction of steam void present as well as the effect of the control rod positioning. Based on these ideas a program was developed to predict fuel batch burnup in a BWR operating cycle and fuel power sharing. The method was used to test two consecutive BWR operating fuel cycles. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  6. Calculation of BWR [Boiling Water Reactor] limit cycle amplitude using Galerkin's method

    International Nuclear Information System (INIS)

    Damiano, B.; March-Leuba, J.; Euler, J.A.

    1990-01-01

    This paper describes the application of Galerkin's method to estimate the amplitude of boiling water reactor (BWR) limit cycle oscillations. It will be shown that Galerkin's method can be applied to a model of BWR dynamics consisting of the point kinetics equations and the LAPUR generated feedback transfer function to calculate the time history of small amplitude limit cycles. This allows results from the linear frequency domain code LAPUR to be used to calculate nonlinear time domain information. 2 refs., 2 figs., 1 tab

  7. Accident sequence analysis for a BWR [Boiling Water Reactor] during low power and shutdown operations

    International Nuclear Information System (INIS)

    Whitehead, D.W.; Hake, T.M.

    1990-01-01

    Most previous Probabilistic Risk Assessments have excluded consideration of accidents initiated in low power and shutdown modes of operation. A study of the risk associated with operation in low power and shutdown is being performed at Sandia National Laboratories for a US Boiling Water Reactor (BWR). This paper describes the proposed methodology for the analysis of the risk associated with the operation of a BWR during low power and shutdown modes and presents preliminary information resulting from the application of the methodology. 2 refs., 2 tabs

  8. Kinetic analyses on startup and shutdown chemistry of BWR plant

    International Nuclear Information System (INIS)

    Domae, Masafumi; Fujiwara, Kazutoshi; Inagaki, Hiromitsu

    2012-09-01

    During startup and shutdown of Boiling Water Reactor (BWR) plants, temperature and dissolved oxygen (DO) concentration of reactor water change in a wide range. The changes result in variation of conductivity and pH of the reactor water. It has been speculated that the water chemistry change is due to dissolution of the oxides on fuel claddings and structural materials. However, detailed mechanism is not known. In the present paper, trend of recent water chemistry in several BWR plants during startup and shutdown is presented. Conductivity and pH are convenient indication of coolant purity. We tried to clarify the mechanism of the change in the conductivity and the pH value during startup and shutdown, based on the water chemistry data measured. In the water chemistry data, change in chromate concentration and Ni 2+ concentration is rather large. It is assumed that change in the chromate concentration and the Ni 2+ concentration results in the time variation of the conductivity and the pH value. It is reasonable to consider that the increase in the chromate concentration and the Ni 2+ concentration is ascribed to dissolution of Cr oxides and Ni oxides, respectively. A model of dissolution of the Cr oxides and the Ni oxides is proposed. A concept of finite inventory of the Cr oxides and the Ni oxides in the coolant system is introduced. The model is as follows. Chromate is generated by oxidation of the Cr oxides and the Cr dissolution rate depends on the DO concentration. The dissolution rate of chromate is in proportion to DO concentration, the inventory of Cr and difference between solubility limit and the chromate concentration. On the other hand, Ni 2+ is formed by dissolution of the Ni oxides, and DO is not necessary in this process. The dissolution rate of Ni 2+ is in proportion to the inventory of Ni and difference between solubility limit and the Ni 2+ concentration. Coolant is continuously purified, and the chromate concentration and the Ni 2+ concentration

  9. Analysis of BWR high burnup fuel in LOCA conditions

    International Nuclear Information System (INIS)

    Garcia Sedano, Pablo; Dey Navarro, Jose Manuel; Gallego Cabezon, Ines; Orive Moreno, Raul

    2004-01-01

    High Burnup Fuel Behaviour has been growing in importance since middle 80's when pellet microstructure changes (rim effect) and cladding oxidation rates increase were observed. Later on, Cadarache reactivity tests revealed cladding integrity failures below safety limits. These phenomena, occurred at high burnup, stressed the necessity of having a wide experimental data base that would allow to dispose non-extrapolated data of material properties submitted to higher burnups than 40000 MWd/TM and data of new materials at the same time. One of the objectives of the EPRI's Fuel Reliability Program is to establish the bases for the licensing of nuclear fuel to burnup levels beyond the current licensed value of 62 GWd/MTU rod average burnup. The technical bases to support those high burnup levels are being developed. One of the licensing points of concern is the behaviour of the high burnup fuel in LOCA conditions. To respond to this concern a series of LOCA experiments are being performed at Argonne National Laboratory using fuel rods from Limerick NPP at 57 GWd/TM and H.B. Robinson at 67 GWd/MTU. When the ANL tests have been finished, a conservative Peak Cladding Temperature/ Equivalent Cladding Reacted (PCT/ECR) limit will be determine from the residual ductility tests to be applied to the high burnup fuel. This makes necessary to determine the behaviour of the high burnup fuel in LOCA conditions and to determine the available safety margin. In licensing LOCA calculations, corresponding to present core designs and future core designs, the calculated PCT and ECR values as a function of the fuel burnup could be used to determine the relative severity of LOCA for the high burnup fuel. This report presents the LOCA analyses performed by IBERDROLA (Spanish utility), using results from the Cofrentes NPP (BWR-6) LOCA evaluations. (authors)

  10. Timing criteria for supplemental BWR emergency response equipment

    International Nuclear Information System (INIS)

    Bickel, John H.

    2015-01-01

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

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

    International Nuclear Information System (INIS)

    Damsohn, M.

    2011-01-01

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

  12. Effect of heat build-up on carbon emissions in chimato compost piles ...

    African Journals Online (AJOL)

    A study was conducted to determine impacts of heat build-up of chimato compost piles TD0, TD20, TD40, TD50, TD60, TD80 and TD100, made by blending maize stalks with 0, 20, 40, 50, 60, 80 and 100% Tithonia diversifolia, respectively, on carbon losses and emissions during composting. Compost piles temperatures ...

  13. "Testing during Study Insulates against the Buildup of Proactive Interference": Correction

    Science.gov (United States)

    Szpunar, Karl K.; McDermott, Kathleen B.; Roedigger, Henry L., III

    2009-01-01

    Reports an error in "Testing during study insulates against the buildup of proactive interference" by Karl K. Szpunar, Kathleen B. McDermott and Henry L. Roediger III ("Journal of Experimental Psychology: Learning, Memory, and Cognition," 2008[Nov], Vol 34[6], 1392-1399). Incorrect figures were printed due to an error in the…

  14. Space charge build-up in XLPE-cable with temperature gradient

    DEFF Research Database (Denmark)

    Holbøll, Joachim; Henriksen, Mogens; Hjerrild, Jesper

    2000-01-01

    Space charge build-up in standard XLPE insulated AC cables has been studied under varying temperature and field conditions. The cables were triple-extruded with the inner semicon on a solid aluminum conductor, 5.5mm XLPE-insulation and an outer semicon. The cables were stressed up to 15kV/mm DC...

  15. Effect of finite sample dimensions and total scatter acceptance angle on the gamma ray buildup factor

    International Nuclear Information System (INIS)

    Singh, Sukhpal; Kumar, Ashok; Singh, Charanjeet; Thind, Kulwant Singh; Mudahar, Gurmel S.

    2008-01-01

    The simultaneous variation of gamma ray buildup factors with absorber thickness (up to 6.5 mfp) and total scatter acceptance angle (which is the sum of incidence and exit beam divergence) in the media of high volume flyash concrete and water was studied experimentally using a point isotropic 137 Cs source

  16. Effects of Fallow Genealogical Cycles on the Build-up of Nutrients in ...

    African Journals Online (AJOL)

    The study examined the effect of fallow generational cycles on the buildup of nutrients in the soil. Fallow sequence of 1st, 2nd, 3rd, 4th and 5th generations were studied. The quadrat approach of sampling was employed to collect soil samples (surface and subsurface) from five plots of 10m x 10m across the five fallow ...

  17. Investigating the build-up of precedence effect using reflection masking

    DEFF Research Database (Denmark)

    Hartcher-O'Brien, Jessica; Buchholz, Jörg

    2006-01-01

    The auditory processing level involved in the build-up of precedence [Freyman et al., J. Acoust. Soc. Am. 90, 874–884 (1991)] has been investigated here by employing reflection masked threshold (RMT) techniques. Given that RMT techniques are generally assumed to address lower levels of the auditory...

  18. Energy absorption buildup factors for thermoluminescent dosimetric materials and their tissue equivalence

    DEFF Research Database (Denmark)

    Manohara, S.R.; Hanagodimath, S.M.; Gerward, Leif

    2010-01-01

    Gamma ray energy-absorption buildup factors were computed using the five-parameter geometric progression (G-P) fitting formula for seven thermoluminescent dosimetric (TLD) materials in the energy range 0.015-15 MeV, and for penetration depths up to 40 mfp (mean free path). The generated energy-absorption...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  1. Repair bond strength of dual-cured resin composite core buildup materials.

    Science.gov (United States)

    El-Deeb, Heba A; Ghalab, Radwa M; Elsayed Akah, Mai M; Mobarak, Enas H

    2016-03-01

    The reparability of dual-cured resin composite core buildup materials using a light-cured one following one week or three months storage, prior to repair was evaluated. Two different dual-cured resin composites; Cosmecore™ DC automix and Clearfil™ DC automix core buildup materials and a light-cured nanofilled resin composite; Filtek™ Z350 XT were used. Substrate specimens were prepared (n = 12/each substrate material) and stored in artificial saliva at 37 °C either for one week or three months. Afterward, all specimens were ground flat, etched using Scotchbond™ phosphoric acid etchant and received Single Bond Universal adhesive system according to the manufacturers' instructions. The light-cured nanofilled resin composite (Filtek™ Z350 XT) was used as a repair material buildup. To determine the cohesive strength of each solid substrate material, additional specimens from each core material (n = 12) were prepared and stored for the same periods. Five sticks (0.8 ± 0.01 mm(2)) were obtained from each specimen (30 sticks/group) for microtensile bond strength (μTBS) testing. Modes of failure were also determined. Two-way ANOVA revealed a significant effect for the core materials but not for the storage periods or their interaction. After one week, dual-cured resin composite core buildup materials (Cosmecore™ DC and Clearfil™ DC) achieved significantly higher repair μTBS than the light-cured nanofilled resin composite (Filtek™ Z350 XT). However, Clearfil™ DC revealed the highest value, then Cosmecore™ DC and Filtek™ Z350 XT, following storage for 3-month. Repair strength values recovered 64-86% of the cohesive strengths of solid substrate materials. The predominant mode of failure was the mixed type. Dual-cured resin composite core buildup materials revealed acceptable repair bond strength values even after 3-month storage.

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

    International Nuclear Information System (INIS)

    Jonas, T.; Gross, R.; Logback, F.; Josyula, R.

    1995-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  4. Update on materials performance and electrochemistry in hydrogen water chemistry at Dresden-2 BWR

    International Nuclear Information System (INIS)

    Indig, M.E.; Weber, J.E.; Davis, R.B.; Gordon, B.M.

    1985-01-01

    Previous studies performed in 1982 indicated that if sufficient hydrogen was injected into the Dresden-2 BWR, IGSCC of sensitized austenitic stainless steel was mitigated. The present series of experiments were aimed at verification of the above finding, determining how much time off hydrogen water chemistry (HWC) could be tolerated and how HWC affected pre-existing cracks

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-01-29

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

  6. International comparison calculations for a BWR lattice with adjacent gadolinium pins

    International Nuclear Information System (INIS)

    Maeder, C.; Wydler, P.

    1984-09-01

    The results of burnup calculations for a simplified BWR fuel element with two adjacent gadolinium rods are presented and discussed. Ten complete solutions were contributed by Denmark, France, Italy (3), Japan (3), Switzerland and the UK. Partial results obtained from Poland and the USA are included in an Appendix. (Auth.)

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

    International Nuclear Information System (INIS)

    Hodge, S.A.; Kress, T.S.; Cleveland, J.C.; Petek, M.

    1992-01-01

    This paper briefly describes the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to evaluate the effectiveness and feasibility of current and proposed strategies for BWR severe accident management. These results are described in detail in the just-released report Identification and Assessment of BWR In-Vessel Severe Accident Mitigation Strategies, NUREG/CR-5869, which comprises three categories of findings. First, an assessment of the current status of accident management strategies for the mitigation of in-vessel events for BWR severe accident sequences is combined with a review of the BWR Owners' Group Emergency Procedure Guidelines (EPGs) to determine the extent to which they currently address the characteristic events of an unmitigated severe accident. Second, where considered necessary, new candidate accident management strategies are proposed for mitigation of the late-phase (after core damage has occurred) events. Finally, two of the four candidate strategies identified by this effort are assessed in detail. These are (1) preparation of a boron solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and (2) containment flooding to maintain the core debris within the reactor vessel if the injection systems cannot be restored

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

    Directory of Open Access Journals (Sweden)

    Abdul Waris

    2008-03-01

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

  9. On the fast estimation of transit times application to BWR simulated data

    International Nuclear Information System (INIS)

    Antonopoulos-Domis, M.; Marseguerra, M.; Padovani, E.

    1996-01-01

    Real time estimators of transit times are proposed. BWR noise is simulated including a global component due to rod vibration. The time obtained form the simulation is used to investigate the robustness and noise immunity of the estimators. It is found that, in presence of a coincident (global) signal, the cross-correlation function is the worst estimator. (authors)

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

    International Nuclear Information System (INIS)

    Cano Aguado, M.; Perlado Martin, J.M.; Minguez Torres, E.

    1977-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-15

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

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

    International Nuclear Information System (INIS)

    Powers, J.; Aoyagi, Y.; Kataoka, K.; Thomas, S.; Mookhoek, B.

    2015-09-01

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

  13. LOSP-initiated event tree analysis for BWR

    International Nuclear Information System (INIS)

    Watanabe, Norio; Kondo, Masaaki; Uno, Kiyotaka; Chigusa, Takeshi; Harami, Taikan

    1989-03-01

    As a preliminary study of 'Japanese Model Plant PSA', a LOSP (loss of off-site power)-initiated Event Tree Analysis for a Japanese typical BWR was carried out solely based on the open documents such as 'Safety Analysis Report'. The objectives of this analysis are as follows; - to delineate core-melt accident sequences initiated by LOSP, - to evaluate the importance of core-melt accident sequences in terms of occurrence frequency, and - to develop a foundation of plant information and analytical procedures for efficiently performing further 'Japanese Model Plant PSA'. This report describes the procedure and results of the LOSP-initiated Event Tree Analysis. In this analysis, two types of event trees, Functional Event Tree and Systemic Event Tree, were developed to delineate core-melt accident sequences and to quantify their frequencies. Front-line System Event Tree was prepared as well to provide core-melt sequence delineation for accident progression analysis of Level 2 PSA which will be followed in a future. Applying U.S. operational experience data such as component failure rates and a LOSP frequency, we obtained the following results; - The total frequency of core-melt accident sequences initiated by LOSP is estimated at 5 x 10 -4 per reactor-year. - The dominant sequences are 'Loss of Decay Heat Removal' and 'Loss of Emergency Electric Power Supply', which account for more than 90% of the total core-melt frequency. In this analysis, a higher value of 0.13/R·Y was used for the LOSP frequency than experiences in Japan and any recovery action was not considered. In fact, however, there has been no experience of LOSP event in Japanese nuclear power plants so far and it is also expected that offsite power and/or PCS would be recovered before core melt. Considering Japanese operating experience and recovery factors will reduce the total core-melt frequency to less than 10 -6 per reactor-year. (J.P.N.)

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

    International Nuclear Information System (INIS)

    Martin del Campo M, C.; Carmona H, R.; Oropeza C, I.P.

    2006-01-01

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

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

    International Nuclear Information System (INIS)

    Garcia V, M.A.

    2006-01-01

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

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

    International Nuclear Information System (INIS)

    Itjeu Karliana; Sumijanto; Dhandhang Purwadi, M.

    2008-01-01

    The needs of fresh water increased by rapid population growth and industrials expansion, but these demands can not be prepared naturally. Following this case, seawater desalination becomes the primer option which can fulfill the need through the nuclear desalination technology. The coupled nuclear power reactor enables to supply thermal energy for auxiliary equipment and pumps operation. The utilization study of power reactor type BWR coupled with desalination process has been performed. The goal of study is to obtain characteristic data of desalted water specification which desalination system coupling with nuclear power plant produced energy for desalination process. The study is carried out by browsing data and information, and comprehensive review of thermal energy correlation between NPP with desalination process installation. According to reviewing are found that the thermal energy and electric power utilization from the nuclear power reactor are enable to remove the seawater to produce desalted water and also to operate auxiliary equipments. The assessment results is VK-300 reactor prototype, BWR type 250 MW(e) power are cogeneration unit can supplied hot steam temperature 285 °C to the extraction turbine to empower 150 MW electric power, and a part of hot steam 130 °C is use to operate desalination process and remind heat is distribute to the municipal and offices at that region. The coupled of VK-300 reactor power type BWR with desalination installation of MED type enable to produce desalted water with high quality distillate. Based on the economic calculation that the VK-300 reactor power of BWR type produced water distillate capacity is 300.000 m 3 /hour with cost US$ 0.58/m 3 . The coupling VK-300 reactor power type BWR with MED desalination plant is competitive economically. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Casado Sanchez, C.; Rubio Oviedo, P.

    2014-07-01

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

  18. THALES, Thermohydraulic LOCA Analysis of BWR and PWR

    International Nuclear Information System (INIS)

    ABE, Kiyoharu

    1990-01-01

    reactor coolant system, combustible gas burning, atmosphere- structure heat transfer, ventilation, containment spray cooling, etc. After the molten core penetrates the reactor bottom head, steam generation, concrete disintegration and noncondensable gas generation are calculated in the reactor cavity or the pedestal. 2 - Method of solution: Each of the THALES member codes first establishes the steady state conditions after reading input data. Then iterative time-dependent calculation is continued, taking account of various phenomena and events and their interactions which will occur in the course of a postulated severe accident. The transient calculations are iterated by the physical times specified by input. Generally the RCS thermal hydraulic analysis with the THALES-PM or THALES-BM code is first carried out and its results are transferred to the following containment analysis with the THALES-CV code. Then both results are transferred to a code for analyzing fission product release and transport behavior. Automatic data transfer is possible in the case the JAERI's ART code is used for fission product behavior analysis. In overall thermal hydraulic analysis, a new method is adopted aiming at sufficiently accurate estimation of mixture levels in the reactor coolant system and the containment in a reasonable computer time. The heat transfer calculation in the core is carried out based on the backward method. 3 - Restrictions on the complexity of the problem: Restrictions relating to storage allocation are: (1) Maximum number of radial regions in the core : 10; (2) Maximum number of axial increments in the fuel rods : 50; (3) Maximum number of loops in the PWR primary system : 4; (4) Maximum number of volumes in the PWR primary system : 11; (5) Number of BWR recirculation loops: 2 (fixed); (6) Number of volumes in the BWR reactor coolant system : 7 (fixed); (7) Maximum number of compartments in the containment : 10. There is another restriction, which relates to time step

  19. Influence of standard RF coil materials on surface and buildup dose from a 6 MV photon beam in magnetic field.

    Science.gov (United States)

    Ghila, A; Fallone, B G; Rathee, S

    2016-11-01

    Magnetic resonance guided teletherapy systems aspire to image the patient concurrently with the radiation delivery. Thus, the radiofrequency (RF) coils used for magnetic resonance imaging, placed on or close to patient skin and in close proximity to the treatment volume, would be irradiated leading to modifications of radiation dose to the skin and in the buildup region. The purpose of this work is to measure and assess these dose modifications due to standard off-the-shelf RF coil materials. A typical surface coil was approximated as layered sheets of polycarbonate, copper tape, and Teflon to emulate the base, conductor, and cover, respectively. A separate investigation used additional coil materials, such as copper pipe, plastic coil housing, a typical coil padding material, and a thin copper conductor. The materials were placed in the path of a 6 MV photon beam at various distances from polystyrene phantoms in which the surface and buildup doses were measured. The experiments were performed on a clinical Varian linac with no magnetic field and with a 0.21 T electromagnet producing a magnetic field parallel to the beam central axis. The authors repeated similar experiments in the presence of a 0.22 T magnetic field oriented perpendicular to the beam central axis using an earlier linac-MR prototype, with a biplanar permanent magnet. The radiation detectors used for the measurements were two different parallel plate ion chambers and GAFChromic films. A typical open beam surface dose of 20% (relative to open beam D max ) was increased to 63% by the coil padding material and to >74% by all other materials when placed in direct contact with the phantom, irrespective of magnetic field presence or orientation. Without a magnetic field, the surface dose decreased as the test materials were moved away from the phantom surface toward the radiation source, reaching between 30% and 40% at 10 cm gap between the phantom and the test materials. In the presence of the transverse

  20. Behavior to the fracture of an AISI 304 stainless steel sensitized in BWR reactor conditions (288 degrees Centigrade and 80 Kg/cm{sup 2}); Comportamiento a la fractura de un acero inoxidable AISI 304 sensibilizado en condiciones de reactor BWR (288 grados Centigrados y 80 Kg/cm{sup 2})

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-07-01

    It is a knew fact that ductility of a lot of structural alloys can be deteriorated by the environment effect which are exposed, and that their consequent embrittlement can put in doubt the safety of their functioning; such is the case of austenitic stainless steels used in internal components of the BWR type reactors which not only is subjected to the effect combined of the aggressive environment which surround it (pressure, temperature, corrosion potential, conductivity medium, local state of efforts, etc.), but also to the action of present neutron radiation, manifesting microstructural changes which are reflected in the augmentation of its susceptibility to the intergranular cracking, phenomena generally known as IASCC ''Irradiation Assisted Stress Corrosion Cracking''. Once appeared the cracking in the material, the useful life of a component is limited by the rapidity to growth of these cracking, making necessary evaluations which can to predict its behavior, therefore the present work shows the preliminary results for determining the behavior to the fracture of an AISI 304 stainless steel sensitized, in a dynamic recirculation circuit which allows to simulate the operation conditions of a BWR reactor (288 Centigrade and 80 kg/cm{sup 2}). (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Garcia V, M.A

    2006-07-01

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

  2. BWR fuel performance under advanced water chemistry conditions – a delicate journey towards zero fuel failures – a review

    International Nuclear Information System (INIS)

    Hettiarachchi, S.

    2015-01-01

    Boiling Water Reactors (BWRs) have undergone a variety of chemistry evolutions over the past few decades as a result of the need to control stress corrosion cracking of reactor internals, radiation fields and personnel exposure. Some of the advanced chemistry changes include hydrogen addition, zinc addition, iron reduction using better filtration technologies, and more recently noble metal chemical addition to many of the modern day operating BWRs. These water chemistry evolutions have resulted in changes in the crud distribution on fuel cladding material, Co-60 levels and the Rod oxide thickness (ROXI) measurements using the conventional eddy current techniques. A limited number of Post-Irradiation Examinations (PIE) of fuel rods that exhibited elevated oxide thickness using eddy current techniques showed that the actual oxide thickness by metallography is much lower. The difference in these observations is attributed to the changing magnetic properties of the crud affecting the rod oxide thickness measurement by the eddy current technique. This paper will review and summarize the BWR fuel cladding performance under these advanced and improved water chemistry conditions and how these changes have affected the goal to reach zero fuel failures. The paper will also provide a brief summary of some of the results of hot cell PIE, results of crud composition evaluation, crud spallation, oxide thickness measurements, hydrogen content in the cladding and some fuel failure observations. (author) Key Words: Boiling Water Reactor, Fuel Performance, Hydrogen Addition, Zinc Addition, Noble Metal Chemical Addition, Zero Leakers

  3. Axial profiles of burned and fraction of holes for calculations of criticality with credit for BWR fuel burning

    International Nuclear Information System (INIS)

    Casado Sanchez, C.; Rubio Oviedo, P.

    2014-01-01

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

  4. Qualification of helium measurement system for detection of fuel failures in a BWR

    Science.gov (United States)

    Larsson, I.; Sihver, L.; Loner, H.; Grundin, A.; Helmersson, J.-O.; Ledergerber, G.

    2014-05-01

    There are several methods for surveillance of fuel integrity during the operation of a boiling water reactor (BWR). The detection of fuel failures is usually performed by analysis of grab samples of off-gas and coolant activities, where a measured increased level of ionizing radiation serves as an indication of new failure or degradation of an already existing one. At some nuclear power plants the detection of fuel failures is performed by on-line nuclide specific measurements of the released fission gases in the off-gas system. However, it can be difficult to distinguish primary fuel failures from degradation of already existing failures. In this paper, a helium measuring system installed in connection to a nuclide specific measuring system to support detection of fuel failures and separate primary fuel failures from secondary ones is presented. Helium measurements provide valuable additional information to measurements of the gamma emitting fission gases for detection of primary fuel failures, since helium is used as a fill gas in the fuel rods during fabrication. The ability to detect fuel failures using helium measurements was studied by injection of helium into the feed water systems at the Forsmark nuclear power plant (NPP) in Sweden and at the nuclear power plant Leibstadt (KKL) in Switzerland. In addition, the influence of an off-gas delay line on the helium measurements was examined at KKL by injecting helium into the off-gas system. By using different injection rates, several types of fuel failures with different helium release rates were simulated. From these measurements, it was confirmed that the helium released by a failed fuel can be detected. It was also shown that the helium measurements for the detection of fuel failures should be performed at a sampling point located before any delay system. Hence, these studies showed that helium measurements can be useful to support detection of fuel failures. However, not all fuel failures which occurred at

  5. Gamma-ray energy absorption and exposure buildup factor studies in some human tissues with endometriosis

    Energy Technology Data Exchange (ETDEWEB)

    Kurudirek, Murat, E-mail: mkurudirek@gmail.co [Faculty of Science, Department of Physics, Ataturk University, 25240 Erzurum (Turkey); Dogan, Bekir [Faculty of Science, Department of Physics, Ataturk University, 25240 Erzurum (Turkey); Ingec, Metin [Faculty of Medicine, Department of Obstetrics and Gynecology, Ataturk University, 25240 Erzurum (Turkey); Ekinci, Neslihan; Ozdemir, Yueksel [Faculty of Science, Department of Physics, Ataturk University, 25240 Erzurum (Turkey)

    2011-02-15

    Human tissues with endometriosis have been analyzed in terms of energy absorption (EABF) and exposure (EBF) buildup factors using the five-parameter geometric progression (G-P) fitting formula in the energy region 0.015-15 MeV up to a penetration depth of 40 mfp (mean free path). Chemical compositions of the tissue samples were determined using a wavelength dispersive X-ray fluorescence spectrometer (WDXRFS). Possible conclusions were drawn due to significant variations in EABF and EBF for the selected tissues when photon energy, penetration depth and chemical composition changed. Buildup factors so obtained may be of use when the method of choice for treatment of endometriosis is radiotherapy.

  6. Simulations of the electron cloud buildups and suppressions in Tevatron and main injector

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Xiaolong; Ostiguy, Jean-Francois; Chou, Weiren; /Fermilab

    2007-06-01

    To assess the effects of the electron cloud on Main Injector intensity upgrades, simulations of the cloud buildup were carried out using POSINST and compared with ECLOUD. Results indicate that even assuming an optimistic 1.3 maximum secondary electron yield, the electron cloud remains a serious concern for the planned future operational mode with 500 bunches, 3e11 proton per bunch. Electron cloud buildup can be mitigated in various ways. We consider a plausible scenario involving solenoids in straight section and a single clearing strip electrode (like SNEG in Tevatron) held at a potential of 500V. Simulations with parameters corresponding to Tevatron and Main Injector operating conditions at locations where special electron cloud detectors have been installed have been carried out and are in satisfactory agreement with preliminary measurements.

  7. Computational fluid dynamics modeling of two-phase flow in a BWR fuel assembly

    International Nuclear Information System (INIS)

    Andrey Ioilev; Maskhud Samigulin; Vasily Ustinenko; Simon Lo; Adrian Tentner

    2005-01-01

    Full text of publication follows: The goal of this project is to develop an advanced Computational Fluid Dynamics (CFD) computer code (CFD-BWR) that allows the detailed analysis of the two-phase flow and heat transfer phenomena in a Boiling Water Reactor (BWR) fuel bundle under various operating conditions. This code will include more fundamental physical models than the current generation of sub-channel codes and advanced numerical algorithms for improved computational accuracy, robustness, and speed. It is highly desirable to understand the detailed two-phase flow phenomena inside a BWR fuel bundle. These phenomena include coolant phase changes and multiple flow regimes which directly influence the coolant interaction with fuel assembly and, ultimately, the reactor performance. Traditionally, the best analysis tools for the analysis of two-phase flow phenomena inside the BWR fuel assembly have been the sub-channel codes. However, the resolution of these codes is still too coarse for analyzing the detailed intra-assembly flow patterns, such as flow around a spacer element. Recent progress in Computational Fluid Dynamics (CFD), coupled with the rapidly increasing computational power of massively parallel computers, shows promising potential for the fine-mesh, detailed simulation of fuel assembly two-phase flow phenomena. However, the phenomenological models available in the commercial CFD programs are not as advanced as those currently being used in the sub-channel codes used in the nuclear industry. In particular, there are no models currently available which are able to reliably predict the nature of the flow regimes, and use the appropriate sub-models for those flow regimes. The CFD-BWR code is being developed as a customized module built on the foundation of the commercial CFD Code STAR-CD which provides general two-phase flow modeling capabilities. The paper describes the model development strategy which has been adopted by the development team for the

  8. Secondary side TSP deposit buildup: lab test investigation focused on electrokinetic considerations

    Energy Technology Data Exchange (ETDEWEB)

    Barale, M.; Guillodo, M.; Foucault, M., E-mail: Morgan.Barale@areva.com [AREVA NP SAS, Technical Centre, Le Creusot (France); Ryckelynck, N.; Clinard, M-H.; Chahma, F.; Brun, C. [AREVA NP SAS, Chemistry and Radiochemistry Group, Paris (France); Corredera, G. [Electricite de France, Centre d' Expertise et d' Inspection dans les domaines de la Realisation et de l' Exploitation, Saint-Denis (France)

    2010-07-01

    Deposit buildup which caused the clogging of the 'foils' of the upper tube-support-plates (TSP) inside a PWR steam generator of French NPPs in 2006 presents certain similarities with deposits observed in lab tests performed in secondary coolant chemistry at the Technical Centre of AREVA NP in 2002. The mechanism of TSP clogging seems not to present obvious phenomenological links with the fouling of the free span of SG since deposits buildup is quite uniform and is currently related to a surface boiling effect due to the surface heat flux. A specific mechanism could account for TSP clogging. In particular, electrokinetic effects were investigated by EDF-CEIDRE and AREVA NP SAS in the framework of a lab test program started in 2007. The electrokinetic approach is to consider that the coupling of local hydrodynamic and surface electrochemistry could lead to the formation of a very localized and heterogeneous deposit at the leading edge between both TSP and SG tubing material. Electrokinetic effects can lead to the oxidation and/or the precipitation of ferrous ions and to a variation of the electrokinetic potential which can produce strong attraction of iron oxide colloids. These electrokinetic effects are dependent of the T/H and local hydrodynamic conditions and surface electrochemistry explaining. The objective of this EDF-AREVA lab test program is to investigate the role of secondary chemistry coolant (pH, DH, N{sub 2}H{sub 4}, amine, redox) and of the nature of materials (SS, Ni base alloy) on deposit buildup. Properties of oxide surface and zeta potential of oxidized metallic materials have been also determined at temperature to understand their potential contribution on mechanism of TSP clogging in secondary side chemistry coolant. In this paper, a set of specific experiments carried out in this frame have been presented and discussed, paying particular attention to the effects of electrokinetic considerations and surface charges at oxide

  9. Secondary side TSP deposit buildup: lab test investigation focused on electrokinetic considerations

    International Nuclear Information System (INIS)

    Barale, M.; Guillodo, M.; Foucault, M.; Ryckelynck, N.; Clinard, M-H.; Chahma, F.; Brun, C.; Corredera, G.

    2010-01-01

    Deposit buildup which caused the clogging of the 'foils' of the upper tube-support-plates (TSP) inside a PWR steam generator of French NPPs in 2006 presents certain similarities with deposits observed in lab tests performed in secondary coolant chemistry at the Technical Centre of AREVA NP in 2002. The mechanism of TSP clogging seems not to present obvious phenomenological links with the fouling of the free span of SG since deposits buildup is quite uniform and is currently related to a surface boiling effect due to the surface heat flux. A specific mechanism could account for TSP clogging. In particular, electrokinetic effects were investigated by EDF-CEIDRE and AREVA NP SAS in the framework of a lab test program started in 2007. The electrokinetic approach is to consider that the coupling of local hydrodynamic and surface electrochemistry could lead to the formation of a very localized and heterogeneous deposit at the leading edge between both TSP and SG tubing material. Electrokinetic effects can lead to the oxidation and/or the precipitation of ferrous ions and to a variation of the electrokinetic potential which can produce strong attraction of iron oxide colloids. These electrokinetic effects are dependent of the T/H and local hydrodynamic conditions and surface electrochemistry explaining. The objective of this EDF-AREVA lab test program is to investigate the role of secondary chemistry coolant (pH, DH, N 2 H 4 , amine, redox) and of the nature of materials (SS, Ni base alloy) on deposit buildup. Properties of oxide surface and zeta potential of oxidized metallic materials have been also determined at temperature to understand their potential contribution on mechanism of TSP clogging in secondary side chemistry coolant. In this paper, a set of specific experiments carried out in this frame have been presented and discussed, paying particular attention to the effects of electrokinetic considerations and surface charges at oxide-solution interfaces

  10. Build-up dynamics of heavy metals deposited on impermeable urban surfaces.

    Science.gov (United States)

    Wicke, D; Cochrane, T A; O'Sullivan, A

    2012-12-30

    A method using thin boards (3 cm thick, 0.56 m(2)) comprising different paving materials typically used in urban environments (2 asphalt types and concrete) was employed to specifically investigate air-borne deposition dynamics of TSS, zinc, copper and lead. Boards were exposed at an urban car park near vehicular traffic to determine the rate of contaminant build-up over a 13-day dry period. Concentration profiles from simulated rainfall wash-off were used to determine contaminant yields at different antecedent dry days. Maximum contaminant yields after 13 days of exposure were 2.7 kg ha(-1) for TSS, 35 g ha(-1) zinc, 2.3 g ha(-1) copper and 0.4 g ha(-1) lead. Accumulation of all contaminants increased over the first week and levelled off thereafter, supporting theoretical assumptions that contaminant accumulation on impervious surfaces asymptotically approaches a maximum. Comparison of different surface types showed approximately four times higher zinc concentrations in runoff from asphalt surfaces and two times higher TSS concentrations in runoff from concrete, which is attributed to different physical and chemical compositions of the pavement types. Contaminant build-up and wash-off behaviours were modelled using exponential and saturation functions commonly applied in the US EPA's Stormwater Management Model (SWMM) showing good correlation between measured and modelled concentrations. Maximum build-up, half-saturation time, build-up rate constants and wash-off coefficients, necessary for stormwater contaminant modelling, were determined for the four contaminants studied. These parameters are required to model contaminant concentrations in urban runoff assisting in stormwater management decisions. Copyright © 2012 Elsevier Ltd. All rights reserved.

  11. Step-by-step build-up of covalent poly(ethylene oxide) nanogel films.

    Science.gov (United States)

    Zahouani, S; Hurman, L; De Giorgi, M; Vigier-Carrière, C; Boulmedais, F; Senger, B; Frisch, B; Schaaf, P; Lavalle, P; Jierry, L

    2017-11-30

    Hydrogels based on poly(ethylene glycol) (PEG) are commonly used for studies related to cell fate and tissue engineering. Here we present a new covalent layer-by-layer build-up process leading to PEG coatings of nanometer size called "nanogel films". Compared to macroscopic hydrogels, such nanogels should provide a fine control over the structure and the thickness of the coating. Alternated deposition of bifunctional and tetra functional PEG molecules reacting through thiol/maleimide click chemistry is evaluated by quartz crystal microbalance. We first study parameters influencing the build-up process of such coatings and demonstrate the importance of (i) the nature of the first deposited layer, (ii) the PEG concentrations and (iii) the length of the PEG chains that appears to be the most significant parameter influencing film growth. The build-up process can be extended to a large variety of substrates like SiO 2 or polymers by using an appropriate anchoring layer. Covalent functionalization of these nanogel films by proteins or enzymes is suited by modifying the biomolecules with thiol or maleimide groups and immobilizing them during the build-up process. Activity of the embedded enzymes can be maintained. Moreover ligands like biotin can be incorporated into the film and recognition by streptavidin can be modulated by playing with the number of PEG layers covering biotin. Compared to well-known PEG hydrogels, these new coatings are promising as they allow to (i) build thin nanometric coatings, (ii) finely control the amount of deposited PEG and (iii) organize the position of the embedded biomolecules inside the film layers.

  12. Oxide charge build-up in MOS-structures by tunnel-injection and irradiation

    International Nuclear Information System (INIS)

    Knoll, M.

    1983-08-01

    Tunnel injection and irradiation experiments on MOS structures are performed in order to compare the effects of the two procedures. A model of oxide charge build-up by tunnel injection is established which yields figures of the electron capture cross section, the detrapping cross section and the impact ionizing coefficient. The interface charge is neutral for midgap band bending and the interface states have acceptor type above and donor type below midgap. (orig.) [de

  13. On spatial stabilization of dielectric barrier discharge microfilaments by residual heat build-up in air

    Science.gov (United States)

    Ráhel, Jozef; Szalay, Zsolt; Čech, Jan; Morávek, Tomás

    2016-04-01

    Microfilaments of dielectric barrier discharge are known for their multiple re-appearance at the same spot on dielectrics. This effect of localized re-appearance is driven by residual excited species and ions, surface charge deposited on the dielectric and the local temperature build-up resulting in the local increase of reduced electric field E/ΔN. To assess the magnitude of the latter, the breakdown voltage vs. temperature up to 180 °C was carefully measured at coplanar DBD and used as an input into the numerical simulation of heat build-up by the train of discharge pulses. An average reduction of breakdown voltage was found to be 20 V/K. The model predicted a quasi-stable microfilament temperature into which the thermal build-up rapidly converges. Its magnitude agreed well with the reported rotational temperature of similar electrode configuration. The impact of quasi-stable temperature on microfilament formation dynamics is further discussed. Contribution to the Topical Issue "Recent Breakthroughs in Microplasma Science and Technology", edited by Kurt Becker, Jose Lopez, David Staack, Klaus-Dieter Weltmann and Wei Dong Zhu.

  14. Neutron Buildup Factors Calculation for Support Vector Regression Application in Shielding Analysis

    International Nuclear Information System (INIS)

    Duckic, P.; Matijevic, M.; Grgic, D.

    2016-01-01

    In this paper initial set of data for neutron buildup factors determination using Support Vector Regression (SVR) method is prepared. The performance of SVR technique strongly depends on the quality of information used for model training. Thus it is very important to provide representable data to the SVR. SVR is a supervised type of learning so it demands data in the input/output form. In the case of neutron buildup factors estimation, the input parameters are the incident neutron energy, shielding thickness and shielding material and the output parameter is the neutron buildup factor value. So far the initial sets of data for different shielding configurations have been obtained using SCALE4.4 sequence SAS3. However, this results were obtained using group constants, thus the incident neutron energy was determined as the average value for each energy group. Obtained this way, the data provided to the SVR are fewer and therefore insufficient. More valuable information is obtained using SCALE6.2beta5 sequence MAVRIC which can perform calculations for the explicit incident neutron energy, which leads to greater maneuvering possibilities when active learning measures are employed, and consequently improves the quality of the developed SVR model.(author).

  15. BWR Spent Nuclear Fuel Interfacial Bonding Efficiency Study

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-04-30

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Trianti, Nuri, E-mail: nuri.trianti@gmail.com, E-mail: szaki@fi.itba.c.id; Su' ud, Zaki, E-mail: nuri.trianti@gmail.com, E-mail: szaki@fi.itba.c.id; Arif, Idam, E-mail: nuri.trianti@gmail.com, E-mail: szaki@fi.itba.c.id [Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Science, Bandung Institute of Technology (Ganesha 10 Bandung, Indonesia) (Indonesia); Riyana, EkaSapta [Nuclear Energy Regulatory Agency (BAPETEN) (Indonesia)

    2014-09-30

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-01

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

  19. Study on the Development of New BWR Core Analysis Scheme Based on the Continuous Energy Monte Carlo Burn-up Calculation Method

    OpenAIRE

    東條, 匡志; tojo, masashi

    2007-01-01

    In this study, a BWR core calculation method is developed. The continuous energy Monte Carlo burn-up calculation code is newly applied to BWR assembly calculations of production level. The applicability of the present new calculation method is verified through the tracking-calculation of commercial BWR.The mechanism and quantitative effects of the error propagations, the spatial discretization and of the temperature distribution in fuel pellet on the Monte Carlo burn-up calculations are clari...

  20. Layer by layer buildup of polysaccharide films: physical chemistry and cellular adhesion aspects.

    Science.gov (United States)

    Richert, Ludovic; Lavalle, Philippe; Payan, Elisabeth; Shu, Xiao Zheng; Prestwich, Glenn D; Stoltz, Jean-François; Schaaf, Pierre; Voegel, Jean-Claude; Picart, Catherine

    2004-01-20

    The formation ofpolysaccharide films based on the alternate deposition of chitosan (CHI) and hyaluronan (HA) was investigated by several techniques. The multilayer buildup takes place in two stages: during the first stage, the surface is covered by isolated islets that grow and coalesce as the construction goes on. After several deposition steps, a continuous film is formed and the second stage of the buildup process takes place. The whole process is characterized by an exponential increase of the mass and thickness of the film with the number of deposition steps. This exponential growth mechanism is related to the ability of the polycation to diffuse "in" and "out" of the whole film at each deposition step. Using confocal laser microscopy and fluorescently labeled CHI, we show that such a diffusion behavior, already observed with poly(L-lysine) as a polycation, is also found with CHI, a polycation presenting a large persistence length. We also analyze the effect of the molecular weight (MW) of the diffusing polyelectrolyte (CHI) on the buildup process and observe a faster growth for low MW chitosan. The influence of the salt concentration during buildup is also investigated. Whereas the CHI/HA films grow rapidly at high salt concentration (0.15 M NaCl) with the formation of a uniform film after only a few deposition steps, it is very difficult to build the film at 10(-4) M NaCl. In this latter case, the deposited mass increases linearly with the number of deposition steps and the first deposition stage, where the surface is covered by islets, lasts at least up to 50 bilayer deposition steps. However, even at these low salt concentrations and in the islet configuration, CHI chains seem to diffuse in and out of the CHI/HA complexes. The linear mass increase of the film with the number of deposition steps despite the CHI diffusion is explained by a partial redissolution of the CHI/HA complexes forming the film during different steps of the buildup process. Finally

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-15

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

  2. Development of a computerized operator support system for BWR power plant

    International Nuclear Information System (INIS)

    Monta, K.; Sekimizu, K.; Sato, N.; Araki, T.; Mori, N.

    1985-01-01

    A computerized operator support system for BWR power plant has been developed since 1980 supported by the Japanese government. The main functions of the systems are post trip operational guidance, disturbance analysis, standby system management, operational margin monitoring and control rod operational guidance. The former two functions aim at protection against incidents during operation of nuclear power plants and the latter three functions aim at their prevention. As the final stage of the development, these functions are combined with the plant supervision function and are organized as an advanced man-machine interface for BWR power plant. During the above process, operator task analyses are performed to enable synthesis of these support functions for right fit to operator tasks and to realize a hierarchical structure for CRT displays for right fit to operators cognitive needs. (author)

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

    International Nuclear Information System (INIS)

    Ogata, H.; Kawamura, S.; Kumasaka, M.; Nishikubo, M.

    2001-01-01

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

  4. Study of the possibility of power uprate of a simplified BWR

    International Nuclear Information System (INIS)

    Kim, H.T.; Yang, A.I.; Tabata, H.; Montani, M.; Yoshioka, Y.

    1995-01-01

    Stability and critical power analyses were performed to study the possibility of power uprate for a large natural circulation BWR beyond 1000 MWe based on the thermal hydraulic performance requirement. The natural circulation flow required to satisfy the stability and critical power design criteria and the chimney height necessary to provide the natural circulation flow were determined by performing a parameter sensitivity study. Based on the study results, relationship between power level, fuel type, and size of reactor pressure vessel (diameter and height) was established. This paper presents the study results showing that power uprate beyond 1000 MWe for a large natural circulation BWR is quite feasible from the thermal hydraulic performance standpoint. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-07-01

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

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

    International Nuclear Information System (INIS)

    Machida, Eiji; Yusa, Noritaka

    2004-01-01

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

  7. Analysis of radiological consequences in a typical BWR with a mark-II containment

    International Nuclear Information System (INIS)

    Funayama, Kyoko; Kajimoto, Mitsuhiro

    2003-01-01

    INS/NUPEC in Japan has been carrying out the Level 3 PSA program. In the program, the MACCS2 code has been extensively applied to analyze radiological consequences for typical BWR and PWR plants in Japan. The present study deals with analysis of effects of the AMs, which were implemented by industries, on radiological consequence for a typical BWR with a Mark-II containment. In the present study, source terms and their frequencies of source terms were used based on results of Level 2 PSA taking into account AM countermeasures. Radiological consequences were presented with dose risks (Sv/ry), which were multiplied doses (Sv) by containment damage frequencies (/ry), and timing of radionuclides release to the environment. The results of the present study indicated that the dose risks became negligible in most cases taking AM countermeasures and evacuations. (author)

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

    International Nuclear Information System (INIS)

    Bregani, F.

    1990-01-01

    In order to check the effectiveness of direct chemical decontamination on small and complex components, usually considered for storage without decontamination because of the small amount, some tests were performed on the DECO experimental loop. Four small stainless steel valves from the primary system of the Garigliano BWR were decontaminated using mainly aggressive chemicals such as HC1, HF, HNO 3 and their mixtures. On two valves, before the treatment with aggressive chemicals, a step with soft chemical (oxalic and citric acid mixture) was performed in order to see whether a softening action enhances the following aggressive decontamination. Moreover, in order to increase as much as possible the decontamination effectiveness, a decontamination process using ultrasounds jointly with aggressive chemicals was investigated. After an intensive laboratory testing programme, two smaller stainless steel valves from the primary system of the Garigliano BWR were decontaminated using ultrasounds in aggressive chemical solutions

  9. Assessment of the fracture toughness of irradiated stainless steel for BWR core shrouds

    International Nuclear Information System (INIS)

    Carter, R.G.; Gamble, R.M.

    2002-01-01

    Data from previously performed experiments were collected and evaluated to determine the relationship between fracture toughness and neutron fluence for conditions representative of BWR core shrouds. This relationship together with EPFM (elastic-plastic fracture mechanics) analysis methods similar to those in Appendix K of Section XI of the ASME Code were used to compute margin against failure as a function of neutron fluence for postulated cracks in BWR core shrouds. The results indicate that EPFM analyses can be used for flaw evaluation of core shrouds at fluence levels less than 3.10 21 n/cm 2 (E > 1 MeV). At fluence levels equal to or greater than 3.10 21 n/cm 2 , LEFM (linear-elastic fracture mechanics) analyses should be used with K Ic = 55 MPa-(m) 0.5 . (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1983-05-01

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

  11. Investigations on the thermal-hydraulics of a natural circulation cooled BWR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Kok, H.V.; Hagen, T.H.J.J. van der; Mudde, R.F. [Delft Univ. of Technology (Netherlands)

    1995-09-01

    A scaled natural circulation loop facility has been built after the Dodewaard Boiling Water Reactor, which is the only operating natural circulation cooled BWR in the world. The loop comprises one fuel assembly, a riser with a downcomer and a condenser with a cooling system. Freon-12 is used as a scaling liquid. This paper reports on the first measurements done with this facility. Quantities like the circulation flow, carry-under and the void-fraction have been measured as a function of power, pressure, liquid level, riser length, condensate temperature and friction factors. The behavior of the circulation flow can be understood by considering the driving force. Special attention has been paid to the carry-under, which has been shown to have a very important impact on the dynamics of a natural circulation cooled BWR.

  12. 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.)

  13. Axial diagnostic system of finished rods BWR type

    International Nuclear Information System (INIS)

    Rivero G, T.; Rojas C, E.L.

    1996-01-01

    This system is employed as a final non destructive diagnostic system to verify the adequate distribution of the different enrichment through the can of nuclear fuel. The system is framed of traction mechanisms, a personal computer, a counting card and another card for the pass motor control, the nuclear electronics and the control program. The performance is based on the gamma radiation counting of the natural decay of uranium 235, this radiation is processed by the nuclear instrumentation for delivering a pulse by each gamma detected. (Author)

  14. Gamma Ray Buildup Factor for Finite Media in Energy Range (4-10) MeV for Al and Pb

    International Nuclear Information System (INIS)

    Al-Ani, L.A.; Goarge, L.E.; Mahdi, M.S.

    2015-01-01

    A computer program based on Monte Carlo method had been designed and written in visual basic computer language and utilized for simulating the classic problem of gamma ray beam incident on finite plane slabs of absorbing materials.The source geometry adopted in this program is plane normal source. Dose buildup factor of gamma photons in the absence and presence pair production effect have been calculated in the energy range (4-10) MeV for Aluminum and Lead up to 5 mean free path thickness.Dose buildup factor in the presence of pair production is higher than dose buildup factor in the absence of pair production effect.The deviation between the values of dose buildup factor in the presence and absence pair production is increased when the energy is increased within the studied energy range because the cross section for pair production is increased within the studied energy range

  15. Document turn-over analysis to determine need of NPP construction in build-up structures of reinforced concrete

    International Nuclear Information System (INIS)

    Vojpe, D.K.; Lyubavin, V.K.

    1986-01-01

    Document turn-over to determine used of NPP construction in build-up structures of reinforced concrete is carried out. Ways of improving determination of needs of NPP construction board in the mentioned structures are pointed out

  16. Application of process computers and colour CRT displays in the plant control room of a BWR

    International Nuclear Information System (INIS)

    Itoh, M.; Hayakawa, H.; Kawahara, H.; Neda, T.; Wakabayashi, Y.

    1983-01-01

    The recent application of a CRT display system in an 1100-MW(e) BWR plant control room and the design features of a new control room whose installation is planned for the next generation are discussed. As reactor unit capacity and the need for plant safety and reliability continue to increase, instrumentation and control equipment is growing in number and complexity. In consequence, control and supervision of plant operations require improvement. Thus, because of recent progress in the field of process computers and display equipment (colour CRTs), efficient improvements of the control room are under way in the Japanese BWR plant. In the recently constructed BWR plant (1100 MW(e)), five CRTs on the bench board and two process computers were additionally installed in the control room during the construction stage to improve plant control and supervisory functions by implementing the lessons learned from the Three Mile Island incident. The major functions of the new computers and display systems are to show integrated graphic displays of the plant status, to monitor the standby condition of the safety system, to show the condition of the integrated alarm system, etc. In practice, in the actual plant, this newly installed system performs well. On the basis of the experience gained in these activities, a new computerized control and monitoring system is now being designed for subsequent domestic BWR plants. This advanced system will incorporate not only the functions already mentioned, but also a surveillance guide system and plant automation. For future plants, a diagnostic system and an instructional system that can analyse a disturbance and give operational guidance to the plant operator are being developed in a government-sponsored programme. (author)

  17. Propagation of cracks by stress corrosion in conditions of BWR type reactor

    International Nuclear Information System (INIS)

    Merino C, F.J.; Fuentes C, P.

    2004-01-01

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

  18. BWR Mark I pressure suppression study: effect of downcomer fill level on the vertical load function

    International Nuclear Information System (INIS)

    Lai, W.; McCauley, E.W.

    1977-02-01

    The investigation reported forms a part of the BWR Mark I Pressure Suppression Experiment Program and is one of a series of small scale studies designed to evaluate limited aspects of the pool dynamics phenomena prior to conduct of the 1/5 scale air test series. Presented is an experimental study of the effect of downcomer fill level (DFL) on the vertical load function

  19. Design-by-analysis application to next generation BWR fuel thermal design method

    International Nuclear Information System (INIS)

    Yamamoto, Yasushi; Mitsutake, Toru

    2008-01-01

    For BWR core thermal design methods, the full mock-up thermal-hydraulic tests have been needed, such as the critical power measurement test, pressure drop measurement test and so on for more than thirty years since 1976. However, the full mock-up test required the high costs and large-scale test facility. Now, there are only a few test facilities in the world to perform the full mock-up thermal-hydraulic test. Moreover, for future BWR, the bundle size tends to be larger, because of reducing the plant construction costs and minimizing days taken for the periodic check. For instance, AB1600, improved ABWR, was proposed from Toshiba (Imaruoka et al., 2005), whose bundle size was 1.2 times larger than the conventional BWR fuel size. It is too expensive and far from realistic to perform the full mock-up thermal-hydraulic test for such a large size fuel bundle on every new bundle design. The new design procedure is exactly expected at reasonable costs and required to perform the large scale bundle design development, especially for the next generation BWR. Then, we propose a kind of Design by Analysis method combined with the alternative to the full mock-up tests. We called Practical Design-by-Analysis (PDBA) method even more effective and with better accuracy than Full Design-by-Analysis (FDBA). The PDBA method was applied successfully to produce the critical power correlation for the current 8x8 fuel. The estimation error was almost equal to that of the design correlation developed empirically with the full mock-up test data. (author)

  20. Valuation of power oscillations in a BWR after control rod banks withdrawal events

    International Nuclear Information System (INIS)

    Costa, A. L.; Pereira, C.; Da Silva, C. A. M.; Veloso, M. A. F.

    2009-01-01

    The out-of-phase mode of oscillation is a very challenging type of instability occurring in BWR (Boiling Water Reactor) and its study is relevant because of the safety implications related to the capability to promptly detect any such inadvertent occurrence by in-core neutron detectors, thus triggering the necessary countermeasures in terms of selected rod insertion or even reactor shutdown. In this work, control rod banks (CRB) withdrawal transient was considered to study the power instability occurring in a BWR. To simulate this transient, the control rod banks were continuously removed from the BWR core in different cases. The simulation resulted in a very large increase of power. To perform the instability simulations, the RELAP5/MOD3.3 thermal hydraulic system code was coupled with the PARCS/2.4 3D neutron kinetic code. Data from a real BWR, the Peach Bottom, have been used as reference conditions and reactor parameters. The trend of the mass flow rate, pressure, coolant temperature and the void fraction to four thermal hydraulic channels symmetrically located in the core with respect to the core centre, were taken. It appears that the velocity of the rod bank withdrawal is a very important aspect for reactor stability. The slowest CRB withdrawal (180 s) did not cause power perturbation while the fast removal (20 s) triggered a slow power oscillation that little by little amplified to reach levels of more 100% of the initial power after about 210 s. The investigation of the related thermo hydraulic parameters showed that the mass flow rate, the void fraction and also the coolant temperature began to oscillate at approximately the same time interval

  1. Improved point-kinetics model for the BWR control rod drop accident

    International Nuclear Information System (INIS)

    Neogy, P.; Wakabayashi, T.; Carew, J.F.

    1985-01-01

    A simple prescription to account for spatial feedback weighting effects in RDA (rod drop accident) point-kinetics analyses has been derived and tested. The point-kinetics feedback model is linear in the core peaking factor, F/sub Q/, and in the core average void fraction and fuel temperature. Comparison with detailed spatial kinetics analyses indicates that the improved point-kinetics model provides an accurate description of the BWR RDA

  2. Using ParaPost Tenax fiberglass and ParaCore build-up material to restore severely damaged teeth.

    Science.gov (United States)

    Caicedo, Ricardo; Castellon, Paulino

    2005-01-01

    This article describes a technique using ParaPost Tenax Fiber White, ParaPost Cement, and ParaPost ParaCore build-up material to restore a tooth with a significant loss of tooth structure. After successful root canal therapy, the posts were bonded in the canals and the core was built using ParaPost ParaCore build-up material. At that point, the tooth was prepared to receive a conventional porcelain-fused-to-metal crown.

  3. Product Evaluation Task Force Phase Two report for BWR/PWR dissolver wastes

    International Nuclear Information System (INIS)

    Francis, A.J.

    1990-01-01

    It has been proposed that all Intermediate Level Wastes arising at Sellafield should be encapsulated prior to ultimate disposal. The Product Evaluation Task Force (PETF) was set up to investigate possible encapsulants and to produce an adequate data base to justify the preferred matrices. This report details the work carried out, under Phase 2 of the Product Evaluation Task Force programme, on BWR/PWR Dissolver Wastes. Three possible types of encapsulants for BWR/PWR Dissolver Wastes:- Inorganic cements, Polymer cements and Polymers are evaluated using the Kepner Tregoe decision analysis technique. This technique provides a methodology for scoring and ranking alternative options and evaluating any risks associated with an option. The analysis shows that for all four stages of waste management operations ie Storage, Transport, handling and emplacement, Disposal and Process, cement matrices are considerably superior to other potential matrices. A matrix, consisting of three parts Blast Furnace Slag (BFS) to one part Ordinary Portland Cement (OPC), is recommended for Phase 3 studies on BWR/PWR Dissolver Wastes. (author)

  4. Effect of the inlet throttling on the thermal-hydraulic instability of the natural circulation BWR

    International Nuclear Information System (INIS)

    Furuya, Masahiro; Inada, Fumio; Yoneda, Kimitoshi

    1997-01-01

    Although it is well-established that inlet restriction has a stabilizing for forced circulation BWR, the effect of inlet on the thermal-hydraulic stability of natural circulation BWR remains unknown since increasing inlet restriction affect thermal-hydraulic stability due to reduction of the recirculation flow rate. Therefore experiments have been conducted to investigate the effect of inlet restriction on the thermal-hydraulic stability. A test facility used in this experiments was designed and constructed to have non-dimensional values which are nearly equal to those of natural circulation BWR. Experimental results showed that driving force of the natural circulation was described as a function of heat flux and inlet subcooling independent of inlet restriction. Stability maps in reference to the channel inlet subcooling, heat flux were presented for various inlet restriction which were carried out by an analysis based on the homogeneous flow various using this function. Instability region during the inlet subcooling shifted to the higher inlet subcooling with increasing inlet restriction and became larger with increasing heat flux. (author)

  5. Assessment of severe accident prevention and mitigation features: BWR, Mark II containment design

    International Nuclear Information System (INIS)

    Lehner, J.R.; Hsu, C.J.; Eltawila, F.; Perkins, K.R.; Luckas, W.J.; Fitzpatrick, R.G.; Pratt, W.T.

    1988-07-01

    Plant features and operator actions, which have been found to be important in either preventing or mitigating severe accidents in BWRs with Mark II containments (BWR Mark II's) have been identified. These features and actions were developed from insights derived from reviews of in-depth risk assessments performed specifically for the Limerick and Shoreham plants and from other relevant studies. Accident sequences that dominate the core-damage frequency and those accident sequences that are of potentially high consequence were identified. Vulnerabilities of the BWR Mark II to severe-accident containment loads were also noted. In addition, those features of a BWR Mark II, which are important for preventing core damage and are available for mitigating fission-product release to the environment were also identified. This report is issued to provide focus to an analyst examining an individual plant. This report calls attention to plant features and operator actions and provides a list of deterministic attributes for assessing those features and actions found to be helpful in reducing the overall risk for Mark II plants. Thus, the guidance is offered as a resource in examining the subject plant to determine if the same, or similar, plant features and operator actions will be of value in reducing overall plant risk. This report is intended to serve solely as guidance

  6. BWR control blade/channel box interaction models for SCDAP/RELAP5

    International Nuclear Information System (INIS)

    Griffin, F.P.

    1993-01-01

    The core of a boiling water reactor (BWR) consists of an array of fuel assemblies with cross-shaped control blades located between these assemblies. Each fuel assembly consists of a fuel rod bundle surrounded by a Zircaloy channel box. Each control blade consists of small stainless steel absorber tubes filled with B 4 C powder surrounded by a stainless steel blade sheath. Under severe accident conditions, material interactions between the B 4 C, stainless steel, and Zircaloy would have a significant impact on the melting and subsequent relocation of the control blade and channel box structures. This paper describes a new BWR control blade/channel box model for the SCDAP/RELAP5 severe accident analysis code that includes the effects of these material interactions. The phenomena represented by this model and the modeling techniques are derived from ORNL analyses of the BWR severe fuel damage experiments. Two examples of the operation of this new model within SCDAP/RELAP5 are provided

  7. Development of a coordinated control system for BWR nuclear power plant and HVDC transmission system

    International Nuclear Information System (INIS)

    Ishikawa, M.; Hara, T.; Hirayama, K.; Sekiya, K.

    1986-01-01

    The combined use of dc and ac transmissions or so-called hybrid transmission was under study, employing both dc and ac systems to enable stable transmission of 10,000 MW of electric power generated by the BWR nuclear plant, scheduled to be built about 800 km away from the center of the load. It was thus necessary to develop a hybrid power transmission control system, the hybrid power transmission system consisting of a high voltage dc transmission system (HVDC) and an ultrahigh ac transmission system (UHVAC). It was also necessary to develop a control system for HVDC transmission which protects the BWR nuclear power plant from being influenced by any change in transmission mode that occurs as a result of faults on the UHVAC side when the entire power of the BWR plant is being sent by the HVDC transmission. This paper clarifies the requirements for the HVDC system control during hybrid transmission and also during dc transmission. The control method that satisfies these requirements was studied to develop a control algorithm

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-10-01

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

  9. Analysis CFD for the hydrogen transport in the primary containment of a BWR

    International Nuclear Information System (INIS)

    Jimenez P, D. A.; Del Valle G, E.; Gomez T, A. M.

    2014-10-01

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

  10. Variation of energy absorption and exposure buildup factors with incident photon energy and penetration depth for boro-tellurite (B2O3-TeO2) glasses

    Science.gov (United States)

    Sayyed, M. I.; Elhouichet, H.

    2017-01-01

    The gamma ray energy absorption (EABF) and exposure buildup factors (EBF) of (100-x)TeO2-xB2O3 glass systems (where x=5, 10, 15, 20, 22.5 and 25 mol%) have been calculated in the energy region 0.015-15 MeV up to a penetration depth of 40 mfp (mean free path). The five parameters (G-P) fitting method has been used to estimate both EABF and EBF values. Variations of EABF and EBF with incident photon energy and penetration depth have been studied. It was found that EABF and EBF values were higher in the intermediate energy region, for all the glass systems. Furthermore, boro-tellurite glass with 5 mol% B2O3, was found to present the lowest EABF and EBF values, hence it is superior gamma-ray shielding material. The results indicate that the boro-tellurite glasses can be used as radiation shielding materials.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-07-01

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

  12. Predictive uncertainty reduction in coupled neutron-kinetics/thermal hydraulics modeling of the BWR-TT2 benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Badea, Aurelian F., E-mail: aurelian.badea@kit.edu [Karlsruhe Institute of Technology, Vincenz-Prießnitz-Str. 3, 76131 Karlsruhe (Germany); Cacuci, Dan G. [Center for Nuclear Science and Energy/Dept. of ME, University of South Carolina, 300 Main Street, Columbia, SC 29208 (United States)

    2017-03-15

    Highlights: • BWR Turbine Trip 2 (BWR-TT2) benchmark. • Substantial (up to 50%) reduction of uncertainties in the predicted transient power. • 6660 uncertain model parameters were calibrated. - Abstract: By applying a comprehensive predictive modeling methodology, this work demonstrates a substantial (up to 50%) reduction of uncertainties in the predicted total transient power in the BWR Turbine Trip 2 (BWR-TT2) benchmark while calibrating the numerical simulation of this benchmark, comprising 6090 macroscopic cross sections, and 570 thermal-hydraulics parameters involved in modeling the phase-slip correlation, transient outlet pressure, and total mass flow. The BWR-TT2 benchmark is based on an experiment that was carried out in 1977 in the NPP Peach Bottom 2, involving the closure of the turbine stop valve which caused a pressure wave that propagated with attenuation into the reactor core. The condensation of the steam in the reactor core caused by the pressure increase led to a positive reactivity insertion. The subsequent rise of power was limited by the feedback and the insertion of the control rods. The BWR-TT2 benchmark was modeled with the three-dimensional reactor physics code system DYN3D, by coupling neutron kinetics with two-phase thermal-hydraulics. All 6660 DYN3D model parameters were calibrated by applying a predictive modeling methodology that combines experimental and computational information to produce optimally predicted best-estimate results with reduced predicted uncertainties. Simultaneously, the predictive modeling methodology yields optimally predicted values for the BWR total transient power while reducing significantly the accompanying predicted standard deviations.

  13. Uncertainty analysis of pollutant build-up modelling based on a Bayesian weighted least squares approach

    International Nuclear Information System (INIS)

    Haddad, Khaled; Egodawatta, Prasanna; Rahman, Ataur; Goonetilleke, Ashantha

    2013-01-01

    Reliable pollutant build-up prediction plays a critical role in the accuracy of urban stormwater quality modelling outcomes. However, water quality data collection is resource demanding compared to streamflow data monitoring, where a greater quantity of data is generally available. Consequently, available water quality datasets span only relatively short time scales unlike water quantity data. Therefore, the ability to take due consideration of the variability associated with pollutant processes and natural phenomena is constrained. This in turn gives rise to uncertainty in the modelling outcomes as research has shown that pollutant loadings on catchment surfaces and rainfall within an area can vary considerably over space and time scales. Therefore, the assessment of model uncertainty is an essential element of informed decision making in urban stormwater management. This paper presents the application of a range of regression approaches such as ordinary least squares regression, weighted least squares regression and Bayesian weighted least squares regression for the estimation of uncertainty associated with pollutant build-up prediction using limited datasets. The study outcomes confirmed that the use of ordinary least squares regression with fixed model inputs and limited observational data may not provide realistic estimates. The stochastic nature of the dependent and independent variables need to be taken into consideration in pollutant build-up prediction. It was found that the use of the Bayesian approach along with the Monte Carlo simulation technique provides a powerful tool, which attempts to make the best use of the available knowledge in prediction and thereby presents a practical solution to counteract the limitations which are otherwise imposed on water quality modelling. - Highlights: ► Water quality data spans short time scales leading to significant model uncertainty. ► Assessment of uncertainty essential for informed decision making in water

  14. CFD predictions of standby liquid control system mixing in lower plenum of a BWR

    International Nuclear Information System (INIS)

    Boyd, Christopher; Skarda, Raymond

    2014-01-01

    Highlights: • Computational fluid dynamics analysis of BWR lower plenum. • Mixing and stratification of the standby liquid control system injection. • Scoping study highlights the expected flow paths and limitations of experiments. - Abstract: During an anticipated transient without scram (ATWS) scenario in certain boiling water reactor (BWR) systems, a standby liquid control system (SLCS) is used to inject a sodium pentaborate solution into the reactor system in order to quickly shut down (scram) the reactor without the use of the control rods. Some BWR designs utilize a SLCS that injects through a set of nozzles on a vertical pipe in the peripheral region of the lower plenum of the reactor vessel. During the scenario, system water levels are reduced and natural circulation flow rates down through the jet pump nozzles and up into the core are a small fraction of the rated system flow. It is during this period that the SLCS flows are considered. This work outlines some initial scoping studies completed by the staff at the Nuclear Regulatory Commission (NRC). An attempt at benchmarking the computational fluid dynamics (CFD) approach using a set of available test data from a small facility is outlined. Due to our lack of information related to specific details of the facility geometry along with the limited data available from the test, the benchmark exercise produced only a qualitative basis for selecting turbulence models and mesh density. A CFD model simulating a full-scale reactor system is developed for the lower plenum of a representative BWR/4 design and SLCS flows and mixing are studied under a range of flow conditions. The full-scale BWR simulation builds upon the lessons learned from the benchmark exercise. One challenge for this work is the large size of the domain and the relatively small size of the geometric details such as flow passages and gaps. The geometry is simplified to make meshing feasible by eliminating some of the small features. The

  15. Optimal Pile Arrangement for Minimizing Excess Pore Water Pressure Build-Up

    DEFF Research Database (Denmark)

    Barari, Amin; Saadati, Meysam; Ibsen, Lars Bo

    2013-01-01

    Numerical analysis of pile group in a liquefiable soil was considered to investigate the influence of pile spacing on excess pore pressure distribution and liquefaction potential. The analysis is conducted using a two-dimensional plain strain finite difference program considering a nonlinear...... constitutive model for sandy soil, strength and stiffness reduction, and pile-soil interaction. The Mohr-Coulomb constitutive model coupled with Byrne pore pressure build-up model have been employed in the analysis. Numerical analysis results show that pile groups have significant influence on the dynamic...... response of sandy soil as they reduce the amount of excess pore pressure development during seismic shaking and may even prevent liquefaction....

  16. The application of the LTSN method in the evaluation of the buildup factor

    International Nuclear Information System (INIS)

    Rodriguez, Barbara A.; Borges, Volnei; Zabadal, Jorge R.

    2005-01-01

    In this paper the LTS N method is used to solve the transport equation for photons in a heterogeneous medium, assuming the Klein-Nishina scattering kernel as the scattering differential cross section as well the multigroup model in the wavelength variable. The flux density of photons and the parameters of the medium are used for the calculation of the exposure buildup factor. We present numerical simulations and comparisons with available results in the literature for different compositions containing water, iron and lead. (author)

  17. Fusion-product ash buildup in tokamak with radial electric field

    International Nuclear Information System (INIS)

    Downum, W.B.; Choi, C.K.; Miley, G.H.

    1979-01-01

    The buildup of thermalized fusion products (ash) in a tokamak can seriously limit burn times. Prior studies have concentrated on deposition profile effects on alpha particle transport in tokamaks but have not considered the effect on ash of radial electric fields (either created internally, e.g. due to high-energy alpha leakage, or generated externally). The present study focuses on this issue since it appears that electric fields might offer one approach to control of the ash. Approximate field and source profiles are used, based on prior calculations

  18. Preliminary model for core/concrete interactions. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    Murfin, W.B.

    1977-08-01

    A preliminary model is described for computing the rate of penetration of concrete by a molten LWR core. Among the phenomena included are convective stirring of the melt by evolved gases, admixture of concrete decomposition products to the melt, chemical reactions, radiative heat loss, and variation of heat transfer coefficients with local pressure. The model is most applicable to a two-phase melt (metallic plus oxidic) having a fairly high metallic content.

  19. Preservice inspection of BWR plants using advanced ISI system

    International Nuclear Information System (INIS)

    Shouji, H.; Kobayashi, T.; Nagao, T.

    1994-01-01

    This paper reports experiences of preservice inspections using advanced Inservice Inspection (ISI) systems. The advanced ISI system consists of newly designed automated scanners, a high speed data acquisition/analysis station and automated pipe inspection system. This system has been developed to reduce examination time and radiation dose. The automated scanners are designed light weight and easy handling. A semi-automated inspection system has been used for pipe inspection. It is successful for taking reliable examination data, but not successful to reduce the radiation dose of personnel. For this reason, the automated pipe inspection system has been developed to replace the semi-automated system. The automated pipe inspection system consists of a small scanner, scanner controller and high speed personal computer for data acquisition. The data is analyzed by the data analysis station. The scanner has a light weight body in order to ease handling, and it requires almost the same clearance with manual examination. The PSI were successfully completed in a shorten time period. From this experience, the advanced ISI system will be very useful for ISI, especially reduction of radiation dose of personnel in future

  20. Bacterial leakage through temporary fillings in core buildup composite material - an in vitro study.

    Science.gov (United States)

    Rechenberg, Dan-Krister; Schriber, Martina; Attin, Thomas

    2012-08-01

    To evaluate the ability of the provisional filling material Cavit-W alone or in combination with different restorative materials to prevent bacterial leakage through simulated access cavities in a resin buildup material. LuxaCore resin cylinders were subdivided into 4 experimental groups (n = 30), plus a positive (n = 5) and a negative (n = 30) control group. One bore hole was drilled through each cylinder, except those in the negative control group (G1). The holes were filled with Cavit-W (G2), Cavit-W and Ketac-Molar (glassionomer cement, G3), Cavit-W and LuxaCore bonded with LuxaBond (G4), Cavit-W and LuxaCore (G5), or left empty (G6). Specimens were mounted in a two-chamber leakage setup. The upper chamber was inoculated with E. faecalis. An enterococci-selective broth was used in the lower chamber. Leakage was assessed for 60 days and compared using Fisher's exact test (α Cavit-W alone or combined with a glass-ionomer cement did not prevent bacterial leakage through a resin buildup material for two months. In contrast, covering Cavit-W with a bonded resin material resulted in a bacteria-tight seal for two months.

  1. Optical fiber sensors and their application in monitoring stress build-up in dental resin cements

    Science.gov (United States)

    Ottevaere, H.; Tabak, M.; Fernandez Fernandez, A.; Berghmans, F.; Thienpont, H.

    2005-09-01

    The field of optical fiber sensing is highly diverse and this diversity is perceived as a great advantage over more conventional sensors in that an optical sensor can be tailored to measure any of a myriad of physical parameters. In this paper we present a niche application for optical fiber sensors in the domain of biophotonics, namely the monitoring of stress build-up during the curing process of dental resin cements. We discuss the origin of this stress build-up and the problems it can cause when treating patients. Optical fiber sensors aim at excelling in two kind of applications: firstly to perform quality control on batch produced dental cements and measure their total material shrinkage, secondly to monitor the hardening of the cement during in-vivo measurements resulting in the dynamic measurement of the shrinkage and to control the stress in a facing based restoration. We therefore investigated two types of optical fiber sensors as alternatives to conventional measurement techniques; namely polarimetric optical fiber sensors and fiber Bragg gratings written in polarization maintaining fibers. After discussing the results obtained with both optical fiber sensors, we will conclude with a critical assessment of the suitability of the two proposed sensing configurations for multi-parameter stress monitoring.

  2. Nonsurgical treatment of skeletal anterior open bite in adult patients: Posterior build-ups.

    Science.gov (United States)

    Vela-Hernández, Arturo; López-García, Rocio; García-Sanz, Verónica; Paredes-Gallardo, Vanessa; Lasagabaster-Latorre, Felicidad

    2017-01-01

    To (1) evaluate the efficacy of build-ups in the correction of anterior open bite in adults, (2) evaluate their efficacy in producing molar intrusion, (3) analyze skeletal and dental changes, and (4) assess the long-term stability. The sample consisted of 93 lateral cephalograms from 31 patients with skeletal and dental anterior open bite. The patients had received orthodontic treatment consisting of bonded resin blocks on the maxillary molars combined with Tip-Edge Plus bracket appliances. Cephalometric measurements were performed on radiographs taken before treatment (T1), after treatment (T2), and after a retention period (T3), which were analyzed and compared. Significant dental and skeletal changes were observed after treatment. Molar intrusion averaging 1 mm; 1.44 and 1.57 mm extrusion of mandibular and maxillary incisors, respectively; and a mean of 3.98 mm overbite increase were observed. The mandibular plane angle showed a mean closure of 1.19°, and there was a mean decrease in anterior facial height of 0.7 mm. A mild relapse tendency was observed, but long-term stability was acceptable. Build-ups are an effective treatment alternative for anterior open bite in adults. Outcomes remain significantly stable during the retention period.

  3. A novel way to reduce thrombus build-up in vena cava filters.

    Science.gov (United States)

    Chen, Zengsheng; Zhan, Fan; Fan, Yubo; Deng, Xiaoyan

    2011-11-01

    The build-up of thromboses in vena cava filters after deployment presents serious problem to the patients. We proposed a novel way to overcome this problem in a belief that intentionally induced swirling flow can optimize blood flow patterns in vena cava filters, enhance the stirring motion of flow, in turn accelerate the dissolution of blood clots captured in the filter and facilitate blood to flow pass through the filters. In this study, we experimentally compared the work efficiency of a vena cava filter under swirling flow condition with that of the same filter under normal flow condition. The results show that when compared to the normal flow, the swirling flow indeed has a significantly beneficial effect on a VCF which can decrease its flow-out time nearly 40% and reduce clot build-up in the filter more than 50%. We therefore believe that the design of an ideal VCF should take how to create swirling flow in the filter into the consideration. Copyright © 2011 Wiley Periodicals, Inc.

  4. Buildup factors for multilayer shieldings in deterministic methods and their comparison with Monte Carlo

    International Nuclear Information System (INIS)

    Listjak, M.; Slavik, O.; Kubovcova, D.; Vermeersch, F.

    2009-01-01

    In general there are two ways how to calculate effective doses. The first way is by use of deterministic methods like point kernel method which is implemented in Visiplan or Microshield. These kind of calculations are very fast, but they are not very convenient for a complex geometry with shielding composed of more then one material in meaning of result precision. In spite of this that programs are sufficient for ALARA optimisation calculations. On other side there are Monte Carlo methods which can be used for calculations. This way of calculation is quite precise in comparison with reality but calculation time is usually very large. Deterministic method like programs have one disadvantage -usually there is option to choose buildup factor (BUF) only for one material in multilayer stratified slabs shielding calculation problems even if shielding is composed from different materials. In literature there are proposed different formulas for multilayer BUF approximation. Aim of this paper was to examine these different formulas and their comparison with MCNP calculations. At first ware compared results of Visiplan and Microshield. Simple geometry was modelled - point source behind single and double slab shielding. For Build-up calculations was chosen Geometric Progression method (feature of the newest version of Visiplan) because there are lower deviations in comparison with Taylor fitting. (authors)

  5. Buildup factors for multilayer shieldings in deterministic methods and their comparison with Monte Carlo

    International Nuclear Information System (INIS)

    Listjak, M.; Slavik, O.; Kubovcova, D.; Vermeersch, F.

    2008-01-01

    In general there are two ways how to calculate effective doses. The first way is by use of deterministic methods like point kernel method which is implemented in Visiplan or Microshield. These kind of calculations are very fast, but they are not very convenient for a complex geometry with shielding composed of more then one material in meaning of result precision. In spite of this that programs are sufficient for ALARA optimisation calculations. On other side there are Monte Carlo methods which can be used for calculations. This way of calculation is quite precise in comparison with reality but calculation time is usually very large. Deterministic method like programs have one disadvantage -usually there is option to choose buildup factor (BUF) only for one material in multilayer stratified slabs shielding calculation problems even if shielding is composed from different materials. In literature there are proposed different formulas for multilayer BUF approximation. Aim of this paper was to examine these different formulas and their comparison with MCNP calculations. At first ware compared results of Visiplan and Microshield. Simple geometry was modelled - point source behind single and double slab shielding. For Build-up calculations was chosen Geometric Progression method (feature of the newest version of Visiplan) because there are lower deviations in comparison with Taylor fitting. (authors)

  6. Fission gas pressure build-up and fast-breeder economy

    International Nuclear Information System (INIS)

    Engelmann, P.

    1962-01-01

    Fuel-cycle costs and doubling time of fast-breeder reactors are strongly affected by the fuel-burn-up obtainable. Use of oxide or carbide fuel offers the possibility of reaching a burn-up of 100 000 MWd/t. In fuel-clad elements, a limiting factor is the fission-gas-pressure build-up. At the high burn-up considered, an appreciable fraction of the fission gases gets into the pores and thus contributes to the pressure on the can. Starting from the known fission-product yields and decay chains, gas production and pressure build-up have been calculated. Three physical models have been employed in calculating the pressure acting upon the can : the gas is contained either in interconnected pores, in separate pores, or in a central hole. The pressure-dependence upon free volume (fuel density) and temperature will be discussed. Cans made of high-strength materials as Ineonel-X and molybdenum could stand the fission-gas pressure at operating temperatures. Unfortunately, these materials have higher absorption cross-sections than stainless steel. Results of a multi-group calculation are given, showing the effect of using these can materials and of decreasing the fuel density on critical mass and breeding ratio in small and medium-size breeders. (author) [fr

  7. Energy-dependent finite-orbit treatment for plasma buildup in mirror fusion devices

    International Nuclear Information System (INIS)

    Campbell, M.M.

    1980-01-01

    A computer simulation of hot plasma buildup in mirror fusion devices and results from this model are presented. In a small, hot magnetically confined plasma, the ion orbit radius (rho/sub i/) can be comparable to the plasma radius (R/sub p/). It a mirror-confined plasma were rho/sub i//R/sub p/ > 1/25 (such as 2XII-B), a point kinetic treatment of ion interactions becomes inaccurate and a finite gyro-radius (FGR) treatment must be used to adequately describe plasma buildup processes. This is particularly true for describing losses due to cold-gas charge exchange (c-x) near the plasma surface, since a particle lost near the vacuum interface may have contributed to the density as far as 2 rho/sub i/ radially inward from the c-x point. A similar FGR effect applies to beam-deposited ions whose large orbits influence the density up to 2 rho/sub i/ from the trapping point

  8. A STUDY INTO ICE BUILD-UP AND MELTING ON VERTICAL COOLED PIPES

    Directory of Open Access Journals (Sweden)

    Y. Zasiadko

    2016-08-01

    Full Text Available The use of cold accumulators based on the principle of ice build up on the cooled surfaces during off-peak periods and ice melting during on-peak periods is an effective method of electricity bills reduction. Within comparatively short periods of on-peak demand a noticeable amount of thermal energy related to ice melting is to be released, it becomes clear that not only sizing of ice accumulators based on balance calculations is actual, but also the determination of time periods of ice accumulation becomes critical. This work presents experimental unit for obtaining data on the ice build-up on the vertical cooled pipes and later on to continuously register data on the ice thickness diminishing at the regimes of ice melting when cooling of pipe stops. The data for ice build-up and melting for some regimes have been presented and analyzed. The data form the base for deriving semi-empirical correlations allowing to determine a time intervals necessary to generate of ice layers of a given thickness.

  9. Impact of eccentricity build-up and graveyard disposal Strategies on MEO navigation constellations

    Science.gov (United States)

    Radtke, Jonas; Domínguez-González, Raúl; Flegel, Sven K.; Sánchez-Ortiz, Noelia; Merz, Klaus

    2015-12-01

    With currently two constellations being in or close to the build-up phase, in a few years the Medium Earth Orbit (MEO) region will be populated with four complete navigation systems in relatively close orbital altitudes: The American GPS, Russian GLONASS, European Galileo, and Chinese BeiDou. To guarantee an appropriate visibility of constellation satellites from Earth, these constellations rely on certain defined orbits. For this, both the repeat pattern, which is basically defined by the semimajor axis and inclination, as well as the orbital planes, which are defined by the right ascension of ascending node, are determining values. To avoid an overcrowding of the region of interest, the disposal of satellites after their end-of-life is recommended. However, for the MEO region, no internationally agreed mitigation guidelines exist. Because of their distances to Earth, ordinary disposal manoeuvres leading to a direct or delayed re-entry due to atmospheric drag are not feasible: The needed fuel masses for such manoeuvres are by far above the reasonable limits and available fuel budgets. Thus, additional approaches have to be applied. For this, in general two options exist: disposal to graveyard orbits or the disposal to eccentricity build-up orbits. In the study performed, the key criterion for the graveyard strategy is that the disposed spacecraft must keep a safe minimum distance to the altitude of the active constellation on a long-term time scale of up to 200 years. This constraint imposes stringent requirements on the stability of the graveyard orbit. Similar disposals are also performed for high LEO satellites and disposed GEO payloads. The eccentricity build-up strategy on the other hand uses resonant effects between the Earth's geopotential, the Sun and the Moon. Depending on the initial conditions, these can cause a large eccentricity build-up, which finally can lead to a re-entry of the satellite. In this paper, the effects of applying either the first or

  10. An Empirical Model for Build-Up of Sodium and Calcium Ions in Small Scale Reverse Osmosis

    Directory of Open Access Journals (Sweden)

    Subriyer Nasir

    2011-05-01

    Full Text Available A simple models for predicting build-up of solute on membrane surface were formulated in this paper. The experiments were conducted with secondary effluent, groundwater and simulated feed water in small-scale of RO with capacity of 2000 L/d. Feed water used in the experiments contained varying concentrations of sodium, calcium, combined sodium and calcium. In order to study the effect of sodium and calcium ions on membrane performance, experiments with ground water and secondary effluent wastewater were also performed. Build-up of salts on the membrane surface was calculated by measuring concentrations of sodium and calcium ions in feed water permeate and reject streams using Atomic Absorption Spectrophotometer (AAS. Multiple linear regression of natural logarithmic transformation was used to develop the model based on four main parameters that affect the build-up of solute in a small scale of RO namely applied pressure, permeate flux, membrane resistance, and feed concentration. Experimental data obtained in a small scale RO unit were used to develop the empirical model. The predicted values of theoretical build-up of sodium and calcium on membrane surface were found in agreement with experimental data. The deviation in the prediction of build-up of sodium and calcium were found to be 1.4 to 10.47 % and 1.12 to 4.46%, respectively.

  11. Support vector regression model for the estimation of γ-ray buildup factors for multi-layer shields

    International Nuclear Information System (INIS)

    Trontl, Kresimir; Smuc, Tomislav; Pevec, Dubravko

    2007-01-01

    The accuracy of the point-kernel method, which is a widely used practical tool for γ-ray shielding calculations, strongly depends on the quality and accuracy of buildup factors used in the calculations. Although, buildup factors for single-layer shields comprised of a single material are well known, calculation of buildup factors for stratified shields, each layer comprised of different material or a combination of materials, represent a complex physical problem. Recently, a new compact mathematical model for multi-layer shield buildup factor representation has been suggested for embedding into point-kernel codes thus replacing traditionally generated complex mathematical expressions. The new regression model is based on support vector machines learning technique, which is an extension of Statistical Learning Theory. The paper gives complete description of the novel methodology with results pertaining to realistic engineering multi-layer shielding geometries. The results based on support vector regression machine learning confirm that this approach provides a framework for general, accurate and computationally acceptable multi-layer buildup factor model

  12. Decay profiles of β and γ for a radionuclide inventory in equilibrium cycle of a BWR type reactor

    International Nuclear Information System (INIS)

    Salaices, M.; Sandoval, S.; Ovando, R.

    2007-01-01

    Presently work the β and γ radiation decay profiles for a radionuclides inventory in equilibrium cycle of a BWR type reactor is presented. The profiles are presented in terms of decay in the activity of the total inventory as well as of the chemical groups that conform the inventory. In the obtaining of the radionuclides inventory in equilibrium cycle the ORIGEN2 code, version 1 was used, which simulates fuel burnup cycles and it calculates the evolution of the isotopic composition as a result of the burnt one, irradiation and decay of the nuclear fuel. It can be observed starting from the results that the decrease in the activity for the initial inventory and the different chemical groups that conform it is approximately proportional to the base 10 logarithm of the time for the first 24 hours of having concluded the burnt one. It can also be observed that the chemical groups that contribute in more proportion to the total activity of the inventory are the lanthanides-actinides and the transition metals, with 39% and 28%, respectively. The groups of alkaline earth metals, halogens, metalloids, noble gases and alkaline metals, contribute with percentages that go from the 8 to 5%. The groups that less they contribute to the total activity of the inventory they are the non metals and semi-metals with smaller proportions that 1%. The chemical groups that more contribute to the energy of β and γ radiation its are the transition metals and the lanthanides-actinides with a change in the order of importance at the end of the 24 hours period. The case of the halogens is of relevance for the case of the γ radiation energy due that occupying the very near third site to the dimensions of the two previous groups. Additionally, the decay in the activity for the total inventory and the groups that conform it can be simulated by means of order 6 polynomials or smaller than describe its behavior appropriately. The results presented in this work, coupled to a distribution model

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-10-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-15

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

  19. Evaluation of simplified analytical models for CO2 plume movement and pressure buildup

    Science.gov (United States)

    Oruganti, Y.; Mishra, S.

    2011-12-01

    CO2 injection into the sub-surface is emerging as a viable technology for reducing anthropogenic CO2 emissions into the atmosphere. When large amounts of CO2 are sequestered, pressure buildup is an associated risk, along with plume movement beyond the injected domain. In this context, simple modeling tools become valuable assets in preliminary CO2 injection project screening and implementation phases. This study presents an evaluation of two commonly used simplified analytical models for plume movement and pressure buildup, (1) the sharp interface model of Nordbotten et al. (2005), and the corresponding pressure distribution solution of Mathias et al. (2008), and (2) the 3-region model of Burton et al. (2008) based on fractional flow and steady-state pressure gradient considerations. The three-region model of Burton et al. assumes a constant pressure outer boundary. In this study, we incorporate the radius of investigation of the pressure front as the transient pressure boundary, in order to represent an infinite-acting system. The sharp-interface model also assumes the system to be infinite-acting. Temperature and pressure conditions used in these models correspond to the "warm, shallow" and "cold, deep" aquifer conditions as defined by Nordbotten et al. The saturation and pressure profiles as well as injection-well pressure buildup predicted by the analytical models are compared with those from the numerical simulator STOMP in order to provide a verification of the simplified modeling assumptions. Both the STOMP results and the three-region model show two sharp fronts (the drying and two-phase fronts), and a good match is obtained between the front positions at any time. For the sharp interface model, the vertically averaged gas saturation does not exhibit two sharp fronts as seen in the STOMP simulations, but shows a gradual change in saturation with radial distance over the two-phase region. The pressure profiles from STOMP and the analytical model are

  20. The Build-Up to Eruptive Solar Events Viewed as the Development of Chiral Systems

    Science.gov (United States)

    Martin, S. F.; Panasenco, O.; Berger, M. A.; Engvold, O.; Lin, Y.; Pevtsov, A. A.; Srivastava, N.

    2012-12-01

    When we examine the chirality or observed handedness of the chromospheric and coronal structures involved in the long-term build-up to eruptive events, we find that they evolve in very specific ways to form two and only two sets of large-scale chiral systems. Each system contains spatially separated components with both signs of chirality, the upper portion having negative (positive) chirality and the lower part possessing positive (negative) chirality. The components within a system are a filament channel (represented partially by sets of chromospheric fibrils), a filament (if present), a filament cavity, sometimes a sigmoid, and always an overlying arcade of coronal loops. When we view these components as parts of large-scale chiral systems, we more clearly see that it is not the individual components of chiral systems that erupt but rather it is the approximate upper parts of an entire evolving chiral system that erupts. We illustrate the typical pattern of build-up to eruptive solar events first without and then including the chirality in each stage of the build-up. We argue that a complete chiral system has one sign of handedness above the filament spine and the opposite handedness in the barbs and filament channel below the filament spine. If the spine has handedness, the observations favor its having the handedness of the filament cavity and coronal loops above. As the separate components of a chiral system form, we show that the system appears to maintain a balance of right-handed and left-handed features, thus preserving an initial near-zero net helicity. We further argue that the chiral systems allow us to identify key sites of energy transformation and stored energy later dissipated in the form of concurrent CMEs, erupting filaments and solar flares. Each individual chiral system may produce many successive eruptive events above a single filament channel. Because major eruptive events apparently do not occur independent of, or outside of, these unique

  1. Differential aquaporin 4 expression during edema build-up and resolution phases of brain inflammation

    Directory of Open Access Journals (Sweden)

    Brochet Bruno

    2011-10-01

    Full Text Available Abstract Background Vasogenic edema dynamically accumulates in many brain disorders associated with brain inflammation, with the critical step of edema exacerbation feared in patient care. Water entrance through blood-brain barrier (BBB opening is thought to have a role in edema formation. Nevertheless, the mechanisms of edema resolution remain poorly understood. Because the water channel aquaporin 4 (AQP4 provides an important route for vasogenic edema resolution, we studied the time course of AQP4 expression to better understand its potential effect in countering the exacerbation of vasogenic edema. Methods Focal inflammation was induced in the rat brain by a lysolecithin injection and was evaluated at 1, 3, 7, 14 and 20 days using a combination of in vivo MRI with apparent diffusion coefficient (ADC measurements used as a marker of water content, and molecular and histological approaches for the quantification of AQP4 expression. Markers of active inflammation (macrophages, BBB permeability, and interleukin-1β and markers of scarring (gliosis were also quantified. Results This animal model of brain inflammation demonstrated two phases of edema development: an initial edema build-up phase during active inflammation that peaked after 3 days (ADC increase was followed by an edema resolution phase that lasted from 7 to 20 days post injection (ADC decrease and was accompanied by glial scar formation. A moderate upregulation in AQP4 was observed during the build-up phase, but a much stronger transcriptional and translational level of AQP4 expression was observed during the secondary edema resolution phase. Conclusions We conclude that a time lag in AQP4 expression occurs such that the more significant upregulation was achieved only after a delay period. This change in AQP4 expression appears to act as an important determinant in the exacerbation of edema, considering that AQP4 expression is insufficient to counter the water influx during the build-up

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

  3. BWR transient analysis using neutronic / thermal hydraulic coupled codes including uncertainty quantification

    International Nuclear Information System (INIS)

    Hartmann, C.; Sanchez, V.; Tietsch, W.; Stieglitz, R.

    2012-01-01

    The KIT is involved in the development and qualification of best estimate methodologies for BWR transient analysis in cooperation with industrial partners. The goal is to establish the most advanced thermal hydraulic system codes coupled with 3D reactor dynamic codes to be able to perform a more realistic evaluation of the BWR behavior under accidental conditions. For this purpose a computational chain based on the lattice code (SCALE6/GenPMAXS), the coupled neutronic/thermal hydraulic code (TRACE/PARCS) as well as a Monte Carlo based uncertainty and sensitivity package (SUSA) has been established and applied to different kind of transients of a Boiling Water Reactor (BWR). This paper will describe the multidimensional models of the plant elaborated for TRACE and PARCS to perform the investigations mentioned before. For the uncertainty quantification of the coupled code TRACE/PARCS and specifically to take into account the influence of the kinetics parameters in such studies, the PARCS code has been extended to facilitate the change of model parameters in such a way that the SUSA package can be used in connection with TRACE/PARCS for the U and S studies. This approach will be presented in detail. The results obtained for a rod drop transient with TRACE/PARCS using the SUSA-methodology showed clearly the importance of some kinetic parameters on the transient progression demonstrating that the coupling of a best-estimate coupled codes with uncertainty and sensitivity tools is very promising and of great importance for the safety assessment of nuclear reactors. (authors)

  4. On the possibility of a decay ratio jump under continuous parameter variation in a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Lange, Carsten; Hennig, Dieter; Hurtado, Antonio [Technische Univ. Dresden (Germany). Professur fuer Wasserstoff- und Kernenergietechnik

    2012-11-01

    The objective of this paper is to demonstrate that specific methods of the nonlinear dynamics like bifurcation analysis are helpful to understand the temporal behavior of the nonlinear dynamical system BWR in many details. So, the stability state of the system changes discontinuously if a nonlinear dynamical system encountered a Hopf bifurcation under parameter variation. BWR stability analysis is performed in the most cases by application of system codes which provide the time evolution of the neutron flux or thermal power at a defined operational point (OP) after imposing a system parameter perturbation (like a control rod sinusoidal movement). However, in general, we are not able to understand the real stability state of the BWR at a specific OP by application of system code analysis alone. We have shown in the paper that complex dynamical states with coexisting stability modes (fixed points, stable and unstable limit cycles) could spontaneously emerge under selected system parameter variations. Hence, even in the relative simple case where stable fixed points and unstable limit cycles coexist, in the framework of a stability test the unstable dynamical mode could be overlooked because the simple stability linear indicator ''decay ratio'' does not indicate an unstable state for small amplitude perturbations. Or, e.g., both, the in-phase and the out-of-phase oscillation mode can encounter a Hopf bifurcation (everyone for itself) resulting in a change of the stability characteristics which cannot be ''detected'' by the asymptotic decay ratio. (orig.)

  5. Humidity Build-Up in a Typical Electronic Enclosure Exposed to Cycling Conditions and Effect on Corrosion Reliability

    DEFF Research Database (Denmark)

    Conseil, Helene; Gudla, Visweswara Chakravarthy; Jellesen, Morten Stendahl

    2016-01-01

    The design of electronic device enclosures plays a major role in determining the humidity build-up inside the device as a response to the varying external humidity. Therefore, the corrosion reliability of electronic devices has direct connection to the enclosure design. This paper describes...... the internal humidity build-up in a typical enclosure prescribed for electronic applications as a function of external humidity conditions and enclosure-related parameters. Investigated parameters include external temperature and humidity conditions, the temperature and time of the internal heating cycle......, thermal mass, and port/opening size. The effect of the internal humidity build-up on corrosion reliability has been evaluated by measuring the leakage current (LC) on interdigitated test comb patterns, which are precontaminated with sodium chloride and placed inside the enclosure. The results showed...

  6. Aesthetic Closure of Maxillary and Mandibular Anterior Spaces Using Direct Composite Resin Build-Ups: A Case Report

    Directory of Open Access Journals (Sweden)

    Schick Simona-Georgiana

    2016-07-01

    Full Text Available The presence of multiple spaces in the anterior aesthetic zone can produce discomfort for patients and its treatment can be difficult for dental professionals. A variety of treatment options are available and these include orthodontic movement, prosthetic indirect restorations or direct composite resin build-ups. Among these, the closure of interdental spaces using composite build-ups combined with orthodontic treatment is considered to be most conservative. This type of treatment has several advantages like the maximum preservation of tooth substance (no tooth preparation, no need for anesthesia, no multiple time-consuming visits, no provisional restorations and also comparably low costs. Clinical Consideration: This case report describes the clinical restorative procedure of direct composite resin build-ups for the closure of multiple anterior spaces.

  7. Energy absorption buildup factors of human organs and tissues at energies and penetration depths relevant for radiotherapy and diagnostics

    DEFF Research Database (Denmark)

    Manohara, S. R.; Hanagodimath, S. M.; Gerward, Leif

    2011-01-01

    Energy absorption geometric progression (GP) fitting parameters and the corresponding buildup factors have been computed for human organs and tissues, such as adipose tissue, blood (whole), cortical bone, brain (grey/white matter), breast tissue, eye lens, lung tissue, skeletal muscle, ovary......, testis, soft tissue, and soft tissue (4-component), for the photon energy range 0.015-15 MeV and for penetration depths up to 40 mfp (mean free path). The chemical composition of human organs and tissues is seen to influence the energy absorption buildup factors. It is also found that the buildup factor...... of human organs and tissues changes significantly with the change of incident photon energy and effective atomic number, Zeff. These changes are due to the dominance of different photon interaction processes in different energy regions and different chemical compositions of human organs and tissues...

  8. A computerized operator support system for BWR power plant during normal and abnormal conditions

    International Nuclear Information System (INIS)

    Monta, K.; Sato, N.; Tsunoyama, S.; Sekimizu, K.; Mori, N.

    1984-01-01

    A computerized operator support system for BWR was developed based on the operators tasks which are divided into problem solving and controlling. Problem solving comprises fault managing and planning tasks. One of the operator supports for the problem solving is the automatic diagnostic support. However, the operator supports for controlling tasks are still important, e.g. to support its rule memory and rule execution. In the system presented here, operator supports are considered for both of these two categories. In the paper two subsystems for the latter are mainly described. (author)

  9. Proceedings of the international specialist meeting on BWR-pressure suppression containment technology. Vol. 1

    International Nuclear Information System (INIS)

    Schultheiss, G.F.

    1981-01-01

    In the frame of R + D-work for BWR-pressure suppression systems the GKSS-Forschungszentrum Geesthacht GmbH organized an international specialist meeting. All important safety relevant aspects of pressure suppression system technology have been included. About 60 experts from USA, Japan, Sweden, Italy, Netherlands and the Federal Republic of Germany participated. They came from licensing authorities, vendors, research centers and universities. In 24 papers they have shown the world-wide present status of theoretical and experimental know-how on pressure suppression system behaviour. In discussions and working groups recommendations for future work have been compiled. (orig.) [de

  10. Connection between end plates and rods in a BWR fuel element

    International Nuclear Information System (INIS)

    Cali', G.P.

    1975-01-01

    The problem of the connection between the end plates and the rods of a BWR fuel element is analytically formulated. The behaviour of the springs coupling the rods with the upper plate is analyzed with particular detail since the deformation of these springs affects the forces at the interface of the fuel element structure components. A tool is given to design the springs according to some considerations regarding the mechanical strength of the interacting components as well as the influence of the possible geometrical unevennes of the system that can arise during the fuel element lifetime. (Cali', G.P.)

  11. The design and use of proficiency based BWR reactor maintenance and refuelling training mockups

    International Nuclear Information System (INIS)

    Ford, G.J.

    1996-01-01

    The purpose of this paper is to describe the ABB experience with the design and use of boiling water reactor training facilities. The training programs were developed and implemented in cooperation with the nuclear utilities. ABB operates two facilities, the ABB ATOM Light Water Reactor Service Center located in Vasteras, Sweden, and the ABB Combustion Engineering Nuclear Operations BWR Training Center located in Chattanooga, Tennessee, USA. The focus of the training centers are reactor maintenance and refueling activities plus the capability to develop and qualify tools, procedures and repair techniques

  12. Analysis of the integrity of the pressure vessel of the BWR type nuclear reactor

    International Nuclear Information System (INIS)

    Silva Luna, O.

    1982-01-01

    The presssure vessel of a BWR type reactor was monitored for cracking during alternating events of its in-service life. The monitoring was to determine criticality of fractures catastrophic fractures and the velocity of fracture propagation. Detected cracks were evaluated as specified in ASME code section XI, of a minimum wall thickness of 2.5% crack growths were compared a) of 1/10 of the critical maximum size and b) at in-service inspection intervals according to ASME recommendations to be established at the Laguna Verde nuclear plant. Finally conclusions are made and discussed. (author)

  13. 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

  14. Electrical equipment performance under severe accident conditions (BWR/Mark 1 plant analysis): Summary report

    Energy Technology Data Exchange (ETDEWEB)

    Bennett, P.R.; Kolaczkowski, A.M.; Medford, G.T.

    1986-09-01

    The purpose of the Performance Evaluation of Electrical Equipment during Severe Accident States Program is to determine the performance of electrical equipment, important to safety, under severe accident conditions. In FY85, a method was devised to identify important electrical equipment and the severe accident environments in which the equipment was likely to fail. This method was used to evaluate the equipment and severe accident environments for Browns Ferry Unit 1, a BWR/Mark I. Following this work, a test plan was written in FY86 to experimentally determine the performance of one selected component to two severe accident environments.

  15. LOFT test support branch data abstract report: one-sixth scale model BWR jet pump test

    International Nuclear Information System (INIS)

    Crapo, H.S.

    1979-01-01

    Pump performance data are presented for a 1/6 scale model jet pump in tests conducted at the LOFT Test Support Blowdown Facility. Steady-state subcooled pump characterization tests were performed over a wide range of forward and reverse flow conditions, both at room temperature, and at elevated temperature (555 0 K). Blowdown tests were also performed to obtain two-phase performance data in configurations simulating the flow patterns in the intact and broken loops of a BWR during a recirculation line break transient

  16. 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

  17. Analysis of results of AZTRAN and AZKIND codes for a BWR

    International Nuclear Information System (INIS)

    Bastida O, G. E.; Vallejo Q, J. A.; Galicia A, J.; Francois L, J. L.; Xolocostli M, J. V.; Rodriguez H, A.; Gomez T, A. M.

    2016-09-01

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

  18. Description of the power plant model BWR-plasim outlined for the Barsebaeck 2 plant

    International Nuclear Information System (INIS)

    Christensen, P. la Cour.

    1979-08-01

    A description is given of a BWR power plant model outlined for the Barsebaeck 2 plant with data placed at our disposal by the Swedish Power Company Sydkraft A/B. The basic operations are derived and simplifications discussed. The model is implemented with a simulation system DYSYS which assures reliable solutions and easy programming. Emphasis has been placed on the models versatility and flexibility so new features are easy to incorporate. The model may be used for transient calculations for both normal plant conditions and for abnormal occurences as well as for control system studies. (author)

  19. Interpolation of Gamma-ray buildup Factors for Arbitrary Source Energies in the Vicinity of the K-edge

    International Nuclear Information System (INIS)

    Michieli, I.

    1998-01-01

    Recently, a new buildup factors approximation formula based on the expanded polynomial set (E-P function) was successfully introduced (Michieli 1994.) with the maximum approximation error below 4% throughout the standard data domain. Buildup factors interpolation in E-P function parameters for arbitrary source energies, near the K-edge in lead, was satisfactory. Maximum interpolation error, for lead, lays within 12% what appears to be acceptable for most Point Kernel application. 1991. Harima at. al., showed that, near the K-edge, fluctuation in energy of exposure rate attenuation factors i.e.: D(E)B(E, μ E r)exp(-μ E r), given as a function of penetration depth (r) in ordinary length units (not mfps.), is not nearly as great as that of buildup factors. That phenomenon leads to the recommendation (ANSI/ANS-6.4.3) that interpolations in that energy range should be made in the attenuation factors B(E, μ E r)exp(-μ E r) rather than in the buildup factors alone. In present article, such interpolation approach is investigated by applying it to the attenuation factors in lead, with E-P function representation of exposure buildup factors. Simple form of the E-P function leads to strait calculation of new function parameters for arbitrary source energy near the K-edge and thus allowing the same representation form of buildup factors as in the standard interpolation procedure. results of the interpolation are discussed and compared with those from standard approach. (author)

  20. Electron-Cloud Build-up in the FNAL Main Injector

    International Nuclear Information System (INIS)

    Furman, M.A.

    2007-01-01

    We present a summary on ongoing simulation results for the electron-cloud buildup in the context of the proposed FNAL Main Injector (MI) intensity upgrade [1] in a fieldfree region at the location of the RFA electron detector [2]. By combining our simulated results for the electron flux at the vacuum chamber wall with the corresponding measurements obtained with the RFA we infer that the peak secondary electron yield (SEY) (delta) max is ∼> 1.4, and the average electron density is n e ∼> 10 10 m -3 at transition energy for the specific fill pattern and beam intensities defined below. The sensitivity of our results to several variables remains to be explored in order to reach more definitive results. Effects from the electron cloud on the beam are being investigated separately [3

  1. On the reversibility of cake buildup and compression in a membrane bioreactor

    DEFF Research Database (Denmark)

    Jørgensen, Mads Koustrup; Keiding, Kristian; Christensen, Morten Lykkegaard

    2014-01-01

    on compressed cake layers is released, the cake swells back to a looser structure. Based on these observations, we discuss the validity of using the critical flux concept to study fouling irreversibility. Modeling data of short-term filtration tests shows that the presence of a critical flux for irreversible......Fouling in a membrane bioreactor was studied by describing the reversibility of fouling developing during short-term experiments. Data were fitted to a recently proposed model of the buildup and compression of fouling layers. Shear stepping experiments performed to characterize the efficiency...... of increased shear rates at removing cake layers indicated that cake layer removal follows the same kinetics as does cake layer development, so the fouling layers can be characterized as removable fouling. Furthermore, transmembrane pressure stepping indicates compression reversibility, so when the pressure...

  2. Aircraft wing weight build-up methodology with modification for materials and construction techniques

    Science.gov (United States)

    York, P.; Labell, R. W.

    1980-01-01

    An aircraft wing weight estimating method based on a component buildup technique is described. A simplified analytically derived beam model, modified by a regression analysis, is used to estimate the wing box weight, utilizing a data base of 50 actual airplane wing weights. Factors representing materials and methods of construction were derived and incorporated into the basic wing box equations. Weight penalties to the wing box for fuel, engines, landing gear, stores and fold or pivot are also included. Methods for estimating the weight of additional items (secondary structure, control surfaces) have the option of using details available at the design stage (i.e., wing box area, flap area) or default values based on actual aircraft from the data base.

  3. Natural refuge crops, buildup of resistance, and zero-refuge strategy for Bt cotton in China.

    Science.gov (United States)

    Qiao, FangBin; Huang, JiKun; Rozelle, Scott; Wilen, James

    2010-10-01

    In the context of genetically modified crops expressing the Bacillus thuringiensis (Bt) toxin, a 'refuge' refers to a crop of the same or a related species that is planted nearby to enable growth and reproduction of the target pest without the selection pressure imposed by the Bt toxin. The goal of this study is to discuss the role of natural refuge crops in slowing down the buildup of resistance of cotton bollworm (CBW), and to evaluate China's no-refuge policy for Bt cotton. We describe in detail the different factors that China should consider in relation to the refuge policy. Drawing on a review of scientific data, economic analyses of other cases, and a simulation exercise using a bio-economic model, we show that in the case of Bt cotton in China, the no-refuge policy is defensible.

  4. Resonant laser power build-up in ALPS. A 'light-shining-through-walls' experiment

    International Nuclear Information System (INIS)

    Ehret, Klaus; Ghazaryan, Samvel; Frede, Maik

    2009-05-01

    The ALPS collaboration runs a light-shining-through-walls (LSW) experiment to search for photon oscillations into weakly interacting sub-eV particles (WISPs) inside of a superconducting HERA dipole magnet at the site of DESY. In this paper we report on the first successful integration of a large-scale optical cavity to boost the available power for WISP production in this type of experiments. The key elements are a frequency tunable narrow line-width continuous wave laser acting as the primary light source and an electronic feed-back control loop to stabilize the power build-up. We describe and characterize our apparatus and demonstrate the data analysis procedures on the basis of a brief exemplary run. (orig.)

  5. Influence of fast alpha diffusion and thermal alpha buildup on tokamak reactor performance

    International Nuclear Information System (INIS)

    Uckan, N.A.; Tolliver, J.S.; Houlberg, W.A.; Attenberger, S.E.

    1988-01-01

    The effect of fast alpha diffusion and thermal alpha accumulation on the confinement capability of a candidate Engineering Test Reactor plasma (Tokamak Ignition/Burn Experimental Reactor) in achieving ignition and steady-state driven operation has been assessed using both global and 1-1/2-dimensional transport models. Estimates are made of the threshold for radial diffusion of fast alphas and thermal alpha buildup. It is shown that a relatively low level of radial transport, when combined with large gradients in the fast alpha density, leads to a significant radial flow with a deleterious effect on plasma performance. Similarly, modest levels of thermal alpha concentration significantly influence the ignition and steady-state burn capability

  6. Particle simulation of pedestal buildup and study of pedestal scaling law in a quiescent plasma edge

    International Nuclear Information System (INIS)

    Chang, C.S.; Ku, S.; Weitzner, H.; Groebner, R.; Osborne, T.

    2005-01-01

    A discrete guiding-center particle code XGC (X-point included Guiding Center code) is used to study pedestal buildup and sheared E r formation in a quiescent plasma edge of a diverted tokamak. A neoclassical pedestal scaling law has been deduced, which shows that the density pedestal width is proportional to T i 1/2 M 1/2 /B t where T i is the ion temperature, M is ion mass and B t is the toroidal magnetic field. Dependence on the pedestal density or the poloidal magnetic field is found to be much weaker. Ion temperature pedestal is not as well defined as the density pedestal. Neoclassical electron transport rate, including the collisional heat exchange rate with ions, is too slow to be considered in the time scale of simulation (∼ 10 ms). (author)

  7. A modified method of calculating the lateral build-up ratio for small electron fields

    International Nuclear Information System (INIS)

    Tyner, E; McCavana, P; McClean, B

    2006-01-01

    This note outlines an improved method of calculating dose per monitor unit values for small electron fields using Khan's lateral build-up ratio (LBR). This modified method obtains the LBR directly from the ratio of measured, surface normalized, electron beam percentage depth dose curves. The LBR calculated using this modified method more accurately accounts for the change in lateral scatter with decreasing field size. The LBR is used along with Khan's dose per monitor unit formula to calculate dose per monitor unit values for a set of small fields. These calculated dose per monitor unit values are compared to measured values to within 3.5% for all circular fields and electron energies examined. The modified method was further tested using a small triangular field. A maximum difference of 4.8% was found. (note)

  8. Study of Gamma Ray Exposure Buildup Factor for Some Ceramics with Photon Energy, Penetration Depth and Chemical Composition

    Directory of Open Access Journals (Sweden)

    Tejbir Singh

    2013-01-01

    Full Text Available Gamma ray exposure buildup factor for some ceramics such as boron nitride (BN, magnesium diboride (MgB2, silicon carbide (SiC, titanium carbide (TiC and ferrite (Fe3O4 has been computed using five parametric geometric progression (G.P. fitting method in the energy range of 0.015 to 15.0 MeV, up to the penetration of 40 mean free path (mfp. The variation of exposure buildup factors for all the selected ceramics with incident photon energy, penetration depth, and chemical composition has been studied.

  9. Thermal-hydraulics stability of natural circulation BWR under startup. Flashing effects

    International Nuclear Information System (INIS)

    Hu, Rui; Kazimi, Mujid S.

    2009-01-01

    To help achieve the necessary natural circulation flow, a fairly long chimney is installed in a boiling natural circulation reactor like the ESBWR. In such systems, thermal-hydraulic stability during low pressure start-up should be examined while considering the flashing induced by the pressure drop in the channel and the chimney due to gravity head. In this work, a BWR stability analysis code in the frequency domain, named FISTAB (Flashing-Induced STability Analysis for BWR), was developed to address the issue of flashing-induced instability. A thermal-hydraulics non-homogeneous equilibrium model (NHEM) based on a drift flux formulation along with a lumped fuel dynamics model is incorporated in the work. The vapor generation rate is derived from the mixture energy conservation equation while considering the effect of flashing. The functionality of the FISTAB code was confirmed by comparison to experimental results from SIRIUS-N facility at CRIEPI, Japan. Both stationary and perturbation results agree well with the experimental results. (author)

  10. Assessment of Reactivity Effects due to Localized Perturbations in BWR Lattices

    International Nuclear Information System (INIS)

    Geemert, R. van; Jatuff, F.; Grimm, P.; Chawla, R.

    2002-01-01

    Optimization criteria for the representability of numerical models for the estimation of relative reactivity changes, due to localized perturbations in boiling water reactor (BWR) lattices, have been theoretically developed and tested. The validity of the derived theoretical expressions has been assessed for the case of a reactivity perturbation corresponding to the removal of an individual fuel pin from a nominal BWR assembly, thus effectively substituting the pin by water. Such reactivity effects are of importance in the context of evaluating advanced fuel element designs, e.g., those employing part-length rods. Two different geometry models have been implemented for the LWR-PROTEUS critical research facility [full core (FC) and a smaller, reduced geometry (RG)], using the light water reactor assembly code BOXER, and calculations have been performed for the nominal cases (all pins present in the central test assembly) and the perturbed cases (individual pins removed). The FC results have been compared with the results of the RG model with two different boundary conditions (reflective and critical albedo). The comparisons have shown that the results of critical albedo calculations feature superior representability. Differences in relative reactivity effects, with respect to results of the FC calculation, are found to be within the range ±1 to ±4%

  11. Neutron noise to detect BWR-4 in-core instrument tube vibrations and impacting

    International Nuclear Information System (INIS)

    Fry, D.N.; Kryter, R.C.; Mathis, M.V.; Mott, J.E.; Robinson, J.C.

    1979-01-01

    Noise analysis was used to investigate in-core instrument tube vibrations in boiling water reactor-4 (BWR-4)-type reactors. Neutron noise signals from in-core fission chambers and acoustic noise signals from externally mounted accelerometers were used in these studies. The results show that neutron noise can be used to detect vibration and, more important, impacting of instrument tubes against adjacent fuel channel boxes. Externally mounted accelerometers detect impacting but not rubbing of instrument tubes against fuel channel boxes. Accelerometers can monitor impacting only on the particular instrument tube where the accelerometer is mounted. Surveillance for instrument tube impacts can be accomplished using standard BWR-4 in-core power range neutron flux detectors at all instrument tube locations containing these detectors. Ex-vessel accelerometers can then be used to monitor instrument tubes that lack power range neutron flux detectors. However, noise on axial flux profiles obtained with movable in-core detectors is not a reliable indicator of impacting because the recorder used to plot the flux profiles does not respond adequately to the noise frequency generated by impacting. Neutron noise signatures show that modification of the bypass cooling by plugging preexisting holes and drilling new holes in the fuel assembly lower tie plate greatly reduces instrument tube vibration and eliminates impacting of instrument tubes against fuel boxes

  12. BWR stability: analysis of cladding temperature for high amplitude oscillations - 146

    International Nuclear Information System (INIS)

    Pohl, P.; Wehle, F.

    2010-01-01

    Power oscillations associated with density waves in boiling water reactors (BWRs) have been studied widely. Industrial research in this area is active since the invention of the first BWR. Stability measurements have been performed in various plants during commissioning phase but especially the magnitude and divergent nature of the oscillations during the LaSalle Unit 2 nuclear power plant event on March 9, 1988, renewed concern about the state of knowledge on BWR instabilities and possible consequences to fuel rod integrity. The objective of this paper is to present a simplified stability tool, applicable for stability analysis in the non-linear regime, which extends to high amplitude oscillations where inlet reverse flow occurs. In case of high amplitude oscillations a cyclical dryout and rewetting process at the fuel rod may take place, which leads in turn to rapid changes of the heat transfer from the fuel rod to the coolant. The application of this stability tool allows for a conservative determination of the fuel rod cladding temperature in case of high amplitude oscillations during the dryout / re-wet phase. Moreover, it reveals in good agreement to experimental findings the stabilizing effect of the reverse bundle inlet flow, which might be obtained for large oscillation amplitudes. (authors)

  13. TORT/MCNP coupling method for the calculation of neutron flux around a core of BWR

    International Nuclear Information System (INIS)

    Kurosawa, M.

    2005-01-01

    For the analysis of BWR neutronics performance, accurate data are required for neutron flux distribution over the In-Reactor Pressure Vessel equipments taking into account the detailed geometrical arrangement. The TORT code can calculate neutron flux around a core of BWR in a three-dimensional geometry model, but has difficulties in fine geometrical modelling and lacks huge computer resource. On the other hand, the MCNP code enables the calculation of the neutron flux with a detailed geometry model, but requires very long sampling time to give enough number of particles. Therefore, a TORT/MCNP coupling method has been developed to eliminate the two problems mentioned above in each code. In this method, the TORT code calculates angular flux distribution on the core surface and the MCNP code calculates neutron spectrum at the points of interest using the flux distribution. The coupling method will be used as the DOT-DOMINO-MORSE code system. This TORT/MCNP coupling method was applied to calculate the neutron flux at points where induced radioactivity data were measured for 54 Mn and 60 Co and the radioactivity calculations based on the neutron flux obtained from the above method were compared with the measured data. (authors)

  14. Validations of BWR nuclear design code using ABWR MOX numerical benchmark problems

    International Nuclear Information System (INIS)

    Takano, Shou; Sasagawa, Masaru; Yamana, Teppei; Ikehara, Tadashi; Yanagisawa, Naoki

    2017-01-01

    BWR core design code package (the HINES assembly code and the PANACH core simulator), being used for full MOX-ABWR core design, has been benchmarked against the high-fidelity numerical solutions as references, for the purpose of validating its capability of predicting the BWR core design parameters systematically from UO 2 to 100% MOX cores. The reference solutions were created by whole core critical calculations using MCNPs with the precisely modeled ABWR cores both in hot and cold conditions at BOC and EOC of the equilibrium cycle. A Doppler-Broadening Rejection Correction (DCRB) implemented MCNP5-1.4 with ENDF/B-VII.0 was mainly used to evaluate the core design parameters, except for effective delayed neutron fraction (β eff ) and prompt neutron lifetime (l) with MCNP6.1. The discrepancies in the results between the design codes HINES-PANACH and MCNPs for the core design parameters such as the bundle powers, hot pin powers, control rod worth, boron worth, void reactivity, Doppler reactivity, β eff and l, are almost within target accuracy, leading to the conclusion that HINES-PANACH has sufficient fidelity for application to full MOX-ABWR core design. (author)

  15. Basic evaluation on nuclear characteristics of BWR high burnup MOX fuel and core

    International Nuclear Information System (INIS)

    Nagano, M.; Sakurai, S.; Yamaguchi, H.

    1997-01-01

    MOX fuel will be used in existing commercial BWR cores as a part of reload fuels with equivalent operability, safety and economy to UO 2 fuel in Japan. The design concept should be compatible with UO 2 fuel design. High burnup UO 2 fuels are being developed and commercialized step by step. The MOX fuel planned to be introduced in around year 2000 will use the same hardware as UO 2 8 x 8 array fuel developed for a second step of UO 2 high burnup fuel. The target discharge exposure of this MOX fuel is about 33 GWd/t. And the loading fraction of MOX fuel is approximately one-third in an equilibrium core. On the other hand, it becomes necessary to minimize a number of MOX fuels and plants utilizing MOX fuel, mainly due to the fuel economy, handling cost and inspection cost in site. For the above reasons, it needed to developed a high burnup MOX fuel containing much Pu and a core with a large amount of MOX fuels. The purpose of this study is to evaluate basic nuclear fuel and core characteristics of BWR high burnup MOX fuel with batch average exposure of about 39.5 GWd/t using 9 x 9 array fuel. The loading fraction of MOX fuel in the core is within a range of about 50% to 100%. Also the influence of Pu isotopic composition fluctuations and Pu-241 decay upon nuclear characteristics are studied. (author). 3 refs, 5 figs, 3 tabs

  16. Containment venting as a mitigation technique for BWR MARK I plant ATWS

    International Nuclear Information System (INIS)

    Harrington, R.M.

    1987-01-01

    Containment venting is studied as a mitigation strategy for preventing or delaying severe fuel damage following hypothetical BWR Anticipated Transient Without Scram (ATWS) accidents initiated by MSIV-closure, and compounded by failure of the Standby Liquid Control (SLC) system injection of sodium pentaborate solution and by the failure of manually initiated control rod insertion. The venting of primary containment after reaching 75 psia (0.52 MPa) is found to result in the release of the vented steam inside the reactor building, and to result in inadequate Net Positive Suction Head (NPSH) for any system pumping from the pressure suppression pool. CONTAIN code calculations show that personnel access to large portions of the reactor building would be lost soon after the initiation of venting and that the temperatures reached would be likely to result in independent equipment failures. It is concluded that containment venting would be more likely to cause or to hasten the onset of severe fuel damage than to prevent or to delay it. Two alternative strategies that do not require containment venting, but that could delay or prevent severe fuel damage, are analyzed. BWR-LTAS code results are presented for a successful mitigation strategy in which the reactor vessel is depressurized, and for one in which the reactor vessel remains at pressure

  17. Leak before break analysis for cracking at multiple weld locations in BWR recirculation piping

    International Nuclear Information System (INIS)

    Zahoor, A.; Gamble, R.

    1984-01-01

    Periodically over the past decade, intergranular stress corrosion cracking (IGSCC) has been found in austenitic stainless steel piping at Boiling Water Reactor facilities. The effect of IGSCC on piping integrity has been evaluated previously in various BWR Owners Group and NRC studies. In these studies, the analyses were performed assuming the presence of a crack at a single weld location in the pipe run. The purpose of this investigation was to compare the leak rate and potential for unstable crack extension associated with a throughwall crack for the following two conditions in a BWR recirculation system: (1) the recirculation piping contains part through cracks at multiple weld locations and a single throughwall crack, and (2) the piping contains only a throughwall crack at one weld location. Two type BWRs were evaluated; namely, the ring header and five individual loop designs. The results from the analyses indicate that the potential for unstable crack extension at large bending loads, and leak rate at normal operation are not affected by the presence of part through cracks at multiple weld locations. The differences in the respective calculated L/sub eff/ and leak rates for the single and multiply cracked conditions are less than 2%

  18. A practical methodology of radiological protection for the reduction of hot particles in BWR type reactors

    International Nuclear Information System (INIS)

    Alvarez G, G.

    1991-01-01

    The purpose of this work, in general form, is to describe a practical method for reduction of hot particles generated as consequence of the operational activities of BWR nuclear reactors. This methodology provides a description of the localizations and/or probable activities of finding particles highly radioactive denominated hot particles. For this purpose it was developed a strategy based on the decontamination lineaments, as well as the manipulation, gathering, registration, contention, documentation, control and final disposition of the hot particles. In addition, some recommendations are reiterated and alternative, in order to gathering the hot particles in a dynamic way given to the activities of the personal occupationally exposed in highly radioactive areas. The structure of the methodology of hot particles is supported in the radiological controls based on the Code of Federal Regulation 10 CFR 20 as well as the applicable regulatory documents. It provides an idea based on administrative controls of radiological protection, in order to suggesting the responsibilities and necessary directing for the control of the hot particles required in nuclear plants of the BWR type. (author)

  19. Identification and initial assessment of candidate BWR late-phase in-vessel accident management strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.

    1991-01-01

    Work sponsored by the United States Nuclear Regulatory Commission (USNRC) to identify and perform preliminary assessments of candidate BWR [boiling water reactor] in-vessel accident management strategies was completed at Oak Ridge National Laboratory (ORNL) during fiscal year 1990. Mitigative strategies for containment events have been the subject of a companion study at Brookhaven National Laboratory. The focus of this Oak Ridge effort was the development of new strategies for mitigation of the late phase events, that is, the events that would occur in-vessel after the onset of significant core damage. The work began with an investigation of the current status of BWR in-vessel accident management procedures and proceeded through a preliminary evaluation of several candidate new strategies. The steps leading to the identification of the candidate strategies are described. The four new candidate late-phase (in-vessel) accident mitigation strategies identified by this study and discussed in the report are: (1) keep the reactor vessel depressurized; (2) restore injection in a controlled manner; (3) inject boron if control blade damage has occurred; and (4) containment flooding to maintain core and structural debris in-vessel. Additional assessments of these strategies are proposed

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

    KERENA is an innovative AREVA GmbH boiling water reactor (BWR) with passive safety systems (Generation III+). In order to verify the functionality of the reactor design an experimental validation program was executed. Therefore the INKA (Integral Teststand Karlstein) test facility was designed and erected. It is a mockup of the BWR containment, with integrated pressure suppression system. While the scaling of the passive components and the levels match the original values, the volume scaling of the containment compartments is approximately 1:24. The storage capacity of the test facility pressure vessel corresponds to approximately 1/6 of the KERENA RPV and is supplied by a benson boiler with a thermal power of 22 MW. In March 2013 the first integral test - Main Steam Line Break (MSLB) - was executed. The test measured the combined response of the passive safety systems to the postulated initiating event. The main goal was to demonstrate the ability of the passive systems to ensure core coverage, decay heat removal and to maintain the containment within defined limits. The results of the test showed that the passive safety systems are capable to bring the plant to stable conditions meeting all required safety targets with sufficient margins. Therefore the test verified the function of those components and the interplay between them. The test proved that INKA is an unique test facility, capable to perform integral tests of passive safety concepts under plant-like conditions. (orig.)

  1. Quantitative evaluation for training results of nuclear plant operator on BWR simulator

    International Nuclear Information System (INIS)

    Sato, Takao; Sato, Tatsuaki; Onishi, Hiroshi; Miyakita, Kohji; Mizuno, Toshiyuki

    1985-01-01

    Recently, the reliability of neclear power plants has largely risen, and the abnormal phenomena in the actual plants are rarely encountered. Therefore, the training using simulators becomes more and more important. In BWR Operator Training Center Corp., the training of the operators of BWR power plants has been continued for about ten years using a simulator having the nearly same function as the actual plants. The recent high capacity ratio of nuclear power plants has been mostly supported by excellent operators trained in this way. Taking the opportunity of the start of operation of No.2 simulator, effort has been exerted to quantitatively grasp the effect of training and to heighten the quality of training. The outline of seven training courses is shown. The technical ability required for operators, the items of quantifying the effect of training, that is, operational errors and the time required for operation, the method of quantifying, the method of collecting the data and the results of the application to the actual training are described. It was found that this method is suitable to quantify the effect of training. (Kako, I.)

  2. In-core failure of the instrumented BWR rod by locally induced high coolant temperature

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki

    1985-12-01

    In the BWR type light water loop instrumented in HBWR, a current BWR type fuel rod pre-irradiated up to 5.6 MWd/kgU was power ramped to 50 kW/m. During the ramp, the diameter of the rod was expanded significantly at the bottom end. The behaviour was different from which caused by pellet-cladding interaction (PCI). In the post-irradiation examination, the rod was found to be failed. In this paper, the cause of the failure was studied and obtained the followings. (1) The significant expansion of the rod diameter was attributed to marked oxidation of cladding outer diameter, appeared in the direction of 0 0 -180 0 degree with a shape of nodular. (2) The cladding in the place was softened by high coolant temperature. Coolant pressure, 7MPa intruded the cladding into inside chamfer void at pellet interface. (3) At the place of the significant oxidation, an instrumented transformer was existed and the coolant flow area was very little. The reduction of the coolant flow was enhanced by the bending of the cladding which was caused in pre-irradiation stage. They are considered to be a principal cause of local closure of coolant flow and resultant high temperature in the place. (author)

  3. First interim examination of defected BWR and PWR rods tested in unlimited air at 2290C

    International Nuclear Information System (INIS)

    Einziger, R.E.; Cook, J.A.

    1983-01-01

    A five-year whole rod test was initiated to investigate the long-term stability of spent fuel rods under a variety of possible dry storage conditions. Both PWR and BWR rods were included in the test. The first interim examination was conducted after three months of testing to determine if there was any degradation in those defected rods stored in an unlimited air atmosphere. Visual observations, diametral measurements and radiographic smears were used to assess the degree of cladding deformation and particulate dispersal. The PWR rod showed no measurable change from the pre-test condition. The two original artificial defects had not changed in appearance and there was no diametral growth of the cladding. One of the defects in BWR rod showed significant deformation. There was approximately 10% cladding strain at the defect site and a small axial crack had formed. The fuel in the defect did not appear to be friable. The second defect showed no visible change and no cladding strain. Following examination, the test was continued at 230 0 C. Another interim examination is planned during the summer of 1983. This paper discusses the details and meaning of the data from the first interim examination

  4. Comparison of metaheuristic optimization techniques for BWR fuel reloads pattern design

    International Nuclear Information System (INIS)

    François, Juan-Luis; Ortiz-Servin, Juan José; Martín-del-Campo, Cecilia; Castillo, Alejandro; Esquivel-Estrada, Jaime

    2013-01-01

    Highlights: ► This paper shows a performance comparison of several optimization techniques for fuel reload in BWR. ► Genetic Algorithms, Neural Networks, Tabu Search and several Ant Algorithms were used. ► All optimization techniques were executed under same conditions: objective function and an equilibrium cycle. ► Fuel bundles with minor actinides were loaded into the core. ► Tabu search and Ant System were the best optimization technique for the studied problem. -- Abstract: Fuel reload pattern optimization is a crucial fuel management activity in nuclear power reactors. Along the years, a lot of work has been done in this area. In particular, several metaheuristic optimization techniques have been applied with good results for boiling water reactors (BWRs). In this paper, a comparison of different metaheuristics: genetic algorithms, tabu search, recurrent neural networks and several ant colony optimization techniques, were applied, in order to evaluate their performance. The optimization of an equilibrium core of a BWR, loaded with mixed oxide fuel composed of plutonium and minor actinides, was selected to be optimized. Results show that the best average values are obtained with the recurrent neural networks technique, meanwhile the best fuel reload was obtained with tabu search. However, according to the number of objective functions evaluated, the two fastest optimization techniques are tabu search and Ant System.

  5. Control rod pattern exchange in a BWR/6 utilizing gang mode withdrawal

    International Nuclear Information System (INIS)

    Auvil, A.B. Jr.; Aldemir, T.; Hajek, B.K.

    1986-01-01

    The use of checkerboard pattern of alternating inserted and fully withdrawn control rods and the uneven void distribution in boiling water reactor (BWR) cores can cause large burnup gradients even after a short time of operation. To compensate for these effects, power has to be reshaped periodically (typically every two full-power months) by individually manipulating the control rods. During this manipulation process (called the control rod pattern exchange), the core power is reduced to 60% of nominal power by means of flow reduction to limit power swings to tolerable levels and to ensure that fuel thermal limits are not exceeded. A control rod pattern exchange by individual rod manipulation typically takes 4 to 8 h and represents a large cost burden to the utility in terms of reduced system output. The latest generation of BWRs, the BWR/6, possesses the capability to simultaneously move up to four symmetrically located control rods. The rods corresponding to a given gang may have rotational symmetry, mirror symmetry, or a combination of the two. This paper presents a pattern exchange procedure that exploits the capability of gang mode rod withdrawal to reduce the pattern exchange execution time and radial power distribution asymmetry associated with individual rod manipulation. The working model used in the study is the Perry Nuclear Power Plant Unit 1, located in Perry, Ohio, and owned by the Cleveland Electric Illuminating Company

  6. Ultrasonic phased array examination of circumferential weld joint in reactor pressure vessel of BWR

    International Nuclear Information System (INIS)

    Nanekar, Paritosh; Jothilakshmi, N.; Jayakumar, T.

    2013-01-01

    Highlights: • Phased array technique developed for weld joint inspection in BWR pressure vessel. • Simulation studies were carried out for conventional and phased array probe. • Conventional ultrasonic test shows in-adequate weld coverage and poor resolution. • Focused sound beam in phased array results in good resolution and sensitivity. • Ultrasonic phased array technique is validated on mock-up with reference defects. - Abstract: The weld joints in the reactor pressure vessel (RPV) of Boiling Water Reactors (BWR) are required to be examined periodically for assurance of structural integrity. Ultrasonic phased array examination technique has been developed in authors’ laboratory for inspection of the top flange to shell circumferential weld joint in RPV of BWRs, which are in operation in India since the late 1960s. The development involved detailed simulation studies for computation of focal laws followed by validation on mock-up. The paper brings out the limitations of the conventional ultrasonic technique and how this can be overcome by the phased array approach for the weld joint under consideration. The phased array technique was successfully employed for field examination of this weld joint in RPV during the re-fuelling outage

  7. Experience using individually supplied heater rods in critical power testing of advanced BWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Majed, M.; Morback, G.; Wiman, P. [ABB Atom AB, Vasteras (Sweden)] [and others

    1995-09-01

    The ABB Atom FRIGG loop located in Vasteras Sweden has during the last six years given a large experience of critical power measurements for BWR fuel designs using indirectly heated rods with individual power supply. The loop was built in the sixties and designed for maximum 100 bar pressure. Testing up to the mid eighties was performed with directly heated rods using a 9 MW, 80 kA power supply. Providing test data to develop critical power correlations for BWR fuel assemblies requires testing with many radial power distributions over the full range of hydraulic conditions. Indirectly heated rods give large advantages for the testing procedure, particularly convenient for variation of individual rod power. A test method being used at Stern Laboratories (formerly Westinghouse Canada) since the early sixties, allows one fuel assembly to simulate all required radial power distributions. This technique requires reliable indirectly heated rods with independently controlled power supplies and uses insulated electric fuel rod simulators with built-in instrumentation. The FRIGG loop was adapted to this system in 1987. A 4MW power supply with 10 individual units was then installed, and has since been used for testing 24 and 25 rod bundles simulating one subbundle of SVEA-96/100 type fuel assemblies. The experience with the system is very good, as being presented, and it is selected also for a planned upgrading of the facility to 15 MW.

  8. TORT/MCNP coupling method for the calculation of neutron flux around a core of BWR.

    Science.gov (United States)

    Kurosawa, Masahiko

    2005-01-01

    For the analysis of BWR neutronics performance, accurate data are required for neutron flux distribution over the In-Reactor Pressure Vessel equipments taking into account the detailed geometrical arrangement. The TORT code can calculate neutron flux around a core of BWR in a three-dimensional geometry model, but has difficulties in fine geometrical modelling and lacks huge computer resource. On the other hand, the MCNP code enables the calculation of the neutron flux with a detailed geometry model, but requires very long sampling time to give enough number of particles. Therefore, a TORT/MCNP coupling method has been developed to eliminate the two problems mentioned above in each code. In this method, the TORT code calculates angular flux distribution on the core surface and the MCNP code calculates neutron spectrum at the points of interest using the flux distribution. The coupling method will be used as the DOT-DOMINO-MORSE code system. This TORT/MCNP coupling method was applied to calculate the neutron flux at points where induced radioactivity data were measured for 54Mn and 60Co and the radioactivity calculations based on the neutron flux obtained from the above method were compared with the measured data.

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

    International Nuclear Information System (INIS)

    Behringer, K.; Hennig, D.

    2002-11-01

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

  10. Study on vertical seismic response model of BWR-type reactor building

    International Nuclear Information System (INIS)

    Konno, T.; Motohashi, S.; Izumi, M.; Iizuka, S.

    1993-01-01

    A study on advanced seismic design for LWR has been carried out by the Nuclear Power Engineering Corporation (NUPEC), under the sponsorship of the Ministry of International Trade and Industry (MITI) of Japan. As a part of the study, it has been investigated to construct an accurate analytical model of reactor buildings for a seismic response analysis, which can reasonably represent dynamic characteristics of the building. In Japan, vibration models of reactor buildings for horizontal ground motion have been studied and examined through many simulation analyses for forced vibration tests and earthquake observations of actual buildings. And now it is possible to establish a reliable horizontal vibration model on the basis of multi-lumped mass and spring model. However, vertical vibration models have not been so much studied as horizontal models, due to less observed data for vertical motions. In this paper, the vertical seismic response models of a BWR-type reactor building including soil-structure interaction effect are numerically studied, by comparing the dynamic characteristics of (1) three dimensional finite element model, (2) multi-stick lumped mass model with a flexible base-mat, (3) multi-stick lumped mass model with a rigid base-mat and (4) single-stick lumped mass model. In particular, the BWR-type reactor building has the long span truss roof which is considered to be one of the critical members to vertical excitation. The modelings of the roof trusses are also studied

  11. A short core study for the consideration of next century BWR

    International Nuclear Information System (INIS)

    Tanabe, A.; Nakajima, Y.; Chuman, K.; Okada, A.; Ogata, T.; Mizumachi, W.; Yoshioka, R.; Takigawa, Y.; Omoto, A.

    1993-01-01

    The height of BWR core has been kept constant value of 3.7 m. A short core has a lot of merits, the 3 meter core of 50 kW/l with forced circulation is studied and evaluated for the consideration of the next century BWR. The pressure drop of the short core decreases and the core and channel stability are much improved and also increases thermal margin. It is possible to decrease rated flow by 10% and burnup reactivity can be compensated using wide adjustable flow window at rated power for internal pump plant. And if the short core is applied to the jetpump plant, the feedwater jetpump system becomes possible, if the high M, N ratio jetpump is developed. The short core needs larger diameter RPV by 10% and also increases number of fuel bundle and CR/CRD but the large size bundle such as twice pitch lattice will decreases bundle number by 70%. The short core makes RPV position lower and decreases the height of reactor building and is good effects on seismic design

  12. Ionization chamber with build-up cup spectral sensitivity to megavoltage (0.5-20 MeV) photon fluences in free air

    International Nuclear Information System (INIS)

    Gorlachev, G.E.

    2002-01-01

    In-air measurements of photon beam properties, used in radiation therapy, is common practice for determining radiation output dependence from the field size, known as head scatter factors (HSF). PMMA and brass build-up caps are most popular miniphantoms for providing electron equilibrium. Discrepancies up to 2% in HSF measurements by different combinations of detectors and equilibrium caps have been published. One of the main reasons of those discrepancies is the detector system spectral sensitivity and differences in primary and scatter radiation spectra. In the light of new model based dose calculation methods direct radiation fluence measurement is of great interest. So, understanding of detector spectral sensitivity is important task for modern dosimetry of radiation therapy. In the present study Monte Carlo (MC) method was employed to calculate ionization chamber response to monoenergetic photon fluences, normalized to water kerma units. Simulation was done using EGS4 package. Electron transport was performed with ESTEPE equal to 4%. PEGS cross sections were generated for maximal energy 20 MeV with cutoff kinetic energy 10 KeV both for photons and electrons. Scanditronix RK-05 ionization chamber was chosen as a prototype. Eight cylindrical miniphantoms, representing four materials (PMMA, Al, Cu, Pb) and two front wall thickness, were simulated. Results are presented. Miniphantom front wall thicknesses in each case are shown in the figure. Diameter depends on the material and equal respectively: PMMA - 4, Al - 2.5, Cu - 1.5, and PB - 1.5 cm. Ionization chamber outer diameter is equal to 0.7 cm. Detector sensitivity has considerable energy dependence. Two effects explain it. First is the radiation attenuation in the miniphantom. Second is pair production, which dominates in high atomic number miniphantoms for energies above 5 MeV. Depending on the miniphantom material detector response changes from 1.5 to 5 times in the energy range from 0.5 to 20 MeV. Correct

  13. Results of the Simulator smart against synthetic signals using a model of reduced order of BWR with additive and multiplicative noise; Resultados del simulador smart frente a senales sinteticas utilizando un modelo de orden reducido de BWR con ruido aditivo y multiplicativo

    Energy Technology Data Exchange (ETDEWEB)

    Munoz-Cobo, J. L.; Montesino, M. E.; Pena, J.; Escriva, A.; Melara, J.

    2011-07-01

    Results of SMART-simulator front of synthetic signals with models of reduced order of BWR with additive and multiplicative noise Under the SMART project, which aims to monitor the signals Cofrentes nuclear plant, we have developed a signal generator of synthetics BWR that will allow together real signals of plant the validation of the monitor.

  14. Removing organics with the GEODE ultraviolet/ozone system [removal of organic materials from BWR process water

    International Nuclear Information System (INIS)

    Head, R.A.; Alexander, J.E.; Lezon, R.J.

    1989-01-01

    The evaluation of an ultraviolet/ozone oxidation process for removing organic materials from BWR process water systems is described. The process has been found to remove over 90 per cent of organic carbon, at an estimated annual operating cost of around $21 000. (author)

  15. Simulation of the radiation fields from ionizing radiation sources inside the containment in an accident

    Science.gov (United States)

    Kalugin, M. A.

    2010-12-01

    In the present work, a set of codes used for simulations of the radiation fields from ionizing radiation sources inside the containment in an accident is described. A method of evaluating the gamma dose rate from a space and energy distributed source is given. The dose rate is calculated by means of the design point kernel method and using buildup factors. The code MCU-REA with the ORIMCU module is used for the burnup calculations.

  16. Orientation-dependent proton double-quantum NMR build-up function for soft materials with anisotropic mobility.

    Science.gov (United States)

    Naumova, Anna; Tschierske, Carsten; Saalwächter, Kay

    In recent years, the analysis of proton double-quantum NMR build-up curves has become an important tool to quantify anisotropic mobility in different kinds of soft materials such as polymer networks or liquid crystals. In the former case, such data provides a measure of orientation-dependent residual (time-averaged) dipolar couplings arising from anisotropic segmental motions, informing about the length and the state of local stretching of the network chains. Previous studies of macroscopically ordered, i.e. stretched, networks were subject to the limitation that a detailed build-up curve analysis on the basis of a universal "Abragam-like" (A-l) build-up function valid for a proton multi-spin system was only possible for an isotropic orientation-averaged response. This situation is here remedied by introducing a generic orientation-dependent build-up function for an anisotropically mobile protonated molecular segment. We discuss an application to the modeling of data for a stretched network measured at different orientations with respect to the magnetic field, and present a validation by fitting data of different liquid-crystal molecules oriented in the magnetic field. Copyright © 2017 Elsevier Inc. All rights reserved.

  17. The build-up dynamic and chemical fractionation of Cu, Zn and Cd in road-deposited sediment.

    Science.gov (United States)

    Zhang, Jin; Hua, Pei; Krebs, Peter

    2015-11-01

    This study investigates the build-up dynamics of heavy metals on impervious urban surfaces with different antecedent dry-weather periods (ADPs) and land-use types. Solid-phase concentration (mg/kg), surface load (mg/m(2)), and chemical fractionation of Zn, Cu and Cd in bulk- and size-fractionated road-deposited sediment were determined. The inherent correlations among particle size distribution, ADP, land use, and chemical fractionation were analysed by hierarchical cluster analysis. Results show a clear build-up phenomenon of Cu and Zn at a city commercial centre and a highway area. Cd had complex build-up patterns. With regard to chemical fractionation, Zn and Cd could pose higher risks than Cu to aquatic biota after a longer ADP. Special attention should be paid to the significant risk in a rural area because of the high proportion of exchangeable chemical fractionation in terms of the unstable Cd component. Hierarchical cluster analysis indicates that ADP had a strong influence on build-up processes of sediments and associated metals. The metal contents were very dependent on the particle size distribution. However, the chemical fractionation of metals was dependent to a lesser extent on the land-use type. Copyright © 2015 Elsevier B.V. All rights reserved.

  18. 76 FR 7238 - Pipeline Safety: Dangers of Abnormal Snow and Ice Build-Up on Gas Distribution Systems

    Science.gov (United States)

    2011-02-09

    ... of moisture in equipment and snow or ice blocking regulator or relief valve vents which could prevent... DEPARTMENT OF TRANSPORTATION Pipeline and Hazardous Materials Safety Administration [Docket No. PHMSA-2011-0028] Pipeline Safety: Dangers of Abnormal Snow and Ice Build-Up on Gas Distribution Systems...

  19. Inherent errors in pollutant build-up estimation in considering urban land use as a lumped parameter.

    Science.gov (United States)

    Liu, An; Goonetilleke, Ashantha; Egodawatta, Prasanna

    2012-01-01

    Stormwater quality modeling results are subject to uncertainty. The variability of input parameters is an important source of overall model error. An in-depth understanding of the variability associated with input parameters can provide knowledge on the uncertainty associated with these parameters and can assist in uncertainty analysis of stormwater quality models and decision making based on modeling outcomes. This paper discusses the outcomes of a research study undertaken to analyze the variability related to pollutant build-up parameters in stormwater quality modeling. The study was based on the analysis of pollutant build-up samples collected from 12 road surfaces in residential, commercial, and industrial land uses. It was found that build-up characteristics vary appreciably even within the same land use. Therefore, using land use as a lumped parameter would contribute significant uncertainties in stormwater quality modeling. Additionally, it was found that the variability in pollutant build-up can be significant depending on the pollutant type. This underlines the importance of taking into account specific land use characteristics and targeted pollutant species when undertaking uncertainty analysis of stormwater quality models or in interpreting the modeling outcomes. Copyright © by the American Society of Agronomy, Crop Science Society of America, and Soil Science Society of America, Inc.

  20. Extremely rare collapse and build-up of turbulence in stochastic models of transitional wall flows

    Science.gov (United States)

    Rolland, Joran

    2018-02-01

    This paper presents a numerical and theoretical study of multistability in two stochastic models of transitional wall flows. An algorithm dedicated to the computation of rare events is adapted on these two stochastic models. The main focus is placed on a stochastic partial differential equation model proposed by Barkley. Three types of events are computed in a systematic and reproducible manner: (i) the collapse of isolated puffs and domains initially containing their steady turbulent fraction; (ii) the puff splitting; (iii) the build-up of turbulence from the laminar base flow under a noise perturbation of vanishing variance. For build-up events, an extreme realization of the vanishing variance noise pushes the state from the laminar base flow to the most probable germ of turbulence which in turn develops into a full blown puff. For collapse events, the Reynolds number and length ranges of the two regimes of collapse of laminar-turbulent pipes, independent collapse or global collapse of puffs, is determined. The mean first passage time before each event is then systematically computed as a function of the Reynolds number r and pipe length L in the laminar-turbulent coexistence range of Reynolds number. In the case of isolated puffs, the faster-than-linear growth with Reynolds number of the logarithm of mean first passage time T before collapse is separated in two. One finds that ln(T ) =Apr -Bp , with Ap and Bp positive. Moreover, Ap and Bp are affine in the spatial integral of turbulence intensity of the puff, with the same slope. In the case of pipes initially containing the steady turbulent fraction, the length L and Reynolds number r dependence of the mean first passage time T before collapse is also separated. The author finds that T ≍exp[L (A r -B )] with A and B positive. The length and Reynolds number dependence of T are then discussed in view of the large deviations theoretical approaches of the study of mean first passage times and multistability

  1. The build-up of polyelectrolyte multilayers of microfibrillated cellulose and cationic polyelectrolytes.

    Science.gov (United States)

    Wågberg, Lars; Decher, Gero; Norgren, Magnus; Lindström, Tom; Ankerfors, Mikael; Axnäs, Karl

    2008-02-05

    A new type of nanocellulosic material has been prepared by high-pressure homogenization of carboxymethylated cellulose fibers followed by ultrasonication and centrifugation. This material had a cylindrical cross-section as shown by transmission electron microscopy with a diameter of 5-15 nm and a length of up to 1 microm. Calculations, using the Poisson-Boltzmann equation, showed that the surface potential was between 200 and 250 mV, depending on the pH, the salt concentration, and the size of the fibrils. They also showed that the carboxyl groups on the surface of the nanofibrils are not fully dissociated until the pH has reached pH = approximately 10 in deionized water. Calculations of the interaction between the fibrils using the Derjaguin-Landau-Verwey-Overbeek theory and assuming a cylindrical geometry indicated that there is a large electrostatic repulsion between these fibrils, provided the carboxyl groups are dissociated. If the pH is too low and/or the salt concentration is too high, there will be a large attraction between the fibrils, leading to a rapid aggregation of the fibrils. It is also possible to form polyelectrolyte multilayers (PEMs) by combining different types of polyelectrolytes and microfibrillated cellulose (MFC). In this study, silicon oxide surfaces were first treated with cationic polyelectrolytes before the surfaces were exposed to MFC. The build-up of the layers was monitored with ellipsometry, and they show that it is possible to form very well-defined layers by combinations of MFC and different types of polyelectrolytes and different ionic strengths of the solutions during the adsorption of the polyelectrolyte. A polyelectrolyte with a three-dimensional structure leads to the build-up of thick layers of MFC, whereas the use of a highly charged linear polyelectrolyte leads to the formation of thinner layers of MFC. An increase in the salt concentration during the adsorption of the polyelectrolyte results in the formation of thicker

  2. Generic Aging Management Programs for License Renewal of BWR Reactor Coolant System Components

    International Nuclear Information System (INIS)

    Shah, V.N.; Liu, Y.Y.

    2002-01-01

    The paper reviews the existing generic aging management programs (AMPs) for the reactor coolant system (RCS) components in boiling water reactors (BWRs), including the reactor pressure vessel and internals, the reactor recirculation system, and the connected piping. These programs have been evaluated in the U.S. Nuclear Regulatory Commission (NRC) report, Generic Aging Lessons Learned (GALL), NUREG-1801, for their use in the license renewal process to manage several aging effects, including loss of material, crack initiation and growth, loss of fracture toughness, loss of preload, wall thinning, and cumulative fatigue damage. The program evaluation includes a review of ten attributes (scope of program, preventive actions, parameters monitored/inspected, detection of aging effects, monitoring and trending, acceptance criteria, corrective actions, confirmative process, administrative control, and operating experience) for their effectiveness in managing a specific aging effect in a given component(s). The generic programs are based on the ASME Section XI inservice inspection requirements; industry guidelines for inspection and evaluation of aging effects in BWR reactor vessel, internals, and recirculation piping; monitoring and control of BWR water chemistry; and operating experience as reported in the USNRC generic communications and industry reports. The review concludes that all generic AMPs are acceptable for managing aging effects in BWR RCS components during an extended period of operation and do not need further evaluation. However, the plant-specific programs for managing aging in certain RCS components during an extended period of operation do require further evaluation. For some plant-specific AMPs, the GALL report recommends an aging management activity to verify their effectiveness. An example of such an activity is a one-time inspection of Class 1 small-bore piping to ensure that service-induced weld cracking is not occurring in the piping. Several of

  3. Generic aging management programs for license renewal of BWR reactor coolant systems components

    International Nuclear Information System (INIS)

    Shah, V.N.; Liu, Y.Y.

    2002-01-01

    The paper reviews the existing generic aging management programs (AMPs) for the reactor coolant system (RCS) components in boiling water reactors (BWRs), including the reactor pressure vessel and internals, the reactor recirculation system, and the connected piping. These programs have been evaluated in the U.S. Nuclear Regulatory Commission (NRC) report, Generic Aging Lessons Learned (GALL), NUREG-1801, for their use in the license renewal process to manage several aging effects, including loss of material, crack initiation and growth, loss of fracture toughness, loss of preload, wall thinning, and cumulative fatigue damage. The program evaluation includes a review of ten attributes (scope of program, preventive actions, parameters monitored/inspected, detection of aging effects, monitoring and trending, acceptance criteria, corrective actions, confirmative process, administrative control, and operating experience) for their effectiveness in managing a specific aging effect in a given component(s). The generic programs are based on the ASME Section XI inservice inspection requirements; industry guidelines for inspection and evaluation of aging effects in BWR reactor vessel, internals, and recirculation piping; monitoring and control of BWR water chemistry; and operating experience as reported in the USNRC generic communications and industry reports. The review concludes that all generic AMPs are acceptable for managing aging effects in BWR RCS components during an extended period of operation and do not need further evaluation. However, the plant-specific programs for managing aging in certain RCS components during an extended period of operation do require further evaluation. For some plant-specific AMPs, the GALL report recommends an aging management activity to verify their effectiveness. An example of such an activity is a one-time inspection of Class 1 small-bore piping to ensure that service-induced weld cracking is not occurring in the piping. Several of

  4. Experimental study on thermo-hydraulic instability on reduced-moderation natural circulation BWR concept

    International Nuclear Information System (INIS)

    Watanabe, Noriyuki; Subki, M.H.; Kikura, Hiroshige; Aritomi, Masanori

    2003-01-01

    Reduced-moderation natural circulation BWR has been promoted to solve the recent challenges in BWR nuclear power technology problems as one of advanced small and medium-sized reactors equipped with the passive safety features in conformity with the natural law. However, the elimination of recirculation pumps and a high-density core due to the increase of conversion ratio could cause various thermo-hydraulic instabilities especially during the start-up stage. The occurrences of the thermo-hydraulic instabilities are not desirable and it is one of the main challenges in establishing reduced-moderation natural circulation BWR as a commercial reactor. The purpose of this present study is to experimentally investigate the driving mechanism of the thermo-hydraulic instabilities and the effect of system pressure on the unstable flow patterns. Hence, as the fundamental research for this study, a natural circulation loop that carries boiling fluid with parallel boiling channel has been constructed. Channel gap that has been set at 2 mm in order to simulate reduced-moderation reactor core. Pressure ranges of 0.1 up to 0.7 MPa, input heat flux range of 0 ou to 577 kW/m 2 , and inlet subcooling temperatures of 5, 10, and 15 K respectively, are imposed in the experiments. This experiment clarifies that changes in unstable flow patterns with increase in heat flux can be classified into two in response to system pressure range. In case of atmospheric pressure, unstable flow patters has been classified in beyond order, (1) in-phase geysering, (2) transition oscillation combined with both features of in-phase geysering and natural circulation oscillation, (3) natural circulation oscillation induced by hydrostatic head fluctuation, (4) density wave oscillation, and finally (5) stable boiling two-phase flow. On the other hand, in the system pressure range from 0.2 to 0.7 MPa, unstable patters have been dramatically changed in the following order (1) out-of-phase geysering, (2

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz S, J.J

    1998-10-01

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

  9. Assessment of potential nutrient build-up around beef cattle production areas using electromagnetic induction.

    Science.gov (United States)

    Cordeiro, Marcos R C; Ranjan, Ramanathan Sri; Cicek, Nazim

    2011-12-01

    Electromagnetic induction (EMI) has been used to map soil properties such as salinity and water content. The objective of this research is to use EMI to map the potential distribution of nutrients around beef cattle pens and to relate this distribution to major physiographic field features. Beef cattle farms in different physiographic locations were surveyed in Manitoba, Canada, using an EM-38 conductivity meter georeferenced with a GPS receiver. Samples were collected using a response surface design and analysed for electrical conductivity (ECe), which was used as a proxy for determining potential build-up of nutrients. Multiple linear regression models (MLR) were used for calibration of the EM readings. The results showed that areas 1 through 4 had ECe soil layer to accumulate the nutrients. Micro-depressions played a major role in salt accumulation, with the depressions corresponding to higher values of ECe. The presence of features such as drainage ditches and compacted soils beneath roads strongly affected the direction of the plumes. Based on these results, the location of the pens on high elevations and the provision to collect the run-off from the pens were identified as good design criteria. Highly permeable soils may require a low permeability liner to capture the deep percolation and redirect it towards a collection area.

  10. Prediction of wax buildup in 24 inch cold, deep sea oil loading line

    Energy Technology Data Exchange (ETDEWEB)

    Asperger, R.G.; Sattler, R.E.; Tolonen, W.J.; Pitchford, A.C.

    1981-10-01

    When designing pipelines for cold environments, it is important to know how to predict potential problems due to wax deposition on the pipeline's inner surface. The goal of this work was to determine the rate of wax buildup and the maximum, equlibrium wax thickness for a North Sea field loading line. The experimental techniques and results used to evaluate the waxing potential of the crude oil (B) are described. Also, the theoretic model which was used for predicting the maximum wax deposit thickness in the crude oil (B) loading pipeline at controlled temperatures of 40 F (4.4 C) and 100 F (38 C), is illustrated. Included is a recommendation of a procedure for using hot oil at the end of a tanker loading period in order to dewax the crude oil (B) line. This technique would give maximum heating of the pipeline and should be followed by shutting the hot oil into the pipeline at the end of the loading cycle which will provide a hot oil soaking to help soften existing wax. 14 references.

  11. Comparison of mechanical properties of five commercial dental core build-up materials.

    Science.gov (United States)

    Passos, Sheila P; Freitas, Anderson P; Jumaily, Sami; Santos, Maria Jacinta M C; Rizkalla, Amin S; Santos, Gildo C

    2013-01-01

    This study aimed to evaluate and compare the mechanical properties of five commercial core materials using fracture toughness (FT), Knoop hardness number (KHN), diametral tensile strength (DTS), and dynamic elastic moduli (DEM). Composite material specimens were produced (Rock Core, CosmeCore, ParaCore, MultiCore Flow, and Filtek Supreme Plus). The FT test (n = 15) was performed using notchless triangular prism (NTP) specimens. FT was determined using an Instron testing machine. KHN (n = 3) was evaluated using three indentations applied on each specimen. DTS test (n = 15) was measured using an Instron testing machine. The density. of the specimens (n = 3) was determined by water displacement method. Dynamic Young's, shear moduli, and Poisson's ratio (n = 3) were measured by an ultrasonic method. Statistical analysis was conducted using ANOVA and a Tukey B rank order test (P = 0.05). Rock Core presented the lowest FT values. Filtek Supreme Plus and CosmeCore exhibited significantly higher KHN values than the rest of the materials. CosmeCore had the highest DTS value, which was statistically significant only compared to Rock Core. For DEM, Filtek Supreme Plus exhibited significantly higher Young's and shear moduli than the rest of the materials (P < 0.05). Results demonstrated significant differences in the FT, KHN, and DTS values of the core build-up materials tested. According to the elastic behavior of the core composite materials, Rock Core had the lowest Young's values.

  12. Kick-Off Point (KOP and End of Buildup (EOB Data Analysis in Trajectory Design

    Directory of Open Access Journals (Sweden)

    Novrianti Novrianti

    2017-06-01

    Full Text Available Well X is a development well which is directionally drilled. Directional drilling is choosen because the coordinate target of Well X is above the buffer zone. The directional track plan needs accurate survey calculation in order to make the righ track for directional drilling. There are many survey calculation in directional drilling such as tangential, underbalance, average angle, radius of curvature, and mercury method. Minimum curvature method is used in this directional track plan calculation. This method is used because it gives less error than other method.  Kick-Off Point (KOP and End of Buildup (EOB analysis is done at 200 ft, 400 ft, and 600 ft depth to determine the trajectory design and optimal inclination. The hole problem is also determined in this trajectory track design. Optimal trajectory design determined at 200 ft depth because the inclination below 35º and also already reach the target quite well at 1632.28 ft TVD and 408.16 AHD. The optimal inclination at 200 ft KOP depth because the maximum inclination is 18.87º which is below 35º. Hole problem will occur if the trajectory designed at 600 ft. The problems are stuck pipe and the casing or tubing will not able to bend.

  13. The Build-Up Course of Visuo-Motor and Audio-Motor Temporal Recalibration

    Directory of Open Access Journals (Sweden)

    Yoshimori Sugano

    2011-10-01

    Full Text Available The sensorimotor timing is recalibrated after a brief exposure to a delayed feedback of voluntary actions (temporal recalibration effect: TRE (Heron et al., 2009; Stetson et al., 2006; Sugano et al., 2010. We introduce a new paradigm, namely ‘synchronous tapping’ (ST which allows us to investigate how the TRE builds up during adaptation. In each experimental trial, participants were repeatedly exposed to a constant lag (∼150 ms between their voluntary action (pressing a mouse and a feedback stimulus (a visual flash / an auditory click 10 times. Immediately after that, they performed a ST task with the same stimulus as a pace signal (7 flashes / clicks. A subjective ‘no-delay condition’ (∼50 ms served as control. The TRE manifested itself as a change in the tap-stimulus asynchrony that compensated the exposed lag (eg, after lag adaptation, the tap preceded the stimulus more than in control and built up quickly (∼3–6 trials, ∼23–45 sec in both the visuo- and audio-motor domain. The audio-motor TRE was bigger and built-up faster than the visuo-motor one. To conclude, the TRE is comparable between visuo- and audio-motor domain, though they are slightly different in size and build-up rate.

  14. Correction of build-up factor one x-ray hvl measurement

    International Nuclear Information System (INIS)

    Yuliati, Helfi; Akhadi, Mukhlis

    2000-01-01

    Research to obtain the value build-up factor (b) on half value layers (HVL) measurement of diagnostic X-Rays using pocket dosimeter behind aluminium (AI) filter with its thickness vary from 1 to 4 mm. From the measurement it was obtained HVL value of 1.997, 2.596 and 2.718 mmAI for X-Rays of kVp : 80 Kv with 1, 2, 3 and 4 mm filter thickness respectively. HVL value significantly increase with increasing AI filter thickness. Increasing of HVL means increasing filter thickness. From the calculation it was obtained increasing b value relative to 1 mm AI filter of 18.26 and 46% for filter thickness of 2, 3 and 4 mm respectively. Experiment result shows the need of involving b value in HVL calculation of X-Rays if the filter is relatively thick. Calculation of HVL of X-Rays can be carried out with thin layers filter. Key words : x-rays, half value layer, build up factor

  15. Comparative study of mechanical properties of direct core build-up materials

    Directory of Open Access Journals (Sweden)

    Girish Kumar

    2015-01-01

    Full Text Available Background and Objectives: The strength greatly influences the selection of core material because core must withstand forces due to mastication and para-function for many years. This study was conducted to evaluate certain mechanical properties of commonly used materials for direct core build-up, including visible light cured composite, polyacid modified composite, resin modified glass ionomer, high copper amalgam, and silver cermet cement. Materials and Methods: All the materials were manipulated according to the manufacturer′s recommendations and standard test specimens were prepared. A universal testing machine at different cross-head speed was used to determine all the four mechanical properties. Mean compressive strength, diametral tensile strength, flexural strength, and elastic modulus with standard deviations were calculated. Multiple comparisons of the materials were also done. Results: Considerable differences in compressive strength, diametral tensile strength, and flexural strength were observed. Visible light cured composite showed relatively high compressive strength, diametral tensile strength, and flexural strength compared with the other tested materials. Amalgam showed the highest value for elastic modulus. Silver cermet showed less value for all the properties except for elastic modulus. Conclusions: Strength is one of the most important criteria for selection of a core material. Stronger materials better resist deformation and fracture provide more equitable stress distribution, greater stability, and greater probability of clinical success.

  16. Comparison between electron cloud build-up measurements and simulations at the CERN PS

    CERN Document Server

    Caspers, F; Mahner, E; Rumolo, G; Yin Vallgren, C; Iadarola, G

    2012-01-01

    The build up of an Electron Cloud (EC) has been observed at the CERN Proton Synchrotron (PS) during the last stages of the LHC high intensity beam preparation, especially after the bunch shortening before extraction. Two dedicated EC experiments, both equipped with two button pick-ups, a pressure gauge, a clearing electrode and a small dipole magnet, are available in two straight sections of the machine. A measurement campaign has been carried out, in order to scan the EC build-up of LHC-type beams with different bunch spacing, bunch intensity and bunch length. Such information, combined with the results from build up simulations, is of relevance for the characterization in terms of Secondary Emission Yield (SEY) of the chamber inner surface. The interest is twofold: this will enable us to predict the EC build up distribution in the PS for higher intensity beams in the frame of the upgrade program, and it will provide validation of the EC simulation models and codes.

  17. Determination of attenuation parameters and energy absorption build-up factor of amine group materials

    Science.gov (United States)

    Lokhande, Rajkumar M.; More, Chaitali V.; Surung, Bharat S.; Pawar, Pravina P.

    2017-12-01

    We have computed radiological parameters of some C- H- N- O based amine group bio material in the energy range 122-1330 keV with the gamma ray count by narrow beam geometry. The NaI(Tl) detector with 8 K multichannel analyser was used having resolution 6.8% at 663 keV. The energy absorption buildup factor (EABF) was determined by using Geometric Progression (G-P) fitting method up to penetration depth of 40 mfp at energy 0.015-15 MeV. The NIST XCOM data were compared with the experimental value and we observed (3-5%) difference. The comparative study of effective atomic number and effective electron density in the energy range 122-1330 keV using Gaussian fit for accuracy were performed. The amino acid has the highest EABF value at 0.1 MeV and the variation in EABF with penetration depth up to 1-40 mean free path (mfp). The calculated radiological data of biological material are applicable in medical physics and dosimetry.

  18. Measurements and Phenomenological Modeling of Magnetic FluxBuildup in Spheromak Plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Romero-Talamas, C A; Hooper, E B; Jayakumar, R; McLean, H S; Wood, R D; Moller, J M

    2007-12-14

    Internal magnetic field measurements and high-speed imaging at the Sustained Spheromak Physics Experiment (SSPX) [E. B. Hooper, L. D. Pearlstein, R. H. Bulmer, Nucl. Fusion 39, 863 (1999)] are used to study spheromak formation and field buildup. The measurements are analyzed in the context of a phenomenological model of magnetic helicity based on the topological constraint of minimum helicity in the open flux before reconnecting and linking closed flux. Two stages are analyzed: (1) the initial spheromak formation, i. e. when all flux surfaces are initially open and reconnect to form open and closed flux surfaces, and (2) the stepwise increase of closed flux when operating the gun on a new mode that can apply a train of high-current pulses to the plasma. In the first stage, large kinks in the open flux surfaces are observed in the high-speed images taken shortly after plasma breakdown, and coincide with large magnetic asymmetries recorded in a fixed insertable magnetic probe that spans the flux conserver radius. Closed flux (in the toroidal average sense) appears shortly after this. This stage is also investigated using resistive magnetohydrodynamic simulations. In the second stage, a time lag in response between open and closed flux surfaces after each current pulse is interpreted as the time for the open flux to build helicity, before transferring it through reconnection to the closed flux. Large asymmetries are seen during these events, which then relax to a slowly decaying spheromak before the next pulse.

  19. Mechanism of internal pressure build-up within the heated coal grain

    Energy Technology Data Exchange (ETDEWEB)

    M. Sciazko; A. Karcz

    2002-07-01

    The model for pressure distribution within a heated solid coal grain was developed. The model is based on known kinetics of coal thermal decomposition, which in this case comprises eight, first order equations describing evolution of basic six gaseous components, i.e. methane, ethane, hydrogen, carbon oxide and dioxide, BTX, water and additionally total volatile matter. The kinetic model allows evaluating temperature dependent volatile matter flow rate and its average molecular mass. The primary internal pressure build-up and its distribution is the main reason for grain swelling while it turns into plastic state. Finally the model of grain swelling was developed which assumes temperature effect and coal properties described by Giseler characteristic temperatures namely minimum and maximum. To verify the model data the research done in Academy of Mining and Metallurgy Cracow were used on coal grain behaviour while heated. The developed model so called 'hollow grain core' and measured data on grain swelling, particularly the volume were used for the evaluation of model parameters. The developed model can be used for the interpretation of possible mechanism of wall pressure in a coking oven. 10 refs., 8 figs,. 3 tabs.

  20. Build-up and decline of organic matter during PeECE III

    Directory of Open Access Journals (Sweden)

    K. G. Schulz

    2008-05-01

    Full Text Available Increasing atmospheric carbon dioxide (CO2 concentrations due to anthropogenic fossil fuel combustion are currently changing the ocean's chemistry. Increasing oceanic [CO2] and consequently decreasing seawater pH have the potential to significantly impact marine life. Here we describe and analyze the build-up and decline of a natural phytoplankton bloom initiated during the 2005 mesocosm Pelagic Ecosystem CO2 Enrichment study (PeECE III. The draw-down of inorganic nutrients in the upper surface layer of the mesocosms was reflected by a concomitant increase of organic matter until day t11, the peak of the bloom. From then on, biomass standing stocks steadily decreased as more and more particulate organic matter was lost into the deeper layer of the mesocosms. We show that organic carbon export to the deeper layer was significantly enhanced at elevated CO2. This phenomenon might have impacted organic matter remineralization leading to decreased oxygen concentrations in the deeper layer of the high CO2 mesocosms as indicated by deep water ammonium concentrations. This would have important implications for our understanding of pelagic ecosystem functioning and future carbon cycling.