WorldWideScience

Sample records for duty bwr plants

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-12-31

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-01

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

  3. In-plant material test experience under hydrogen water chemistry at a Japanese BWR plant

    Energy Technology Data Exchange (ETDEWEB)

    Ando, Masami; Koshiishi, Masato; Kato, Takahiko [Hitachi Ltd., Ibaraki (Japan). Hitachi Works; Abe, Ayumi; Sekiguchi, Masahiko; Takiguchi, Hideki

    1999-07-01

    Hydrogen injection technology has been applied to Japanese domestic aged BWR plants since 1994 to mitigate corrosive environment regarding Intergranular Stress Corrosion Cracking (IGSCC) of Reactor Internals (RINs). The Tsuruga Unit-1 plant has also been operated with this technology since 1997, considering suppression of radiation increase in the main steam piping system besides mitigation of corrosive environment in the reactor; the hydrogen injection rate in the feed water was about 0.5 ppm. In order to confirm the effects of the hydrogen injection on suppression of SCC susceptibility of the RIN materials, several in-plant material tests have been conducted using the reactor water clean up system (RWCU). Cyclic-Slow Strain Rate Tensile (C-SSRT) test, Slow Strain Rate Tensile (SSRT) test and Compact Tension (CT) test were performed in the test facilities which were installed at the sampling line from the RWCU. Evaluation of SCC life by means of the C-SSRT test was the first application as an accelerated SCC test for in-plant material tests. It was confirmed that the hydrogen injection in the feed water has a good mitigation effects on IGSCC performance of the RIN materials. Results will be discussed from a viewpoint of the test condition such as total oxidant, ECP, conductivity and loading/unloading. (author)

  4. Plant analyzer for high-speed interactive simulation of BWR power plant transients

    Energy Technology Data Exchange (ETDEWEB)

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

    1984-04-01

    A combination of advanced modeling techniques and modern, special-purpose peripheral minicomputer technology is presented which affords realistic predictions of plant transient and severe off-normal events in LWR power plants through on-line simulations at a speed ten times faster than actual process speeds. Results are shown for a BWR plant simulation. The mathematical models account for nonequilibrium, nonhomogeneous two-phase flow effects in the coolant, for acoustical effects in the steam line and for the dynamics of the recirculation loop and feedwater train. Point kinetics incorporate reactivity feedback due to void fraction, fuel temperature, coolant temperature, and boron concentration. Control systems and trip logic are simulated for the nuclear steam supply system. The AD10 of Applied Dynamics International is the special-purpose peripheral processor. It is specifically designed for high-speed digital system simulation, accommodates hardware (instrumentation) in the input/output loop, and operates interactively on-line, like an analog computer. Results are shown to demonstrate computing capacity, accuracy, and speed. Simulation speeds have been achieved which are orders of magnitude faster than those of a CDC-7600 mainframe computer or ten times faster than real-time speed.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-04-01

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

  6. A safety and regulatory assessment of generic BWR and PWR permanently shutdown nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Travis, R.J.; Davis, R.E.; Grove, E.J.; Azarm, M.A. [Brookhaven National Lab., Upton, NY (United States)

    1997-08-01

    The long-term availability of less expensive power and the increasing plant modification and maintenance costs have caused some utilities to re-examine the economics of nuclear power. As a result, several utilities have opted to permanently shutdown their plants. Each licensee of these permanently shutdown (PSD) plants has submitted plant-specific exemption requests for those regulations that they believe are no longer applicable to their facility. This report presents a regulatory assessment for generic BWR and PWR plants that have permanently ceased operation in support of NRC rulemaking activities in this area. After the reactor vessel is defueled, the traditional accident sequences that dominate the operating plant risk are no longer applicable. The remaining source of public risk is associated with the accidents that involve the spent fuel. Previous studies have indicated that complete spent fuel pool drainage is an accident of potential concern. Certain combinations of spent fuel storage configurations and decay times, could cause freshly discharged fuel assemblies to self heat to a temperature where the self sustained oxidation of the zircaloy fuel cladding may cause cladding failure. This study has defined four spent fuel configurations which encompass all of the anticipated spent fuel characteristics and storage modes following permanent shutdown. A representative accident sequence was chosen for each configuration. Consequence analyses were performed using these sequences to estimate onsite and boundary doses, population doses and economic costs. A list of candidate regulations was identified from a screening of 10 CFR Parts 0 to 199. The continued applicability of each regulation was assessed within the context of each spent fuel storage configuration and the results of the consequence analyses.

  7. A conceptual study on large-capacity safety relief valve (SRV) for future BWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Yamada, Katsumi; Tokunaga, Takashi; Iwanaga, Masakazu; Kurosaki, Toshikazu [Toshiba Corporation, Isogo Nuclear Engineering Center, Yokohama (Japan)

    1999-07-01

    This paper presents a conceptual study of Safety Relief Valve (SRV) which has larger flow capacity than that of the conventional one and a new structure. Maintenance work of SRVs is one of the main concerns for next-generation Boiling Water Reactor (BWR) plants whose thermal power is planned to be increased. Because the number of SRVs increases with the thermal power, their maintenance would become critical during periodic inspections. To decrease the maintenance work, reduction of the number by increasing the nominal flow rate per SRV and a new structure suitable for easier treatment have been investigated. From a parameter survey of the initial and maintenance cost, the optimum capacity has been estimated to be between 180 and 200 kg/s. Primarily because the number of SRVs decreases in inversely proportional to the capacity, the total maintenance work decreases. The new structure of SRV, with an internally mounted actuator, decreases the number of the connecting parts and will make the maintenance work easier. A 1/4-scale model of the new SRV has been manufactured and performance tests have been conducted. The test results satisfied the design target, which shows the feasibility of the new structure. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-03-15

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

  9. 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.

  10. 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).

  11. Study of environmental noise in a BWR plant like the Nuclear Power Plant Laguna Verde; Estudio de ruido ambiental en una planta BWR como la Central Nuclear Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Tijerina S, F.; Cruz G, M.; Amador C, C., E-mail: francisco.tijerina@cfe.gob.mx [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Subgerencia de Ingenieria, Carretera Cardel-Nautla Km. 42.5, Alto Lucero, Veracruz (Mexico)

    2013-10-15

    In all industry type the health costs generated by the noise are high, because the noise can cause nuisance and to harm the capacity to work when causing tension and to perturb the concentration, and in more severe cases to reach to lose the sense of the hearing in the long term. The noise levels in the industry have been designated for the different types of use like residential, commercial, and industrial and silence areas. The noise can cause accidents when obstructing the communications and alarm signs. For this reason the noise should be controlled and mitigated, at a low level as reasonably is possible, taking into account that the noise is an acoustic contamination. The present study determines a bases line of the environmental noise levels in a nuclear power plant BWR-5 as Laguna Verde, (like reference) to be able to determine and to give pursuit to the possible solutions to eliminate or to limit the noise level in the different job areas. The noise levels were registered with a meter of integrative noise level (sonometer) and areas of noise exposure levels mapping the general areas in the buildings were established, being the registered maximum level of 96.94 dba in the building of the Reactor-elevation 0.65 m under the operation conditions of Extended Power Up rate (EPU) of 120% PTN. Knowing that the exposition to noises and the noise dose in the job place can influence in the health and in the safety of the workers, are extensive topics that they should be analyzed for separate as they are: to) the effects in the health of the exposure to the noise, b) how measuring the noise, c) the methods and technologies to combat and to control the noise in the industry by part of engineering area and d) the function of the industrial safety bodies as delegates of the health and safety in the task against the noise in the job. (author)

  12. Guidelines for confirmatory inplant tests of safety-relief valve discharges for BWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Su, T.M.

    1981-05-01

    Inplant tests of safety/relief valve (SRV) discharges may be required to confirm generically established specifications for SRV loads and the maximum suppression pool temperature, and to evaluate possible effects of plant-unique parameters. These tests are required in those plants which have features that differ substantially from those previously tested. Guidelines for formulating appropriate test matrices, establishing test procedures, selecting necessary instrumentation, and reporting the test results are provided in this report. Guidelines to determine if inplant tests are required on the basis of the plant unique parameters are also included in the report.

  13. Near-term improvements for nuclear power plant control room annunciator systems. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Rankin, W.L.; Duvernoy, E.G.; Ames, K.R.; Morgenstern, M.H.; Eckenrode, R.J.

    1983-04-01

    This report sets forth a basic design philosophy with its associated functional criteria and design principles for present-day, hard-wired annunciator systems in the control rooms of nuclear power plants. It also presents a variety of annunciator design features that are either necessary for or useful to the implementation of the design philosophy. The information contained in this report is synthesized from an extensive literature review, from inspection and analysis of control room annunciator systems in the nuclear industry and in related industries, and from discussions with a variety of individuals who are knowledgeable about annunciator systems, nuclear plant control rooms, or both. This information should help licensees and license applicants in improving their hard-wired, control room annunciator systems as outlined by NUREG-0700.

  14. Uncertainty analysis of suppression pool heating during an ATWS in a BWR-5 plant. An application of the CSAU methodology using the BNL engineering plant analyzer

    Energy Technology Data Exchange (ETDEWEB)

    Wulff, W.; Cheng, H.S.; Mallen, A.N. [Brookhaven National Lab., Upton, NY (United States); Johnsen, G.W. [Idaho National Engineering Lab., Idaho Falls, ID (United States); Lellouche, G.S. [Technical Data Services, Chicago, IL (United States)

    1994-03-01

    The uncertainty has been estimated of predicting the peak temperature in the suppression pool of a BWR power plant, which undergoes an NRC-postulated Anticipated Transient Without Scram (ATWS). The ATWS is initiated by recirculation-pump trips, and then leads to power and flow oscillations as they had occurred at the LaSalle-2 Power Station in March of 1988. After limit-cycle oscillations have been established, the turbines are tripped, but without MSIV closure, allowing steam discharge through the turbine bypass into the condenser. Postulated operator actions, namely to lower the reactor vessel pressure and the level elevation in the downcomer, are simulated by a robot model which accounts for operator uncertainty. All balance of plant and control systems modeling uncertainties were part of the statistical uncertainty analysis that was patterned after the Code Scaling, Applicability and Uncertainty (CSAU) evaluation methodology. The analysis showed that the predicted suppression-pool peak temperature of 329.3 K (133{degrees}F) has a 95-percentile uncertainty of 14.4 K (26{degrees}F), and that the size of this uncertainty bracket is dominated by the experimental uncertainty of measuring Safety and Relief Valve mass flow rates under critical-flow conditions. The analysis showed also that the probability of exceeding the suppression-pool temperature limit of 352.6 K (175{degrees}F) is most likely zero (it is estimated as < 5-104). The square root of the sum of the squares of all the computed peak pool temperatures is 350.7 K (171.6{degrees}F).

  15. 78 FR 13100 - Models for Plant-Specific Adoption of Technical Specifications Task Force Traveler TSTF-535...

    Science.gov (United States)

    2013-02-26

    ... (STS): NUREG-1433, ``Standard Technical Specifications General Electric Plants BWR/4,'' and NUREG- 1434... the STS, NUREG-1433, ``Standard Technical Specifications General Electric Plants BWR/4,'' and...

  16. Alternative cooling water flow path for RHR heat exchanger and its effect on containment response during extended station blackout for Chinshan BWR-4 plant

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-04-15

    Highlights: • Motivating alternative RHR heat exchanger tube-side flow path and determining required capacity. • Calculate NSSS and containment response during 24-h SBO for Chinshan BWR-4 plant. • RETRAN and GOTHIC models are developed for NSSS and containment, respectively. • Safety relief valve blowdown flow and energy to drywell are generated by RETRAN. • Analyses are performed with and without reactor depressurization, respectively. - Abstract: The extended Station Blackout (SBO) of 24 h has been analyzed with respect to the containment response, in particular the suppression pool temperature response, for the Chinshan BWR-4 plant of MARK-I containment. The Chinshan plant, owned by Taiwan Power Company, has twin units with rated core thermal power of 1840 MW each. The analysis is aimed at determining the required alternative cooling water flow capacity for the residual heat removal (RHR) heat exchanger when its tube-side sea water cooling flow path is blocked, due to some reason such as earthquake or tsunami, and is switched to the alternative raw water source. Energy will be dissipated to the suppression pool through safety relief valves (SRVs) of the main steam lines during SBO. The RETRAN model is used to calculate the Nuclear Steam Supply System (NSSS) response and generate the SRV blowdown conditions, including SRV pressure, enthalpy, and mass flow rate. These conditions are then used as the time-dependent boundary conditions for the GOTHIC code to calculate the containment pressure and temperature response. The shaft seals of the two recirculation pumps are conservatively assumed to fail due to loss of seal cooling and a total leakage flow rate of 36 gpm to the drywell is included in the GOTHIC model. Based on the given SRV blowdown conditions, the GOTHIC containment calculation is performed several times, through the adjustment of the heat transfer rate of the RHR heat exchanger, until the criterion that the maximum suppression pool temperature

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

    Directory of Open Access Journals (Sweden)

    Tanaka Ken-ichi

    2016-01-01

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

  18. BWR AXIAL PROFILE

    Energy Technology Data Exchange (ETDEWEB)

    J. Huffer

    2004-09-28

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

  19. Organizational analysis and safety for utilities with nuclear power plants: an organizational overview. Volume 1. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Osborn, R.N.; Olson, J.; Sommers, P.E.; McLaughlin, S.D.; Jackson, M.S.; Scott, W.G.; Connor, P.E.

    1983-08-01

    This two-volume report presents the results of initial research on the feasibility of applying organizational factors in nuclear power plant (NPP) safety assessment. A model is introduced for the purposes of organizing the literature review and showing key relationships among identified organizational factors and nuclear power plant safety. Volume I of this report contains an overview of the literature, a discussion of available safety indicators, and a series of recommendations for more systematically incorporating organizational analysis into investigations of nuclear power plant safety.

  20. Seismic risk assessment of a BWR: status report

    Energy Technology Data Exchange (ETDEWEB)

    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.

  1. Organizational analysis and safety for utilities with nuclear power plants: perspectives for organizational assessment. Volume 2. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Osborn, R.N.; Olson, J.; Sommers, P.E.; McLaughlin, S.D.; Jackson, M.S.; Nadel, M.V.; Scott, W.G.; Connor, P.E.; Kerwin, N.; Kennedy, J.K. Jr.

    1983-08-01

    This two-volume report presents the results of initial research on the feasibility of applying organizational factors in nuclear power plant (NPP) safety assessment. Volume 1 of this report contains an overview of the literature, a discussion of available safety indicators, and a series of recommendations for more systematically incorporating organizational analysis into investigations of nuclear power plant safety. The six chapters of this volume discuss the major elements in our general approach to safety in the nuclear industry. The chapters include information on organizational design and safety; organizational governance; utility environment and safety related outcomes; assessments by selected federal agencies; review of data sources in the nuclear power industry; and existing safety indicators.

  2. Procedures for using expert judgment to estimate human-error probabilities in nuclear power plant operations. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Seaver, D.A.; Stillwell, W.G.

    1983-03-01

    This report describes and evaluates several procedures for using expert judgment to estimate human-error probabilities (HEPs) in nuclear power plant operations. These HEPs are currently needed for several purposes, particularly for probabilistic risk assessments. Data do not exist for estimating these HEPs, so expert judgment can provide these estimates in a timely manner. Five judgmental procedures are described here: paired comparisons, ranking and rating, direct numerical estimation, indirect numerical estimation and multiattribute utility measurement. These procedures are evaluated in terms of several criteria: quality of judgments, difficulty of data collection, empirical support, acceptability, theoretical justification, and data processing. Situational constraints such as the number of experts available, the number of HEPs to be estimated, the time available, the location of the experts, and the resources available are discussed in regard to their implications for selecting a procedure for use.

  3. Common Cause Failure Analysis of Control Rods and Drives in the Swedish and Finnish BWR Plants. Operating Experiences in 1983 - 2003

    Energy Technology Data Exchange (ETDEWEB)

    Mankamo, Tuomas [Avaplan Oy, Espoo (Finland)

    2006-11-15

    substantial time difference and/or spatial distance within the core. The exploration of CCF cases showed that the most prevalent factor in the CCF mechanisms was the time coupling by sequencing maintenance into refueling outages. Essential contributing factors were design changes, deviations or errors in maintenance or new types of replacement parts, accompanied by unexpected influences. An evident positive trend could be observed both for single failures and CCFs . Impact Vectors were used to expresses the conditional failure probability for the various multiplicity in CCF events, linking event analysis to the estimation of CCF model parameters. A reference application was made for the Forsmark 1 and 2 plant. The Common Load Model was used as parametric CCF model, which proved to be a practicable approach. This method provides a consistent handling of failure combinations and workable extension to evaluate localized dependence between adjacent control rod and drives. Also international experience and reference information were surveyed. The developed methods and collected data are utilized in the ongoing PSA updates for the Swedish BWRs and Olkiluoto 1 and 2. Review - within the project a detailed and project extern review has been performed, covering also the older CCF events. This do now guarantee that the CCF data for the control rods and drives in Swedish and Finnish BWR:s during the observation period 1983 - 2003, now can be judged as quality assured. The scope of this project was limited to collection, analysis and classification of CCF data, and reference application using the industry average of pooled data. It has not been the scope of this project to perform more comprehensive probabilistic studies on e.g., positive learning trends, impact of plant specific design details or different amount of failing control rods at different operational conditions in the reactor vessel and with different safety and support systems in operation. It has either been the scope to

  4. 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.

  5. An overview of the BWR ECCS strainer blockage issues

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-03-01

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

  6. 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.

  7. Validation and Application of the Thermal Hydraulic System Code TRACE for Analysis of BWR Transients

    Directory of Open Access Journals (Sweden)

    V. H. Sánchez

    2012-01-01

    Full Text Available The Karlsruhe Institute of Technology (KIT is participating on (Code Applications and Maintenance Program CAMP of the US Nuclear Regulatory Commission (NRC to validate TRACE code for LWR transient analysis. The application of TRACE for the safety assessment of BWR requires a throughout verification and validation using experimental data from separate effect and integral tests but also using plant data. The validation process is normally focused on safety-relevant phenomena for example, pressure drop, void fraction, heat transfer, and critical power models. The purpose of this paper is to validate selected BWR-relevant TRACE-models using both data of bundle tests such as the (Boiling Water Reactor Full-Size Fine-Mesh Bundle Test BFBT and plant data recorded during a turbine trip event (TUSA occurred in a Type-72 German BWR plant. For the validation, TRACE models of the BFBT bundle and of the BWR plant were developed. The performed investigations have shown that the TRACE code is appropriate to describe main BWR-safety-relevant phenomena (pressure drop, void fraction, and critical power with acceptable accuracy. The comparison of the predicted global BWR plant parameters for the TUSA event with the measured plant data indicates that the code predictions are following the main trends of the measured parameters such as dome pressure and reactor power.

  8. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Alberdi, R.; Gonzalez, E.

    2001-07-01

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

  10. Selection of a closed Brayton cycle gas turbine for an intermediate-duty solar-electric power plant

    Science.gov (United States)

    Vieth, G. L.; Plummer, D. F.

    1980-03-01

    Subsystem and system analyses were performed to select the preferred working gas, performance characteristics and size of a closed cycle gas turbine for an intermediate-duty solar-electric power plant. Capital costs for all major subsystems were evaluated, but the principal selection criterion was the projected cost of electricity produced by the plant. Detailed analyses of the power conversion loop were conducted for both air and helium systems. Since the plant was intended for use on an intermediate-duty cycle, thermal storage was required. The coupling of the storage and power conversion loops in combination with the daily operating cycle influenced plant performance and energy costs in addition to the selection of the power conversion cycle.

  11. Analysis of the accident at Fukushima Daiichi nuclear power plant in an A BWR reactor; Analisis del accidente de la planta nucleoelectrica de Fukushima Daiichi en un reactor tipo ABWR

    Energy Technology Data Exchange (ETDEWEB)

    Escorcia O, D. [UNAM, Facultad de Ingenieria, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Salazar S, E., E-mail: daniel.escorcia.ortiz@gmail.com [UNAM, Facultad de Ingenieria, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, 62250 Jiutepec, Morelos (Mexico)

    2016-09-15

    The present work aims to recreate the accident occurred at the Fukushima Daiichi nuclear power plant in Japan on March 11, 2011, making use of an academic simulator of forced circulation of the A BWR reactor provided by the IAEA to know the scope of this simulator. The simulator was developed and distributed by the IAEA for academic purposes and contains the characteristics and general elements of this reactor to be able to simulate transients and failures of different types, allowing also to observe the general behavior of the reactor, as well as several phenomena and present systems in the same. Is an educational tool of great value, but it does not have a scope that allows the training of plant operators. To recreate the conditions of the Fukushima accident in the simulator, we first have to know what events led to this accident, as well as the actions taken by operators and managers to reduce the consequences of this accident; and the sequence of events that occurred during the course of the accident. Differences in the nuclear power plant behavior are observed and interpreted throughout the simulation, since the Fukushima plant technology and the simulator technology are not the same, although they have several elements in common. The Fukushima plant had an event that by far exceeded the design basis, which triggered in an accident that occurred in the first place by a total loss of power supply, followed by the loss of cooling systems, causing a level too high in temperature, melting the core and damaging the containment accordingly, allowing the escape of hydrogen and radioactive material. As a result of the simulation, was determined that the scope of the IAEA academic simulator reaches the entrance of the emergency equipment, so is able to simulate almost all the events occurred at the time of the earthquake and the arrival of the tsunami in the nuclear power plant of Fukushima Daiichi. However, due to its characteristics, is not able to simulate later

  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. Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Appendices VII, VIII, IX, and X. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    1975-10-01

    Information is presented concerning the release of radioactivity in reactor accidents; physical processes in reactor meltdown accidents; safety design rationale for nuclear power plants; and design adequacy.

  14. 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)

  15. Development of Heavy-Duty and High-Precision Hydraulic Manipulator for Inspection, Maintenance and Decommission of Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sung Uk; Seo, Yong-chil; Jung, Kyung Min; Kim, Chang-hoi; Choi, Byung-seon; Moon, Jei-kwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Robotic manipulators have been used for inspection, maintenance and decommission of nuclear power plants because nuclear power plants have high radiation and human workers cannot easily access the plants. And also, to inspecting, maintaining and decommissioning nuclear power plants require various manipulators. Only one manipulator cannot response to many required tasks. The existing manipulators that was used at nuclear power plants can only operate only focused specific task and cannot be used at several tasks. The actuators used at manipulators are varied and many companies sell actuators depending on power, torque and speed. However, the commercial product is not standardized. Therefore, the development of manipulator is time consuming and expensive. The essential item of a manipulator is an actuator module. If actuator module is standardized, it’s easier to develop a manipulator and also maintain a manipulator. Recently, manipulator having high-radiation, high-duty and high-precision is necessary to inspection, maintain and decommissioning of nuclear power plants. Hydraulic actuator has been used to development high-duty manipulator. But control performance of a hydraulic actuator is not better than that of an electric actuator so that hydraulic manipulator cannot easily satisfy the required precision. In this paper, we developed high-duty and high-precision actuator modules and hydraulic manipulator using the developed actuator modules. The developed hydraulic manipulator have a payload of 250kg and a precision of ±1mm. Four modularized hydraulic actuator modules were developed for inspection, maintenance and decommission. Using the developed actuator modules, the manipulator for decommissioning is easily developed. And also, various manipulators having different kinematic structure for specific tasks will be easily developed by using hydraulic modules.

  16. 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).

  17. 44 BWR Waste Package Loading Curve Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    J.M. Scaglione

    2001-11-05

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Dionne, B.J.; Baum, J.W.

    1984-01-01

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

  19. Stability prediction of continuous surveillance in BWR reactor; Predictor de estabilidad para la vigilancia continua de un reactor tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Tofino Gomez, Y.

    2006-07-01

    As result of the susceptibility of the Boiling Water Reactors (BWR) to suffer from power instabilities, the program LIP has been developed (LAPUR Input Preprocessor), which automatically determines the decay ratio (DR), as stability margin indication. For DR calculation, LAPUR program is a good predictive alternative: a fast execution for an acceptable precision. LAPUR demands a complex input, dependent on the instantaneous core configuration, requiring an exhaustive control of its generation. LIP, with a modular character, automatically generates the input from the core monitoring system, CAPRICORE (based on Simulate-3), obtaining the DR during the operation. This tool can accelerate the start-up maneuvers and other transients, increasing the plant availability. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Hermann, O.W.

    1998-01-01

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

  1. Validation of SCALE (SAS2H) isotopic predictions for BWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hermann, O.W.; DeHart, M.D.

    1998-09-01

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

  2. BWR online monitoring system based on noise analysis

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz-Villafuerte, Javier [Departamento de Sistemas Nucleares, Instituto Nacional de Investigaciones Nucleares, Carr. Mexico-Toluca S/N, Ocoyoacac, Edo. de Mexico, 52750 (Mexico)]. E-mail: jov@nuclear.inin.mx; Castillo-Duran, Rogelio [Departamento de Sistemas Nucleares, Instituto Nacional de Investigaciones Nucleares, Carr. Mexico-Toluca S/N, Ocoyoacac, Edo. de Mexico, 52750 (Mexico)]. E-mail: rcd@nuclear.inin.mx; Alonso, Gustavo [Departamento de Sistemas Nucleares, Instituto Nacional de Investigaciones Nucleares, Carr. Mexico-Toluca S/N, Ocoyoacac, Edo. de Mexico, 52750 (Mexico)]. E-mail: galonso@nuclear.inin.mx; Calleros-Micheland, Gabriel [Central Nuclear de Laguna Verde, Comision Federal de Electricidad, Carr. Cardel-Nautla, km. 42.5, Alto Lucero, Veracruz (Mexico)]. E-mail: gcm9acpp@cfe.gob.mx

    2006-11-15

    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.

  3. Digital implementation, simulation and tests in MATLAB of the models of Steam line, the turbines, the pressure regulator of a BWR type nucleo electric power plant; Implementacion digital, simulacion y pruebas en MATLAB de los modelos de la linea de vapor, las turbinas y el regulador de presion de una central Nucleoelectrica tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Lopez R, A. [UNAM, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: andyskamx@yahoo.com.mx

    2004-07-01

    In this phase of the project they were carried out exhaustive tests to the models of the steam lines, turbines and pressure regulator of a BWR type nucleo electric central for to verify that their tendencies and behaviors are it more real possible. For it, it was necessary to also analyze the transfer functions of the different components along the steam line until the power generator. Such models define alone the dominant poles of the system, what is not limitation to reproduce a wide range of anticipated transitoriness of a power station operation. In the same manner, it was integrated and proved the integrated model form with the models of feeding water of the SUN-RAH, simulating the nuclear reactor starting from predetermined entrances of the prospective values of the vessel. Also it was coupled with the graphic interface developed with the libraries DirectX implementing a specific monitoring panel for this system. (Author)

  4. Full system decontamination experience in BWR

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-07-01

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

  5. Validation and application of the system code ATHLET-CD for BWR severe accident analyses

    Energy Technology Data Exchange (ETDEWEB)

    Di Marcello, Valentino, E-mail: valentino.marcello@kit.edu; Imke, Uwe; Sanchez, Victor

    2016-10-15

    Highlights: • We present the application of the system code ATHLET-CD code for BWR safety analyses. • Validation of core in-vessel models is performed based on KIT CORA experiments. • A SB-LOCA scenario is simulated on a generic German BWR plant up to vessel failure. • Different core reflooding possibilities are investigated to mitigate the accident consequences. • ATHLET-CD modelling features reflect the current state of the art of severe accident codes. - Abstract: This paper is aimed at the validation and application of the system code ATHLET-CD for the simulation of severe accident phenomena in Boiling Water Reactors (BWR). The corresponding models for core degradation behaviour e.g., oxidation, melting and relocation of core structural components are validated against experimental data available from the CORA-16 and -17 bundle tests. Model weaknesses are discussed along with needs for further code improvements. With the validated ATHLET-CD code, calculations are performed to assess the code capabilities for the prediction of in-vessel late phase core behaviour and reflooding of damaged fuel rods. For this purpose, a small break LOCA scenario for a generic German BWR with postulated multiple failures of the safety systems was selected. In the analysis, accident management measures represented by cold water injection into the damaged reactor core are addressed to investigate the efficacy in avoiding or delaying the failure of the reactor pressure vessel. Results show that ATHLET-CD is applicable to the description of BWR plant behaviour with reliable physical models and numerical methods adopted for the description of key in-vessel phenomena.

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

    Energy Technology Data Exchange (ETDEWEB)

    Prill, Dennis; Class, Andreas G. [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany). AREVA Nuclear Professional School (ANPS)

    2013-07-01

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

  7. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.

    1985-01-01

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

  9. Assessment of two BWR accident management strategies

    Energy Technology Data Exchange (ETDEWEB)

    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.

  10. 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)

  11. 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)

  12. Analysis of high fidelity of a BWR fuel element with COBRA-TF/PARCS codes and TRACE; Analisis de Alta Fidelidad de un Elemento Combustible BWR con los codigos COBRA-TF/PARCS y TRACE

    Energy Technology Data Exchange (ETDEWEB)

    Abarca, A.; Miro, R.; Barrachina, T.; Verdu, G.; Solar, A.; Concejal, A.; Melara, J.; Albendea, M.

    2013-07-01

    It has been modeled a 10 x 10 BWR fuel element, containing 91 fuel rods (81 of 10 partial length and total length) and a great water bar of square section in the central part of it. Such fuel element has been modeled in detail: at the level of sub-channel code COBRA-TF and using parametric models for fuel elements BWR that owns the plant code TRACE. Has been an exercise in comparison of the results obtained by both codes in the simulation of a stationary and a small transient flow injection, highlighting the differences observed.

  13. Behavior of irradiated BWR fuel under reactivity-initiated-accident conditions; Results of tests FK-1, -2 and -3

    OpenAIRE

    2004-01-01

    Boiling water reactor (BWR) fuels with burnups of 41 to 45 GWd/tU were pulse-irradiated in the Nuclear Safety Research Reactor (NSRR) to investigate fuel behavior under cold startup reactivity-initiated-accident (RIA) conditions. BWR fuel segment rods of 8times8BJ (STEP I) type from Fukushima-Daiichi Unit 3 nuclear power plant were refabricated into short test rods, and they were subjected to prompt enthalpy insertion from 293 to 607 J/g (70 to 145 cal/g) within about 20 ms. The fuel cladding...

  14. BWR mechanics and materials technology update

    Energy Technology Data Exchange (ETDEWEB)

    Kiss, E.

    1983-05-01

    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.

  15. BWR Refill-Reflood Program, Task 4. 7 - model development: TRAC-BWR component models

    Energy Technology Data Exchange (ETDEWEB)

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

    1983-09-01

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

  16. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-03-01

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

  18. Ecophysiological Response of Plants to Combined Pollution from Heavy-duty Vehicles and Industrial Emissions in Higher Humidity

    Institute of Scientific and Technical Information of China (English)

    Hong-Xia Cui; Gao-Ming Jiang; Shu-Li Niu; Chuang-Dao Jiang; Mei-Zhen Liu; Shun-Li Yu; Lei-Ming Gao

    2006-01-01

    Pollution can be aggravated in industrial areas if traffic exhausts are mixed with industrial emissions under high humidity conditions. Plants growing in such environments may suffer from severe stress. The impact of vehicle emissions on urban vegetation in an industrial area in Qingdao, China, was investigated by studying seven plant species at visible, physiological and chemical levels. The traits of plant species in certain environmental conditions were compared between a clear area, Badaguan (BDG), and polluted area,Roadside (RS). We found that foliar sulfur uptake for all species was not significantly high at RS compared with BDG, although the sulfur content of atmosphere and surface soils at RS were much higher than those at BDG. For Ailanthus altissima Swingle, the content of foliar pigment and net photosynthesis rate (PN)decreased by 20%. Meanwhile, leaves became incrassate and no visible leaf damage was noted, suggesting this species could adapt well to pollution. A 50% decrease in PN occurred in Hibiscus syriacus L., but there was no statistical change in content of chlorophyll a and b and water uptake. Also, thickened leaves may prevent the pollutant from permeation. Foliar water content was still at a low level, although a water compensation mechanism was established for Fraxinus chinensis Rosb. reflected by low water potential and high water use efficiency. More adversely, a 65% decrease in PN happened inevitably with the significant decomposition of photosynthetic pigments, which exhibited visible damage. We also noted in one evergreen species (Magnolia grandiflora L.) that water absorption driven by low water potential should be helpful to supply water loss induced by strong stomatal transpiration and maintain normal growth. Furthermore, photosynthetic pigment content did not decline statistically, but supported a stable net assimilation. Two herbaceous species, Poa annua L. and Ophiopogon japonicus Ker-Gawl., were very tolerant to adverse stress

  19. 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)

  20. Results of the benchmarking in radiological protection practices during fuel reloads in the nuclear power plants of Limerick (BWR) and Ginna (PWR) in the United States of North America; Resultados del benchmarking en practicas de proteccion radiologica durante recargas de combustible en las centrales nucleoelectricas de Limerick (BWR) y Ginna (PWR) en los Estados Unidos de Norteamerica

    Energy Technology Data Exchange (ETDEWEB)

    Lara H, M. A., E-mail: marco.lara@cfe.gob.mx [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Carretera Cardel-Nautla Km 42.5, Alto Lucero, Veracruz (Mexico)

    2011-11-15

    The nuclear power plant of Laguna Verde, unique in our country, has been imposed several goals related with the continuous improvement of their acting; increase in the quantity of continuous days for operation cycle, improvement in the chemical indexes of the reactor coolant, improvement in the indexes of nuclear security, improvement in the indicators of industrial security, improvement in the standards of radiological protection, etc.; in this last item is precisely where is necessary to search creative solutions to be able to maintain the collective doses of the personnel so low as reasonably it is possible (ALARA) especially due to the last projects of extension of useful life of the nuclear power plant (zinc injection, noble metals and hydrogen) and of power increment of the nuclear power plant of Laguna Verde, same that represent in the short period an increment of collective dose and of exposition levels (until 200%) in very specific points of the primary systems of the reactor. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-01

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

  2. Entrainment and deposition modeling of liquid films with applications for BWR fuel rod dryout

    Science.gov (United States)

    Ratnayake, Ruwan Kumara

    While best estimate computer codes provide the licensing basis for nuclear power facilities, they also serve as analytical tools in overall plant and component design procedures. An ideal best estimate code would comprise of universally applicable mechanistic models for all its components. However, due to the limited understanding in these specific areas, many of the models and correlations used in these codes reflect high levels of empiricism. As a result, the use of such models is strictly limited to the range of parameters within which the experiments have been conducted. Disagreements between best estimate code predictions and experimental results are often explained by the mechanistic inadequacies of embedded models. Significant mismatches between calculated and experimental critical power values are common observations in the analyses of Boiling Water Reactors (BWR). Based on experimental observations and calculations, these mismatches are attributed to the additional entrainment and deposition caused by spacer grids in BWR fuel assemblies. In COBRA-TF (Coolant Boiling in Rod Arrays-Two Fluid); a state of the art industrial best estimate code, these disagreements are hypothesized to occur due the absence of an appropriate spacer grid model. In this thesis, development of a suitably detailed spacer grid model and integrating it to COBRA-TF is documented. The new spacer grid model is highly mechanistic so that the applicability of it is not seriously affected by geometric variations in different spacer grid designs. COBRA-TF (original version) simulations performed on single tube tests and BWR rod bundles with spacer grids showed that single tube predictions were more accurate than those of the rod bundles. This observation is understood to arise from the non-availability of a suitable spacer grid model in COBRA-TF. Air water entrainment experiments were conducted in a test section simulating two adjacent BWR sub channels to visualize the flow behavior at

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-10-01

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

  4. 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)

  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. 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.

  7. 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)

  8. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Verdugo, M.

    2015-07-01

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

  10. BWR ASSEMBLY SOURCE TERMS FOR WASTE PACKAGE DESIGN

    Energy Technology Data Exchange (ETDEWEB)

    T.L. Lotz

    1997-02-15

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

  11. 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

  12. 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

  13. Analysis of containment venting following a core damage at a BWR Mark I using THALES-2

    Energy Technology Data Exchange (ETDEWEB)

    Widodo, Surip [Nuclear Safety Technology Development Center, National Nuclear Energy Agency (BATAN), Tangerang (Indonesia); Ishikawa, Jun; Muramatsu, Ken [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Sakamoto, Toru [Toshiba Advanced System Co., Kawasaki, Kanagawa (Japan)

    2000-11-01

    Analysis of containment venting following a core damage at a boiling water reactor (BWR) Mark I using THALES-2 was performed. In this analysis, the effect of various parameters, namely, the areas of the vent path, containment venting pressure, and accident sequences on the containment thermodynamic response, and radionuclide transport and release in the containment venting at a BWR was examined. The code THALES-2B developed by the Japan Atomic Energy Research Institute (JAERI) was used in this analysis. The model plant in this analysis was the Browns Ferry plant. From this analysis was found that the 4-inch pipe of containment venting flow path is sufficient to maintain the containment pressure in the specified range if the containment was pressurized by the decay heat power. The entrainment by the pool swelling as well as by the flashing was not occurred during the containment venting. The source terms are not sensitive to the variation of containment venting flow path area. The containment venting pressure operation setting point has important rule in the containment venting. In the containment venting, the source terms are not sensitive to the accident sequence, except for Sr source term. In order to get better understanding on the containment venting strategy, the following analyses are necessary. Analyses of accident sequence which has a high power such as anticipated transient without scram are necessary, as well as analyses of accident sequence which pressurize the containment before the core damage. (author)

  14. TRACE Model for Simulation of Anticipated Transients Without Scram in a BWR

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-11-10

    A TRACE model has been developed for using theTRACE/PARCS computational package [1, 2] to simulate anticipated transients without scram (ATWS) events in a boiling water reactor (BWR). The model represents a BWR/5 housed in a Mark II containment. The reactor and the balance of plant systems are modeled in sufficient detail to enable the evaluation of plant responses and theeffectiveness of automatic and operator actions tomitigate this beyond design basis accident.The TRACE model implements features thatfacilitate the simulation of ATWS events initiated by turbine trip and closure of the main steam isolation valves (MSIV). It also incorporates control logic to initiate actions to mitigate the ATWS events, such as water levelcontrol, emergency depressurization, and injection of boron via the standby liquid control system (SLCS). Two different approaches have been used to model boron mixing in the lower plenum of the reactor vessel: modulate coolant flow in the lower plenum by a flow valve, and use control logic to modular.

  15. TRACE Model for Simulation of Anticipated Transients Without Scram in a BWR

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-11-10

    A TRACE model has been developed for using theTRACE/PARCS computational package [1, 2] to simulate anticipated transients without scram (ATWS) events in a boiling water reactor (BWR). The model represents a BWR/5 housed in a Mark II containment. The reactor and the balance of plant systems are modeled in sufficient detail to enable the evaluation of plant responses and theeffectiveness of automatic and operator actions tomitigate this beyond design basis accident.The TRACE model implements features thatfacilitate the simulation of ATWS events initiated by turbine trip and closure of the main steam isolation valves (MSIV). It also incorporates control logic to initiate actions to mitigate the ATWS events, such as water levelcontrol, emergency depressurization, and injection of boron via the standby liquid control system (SLCS). Two different approaches have been used to model boron mixing in the lower plenum of the reactor vessel: modulate coolant flow in the lower plenum by a flow valve, and use control logic to modular.

  16. 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.

  17. Estimate of coolant flow in assemblies of a natural circulation BWR applying and equivalent electric model; Estimacion del flujo de refrigerante en los ensambles de un BWR de circulacion natural aplicando un modelo electrico equivalente

    Energy Technology Data Exchange (ETDEWEB)

    Valle H, J.; Morales S, J. B. [UNAM, DEPFI, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico); Espinosa P, G. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)], e-mail: julfi_jg@yahoo.com.mx

    2009-10-15

    The present work exposes the design and implementation of an advanced controller that it allows to estimate the coolant flow in fuel assemblies of a natural circulation BWR in real time. the complete development of this study is part of a doctoral project in course. In this work the construction of optimal controller is shown that allows to estimate the coolant flows in reactor and its operation applied to an equivalent electric model to natural circulation ESBWR. The controller design that allows the completely automatic starter of natural circulation reactor, required of a variables estimator not meter directly of nuclear power plant and use of local distributions estimates of coolant flow, (this controller type at the moment is utilized in the A BWR and several BWR in operation in Japan). The construction of estimator controller is mathematically based in the theory referring to Kalman filter, whose algorithm provides an advanced control of system. To prove the estimator operation was developed a simplified model that reproduces the basic dynamic of coolant flowing in the ESBWR, a practice way and very interesting of representing this phenomenon is by means the use of an equivalent electric model, which was developed starting from analogies that there is among the relation that keep the pressure differences with the mass flow and differences of electric potential with electric current. A detailed analysis of equivalence among models will be presented in a later article. (Author)

  18. Evaluation of PWR and BWR pin cell benchmark results

    Energy Technology Data Exchange (ETDEWEB)

    Pijlgroms, B.J.; Gruppelaar, H.; Janssen, A.J. (Unit Nuclear Energy, Netherlands Energy Research Foundation ECN, Petten (Netherlands)); Hoogenboorm, J.E.; De Leege, P.F.A. (International Reactor Institute IRI, University of Leiden, Leiden (Netherlands)); Van de Voet, J.; Verhagen, F.C.M. (KEMA NV, Arnhem (Netherlands))

    1992-01-01

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

  19. 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)

  20. 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)

  1. 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)

  2. 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)

  3. Safety/relief valve quencher loads: evaluation for BWR Mark II and III containments

    Energy Technology Data Exchange (ETDEWEB)

    Su, T.M.

    1982-10-01

    Boiling water reactor (BWR) plants are equipped with safety/relief valves (SRVs) to protect the reactor from overpressurization. Plant operational transients, such as turbine trips, will actuate the SRV. Once the SRV opens, the air column within the partially submerged discharge line is compressed by the high-pressure steam released from the reactor. The compressed air discharged into the suppression pool produces high-pressure bubbles. Oscillatory expansion and contraction of these bubbles create hydrodynamic loads on the containment structures, piping, and equipment inside containment. This report presents the results of the staff's evaluation of SRV loads. The evaluation, however, is limited to the quencher devices used in Mark II and III containments. With respect to Mark I containments, the SRV acceptance criteria are presented in NUREG-0661 issued July 1980. The staff acceptance criteria for SRV loads for Mark II and III containments are presented in this report.

  4. Evaluation of PWR and BWR pin cell benchmark results

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-12-01

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

  5. BWR refill-reflood program: core spray distribution experimental task plan

    Energy Technology Data Exchange (ETDEWEB)

    Eckert, T.

    1981-02-01

    An experimental task plan for the BWR/4 core spray task of the Refill-Reflood Test Program is presented. The test program will provide core spray distribution data for a 30 degree sector of the BWR/4 and 5-218 design. This design uses different nozzle types and different sparger elevations than the BWR/6-218 design which was tested previously. Test parameter ranges are specified; individual tests are defined; and measurement and data utilization plans are defined.

  6. CCF analysis of high redundancy systems safety/relief valve data analysis and reference BWR application. Main report

    Energy Technology Data Exchange (ETDEWEB)

    Mankamo, T. [Avaplan Oy (Finland); Bjoere, S.; Olsson, Lena [ABB Atom AB, Vaesteraas (Sweden)

    1992-12-01

    Dependent failure analysis and modeling were developed for high redundancy systems. The study included a comprehensive data analysis of safety and relief valves at the Finnish and Swedish BWR plants, resulting in improved understanding of Common Cause Failure mechanisms in these components. The reference application on the Forsmark 1/2 reactor relief system, constituting of twelve safety/relief lines and two regulating relief lines, covered different safety criteria cases of reactor depressurization and overpressure protection function, and failure to re close sequences. For the quantification of dependencies, the Alpha Factor Model, the Binomial Probability Model and the Common Load Model were compared for applicability in high redundancy systems.

  7. Kuosheng BWR/6 containment pressure and temperature responses after recirculation line break using GOTHIC code

    Energy Technology Data Exchange (ETDEWEB)

    Lin, A.; Wang, J-R.; Chen, Y-S., E-mail: samuellin1999@iner.gov.tw, E-mail: jrwang@iner.gov.tw, E-mail: yschen@iner.gov.tw [Inst. of Nuclear Energy Research Atomic Energy Council (China); Shih, C., E-mail: ckshih@ess.nthu.edu.tw [National Tsing Hua Univ., Dept. of Engineering and System Science (China)

    2011-07-01

    In this study, we presented the calculated results of the containment P/T (pressure and temperature) response after the recirculation line break (RCLB) accident of a GE-designed twin-unit BWR/6 plant, which can be served as the design basis for the containment system. During the simulation, a power of SPU (stretch power uprate) range was used and a model of the Mark III type containment was built using the GOTHIC (Generation of Thermal-Hydraulic Information for Containments) code. The calculated results, similar to the FSAR (Final Safety Analysis Report) results, indicate the GOTHIC code has the capability to simulate the containment P/T response to the RCLB accident. (author)

  8. Optimized clearing work concept for the BWR containment; Optimiertes Raeumungskonzept fuer SWR-Sicherheitsbehaelter

    Energy Technology Data Exchange (ETDEWEB)

    Kraps, Uwe [AREVA NP GmbH (Germany)

    2012-11-01

    Based on the experiences of reactor dismantling in the NPPs Wuergasse, Obrigheim and Stade an optimized clearing work concept for the BWR containment including the reactor pressure vessel and the biological shield was developed. The concept is focused on the safety objective, the reduction of the collective dose and the reduction of the execution time. Precondition for the decommissioning license was up to now the removal of fuel elements from the reactor; due to the significantly increased period until fulfillment of this premises concepts are developed that can be performed with simultaneous reduction of the radiological inventories and the fire loads. The most important step of the guideline of the concept is the transition from hot to cold. The in-situ disassembling of the reactor internals can be performed with decreased water level in the reactor pressure vessel, with following water treatment and complete shutdown of operational systems. This status allows an accelerated further dismantling of the plant.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-04-01

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

  10. Security during the Construction of New Nuclear Power Plants: Technical Basis for Access Authorization and Fitness-For-Duty Requirements

    Energy Technology Data Exchange (ETDEWEB)

    Branch, Kristi M.; Baker, Kathryn A.

    2009-09-01

    A technical letter report to the NRC summarizing the findings of a benchmarking study, literature review, and workshop with experts on current industry standards and expert judgments about needs for security during the construction phase of critical infrastructure facilities in the post-September 11 U.S. context, with a special focus on the construction phase of nuclear power plants and personnel security measures.

  11. Influence of iron and nickel species upon activity buildup under simulated BWR conditions

    Energy Technology Data Exchange (ETDEWEB)

    Bjornsson, S.; Chen, J. [Studsvik Nuclear AB, Nykoping (Sweden); Lejon, J. [OKG AB, Oskarshamn (Sweden); Granath, G. [Ringhals AB, Varobacka (Sweden); Tanse-Larsson, M. [Forsmarks Kraftgrupp AB, Osthammar (Sweden)

    2010-07-01

    Activity build-up in BWR systems are of importance for service- and maintenance work performed at the plants. Minimizing the activity build-up is desirable for minimizing doses of personnel at the plants. Numerous studies have been carried out in this important field to understand the activity uptake mechanisms. This paper studied the possible role of Fe(II/III) and Ni(II) impurities in reactor water in activity uptake on stainless steel surfaces. The study was carried out by using a test loop with simulated BWR water containing Fe(II/III), Ni(II) and Co-60 marked Co(II) species of varied concentration and 500 ppb O{sub 2}. The test tube section in the loop system was pre-exposed type 316L stainless steel material. The microstructures of the formed oxide films were examined with high resolution electron microscopy (FE-SEM and FE-TEM). The activity monitoring on the test section showed that injection of 10 ppb Ni(II) and 0.1 ppb Fe(II/III) in the water with 0.1 ppb Co(II) was capable of stopping completely activity uptake. When Co(II) addition in the loop was stopped no activity return to the water could be seen. In another exposure test, injection of combined 2 ppb Fe(II/III) and 0.5∼10 ppb Ni(II) profoundly increased activity uptake on the test section with a maximum in activity buildup at 5 ppb Ni(II). When Co(II) addition in the loop was stopped a slight activity return was seen. The observed differences as seen in the two tests are discussed in view of the microstructures of the oxide films formed. (author)

  12. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    A.K. Scheider

    2005-08-23

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

  14. Development of advanced BWR fuel bundle with spectral shift rod - BWR core characteristics with SSR

    Energy Technology Data Exchange (ETDEWEB)

    Hino, T.; Kondo, T.; Chaki, M.; Ohga, Y. [Hitachi-GE Nuclear Energy, Ltd., 1-1, Saiwai-cho, 3-chome, Hitachi-shi, Ibaraki-ken, 317-0073 (Japan); Makigami, T. [Tokyo Electric Power Company Inc., 1-1-3, Uchisaiwai-cho, Chiyoda-ku, Tokyo, 100-0011 (Japan)

    2012-07-01

    The neutron energy spectrum can be varied during an operation cycle to generate and utilize more plutonium from the non-fissile {sup 238}U by changing the void fraction in the core through control of the core coolant flow rate. This operation method, which is called a spectral shift operation, is practiced in BWRs to save natural uranium. A new component called a spectral shift rod (SSR), which is utilized instead of a conventional water rod, has been introduced to amplify the void fraction change and increase the spectral shift effect. In this study, fuel bundle design with the SSR and core design were carried out for the ABWR and the next generation BWR, HP-ABWR (High-Performance ABWR).The core characteristics with the SSR were evaluated and compared with those when using the conventional water rod. Influences of uncertainty of the water level in the SSR on the safety limit minimum critical power ratio (SLMCPR) were considered for evaluation of the uranium saving effect attained by the SSR. As a result, it was found that the amount of natural uranium needed for an operation cycle could be reduced more than 3% with 20% core coolant flow change and more than 5% with 30% core coolant flow change, in the form of increased discharge exposure by using the SSR compared with the conventional water rod use. (authors)

  15. Evaluation of PWR and BWR pin cell benchmark results

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-11-01

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

  17. Analysis of a German BWR core with TRACE/PARCS using different cross section sets

    Energy Technology Data Exchange (ETDEWEB)

    Hartmann, C., E-mail: Christoph.Hartmann@kit.edu [Karlsruhe Inst. of Tech. (KIT), Inst. for Neutron Physics and Reactor Technology (INR), Eggenstein-Leopoldshafen (Germany); Westinghouse Electric Germany GmbH, Mannheim (Germany); Sanchez, V.H. [Karlsruhe Inst. of Tech. (KIT), Inst. for Neutron Physics and Reactor Technology (INR), Eggenstein-Leopoldshafen (Germany); Tietsch, W. [Westinghouse Electric Germany GmbH, Mannheim (Germany)

    2011-07-01

    'Full text:' Coupled Thermal-hydraulic/Neutron-kinetic (TH/NK) simulations of Boiling Water Reactor transients require well validated and accurate simulation tools as well as appropriate cross sections (XS) libraries depending on the individual thermal hydraulic state parameters. Problem-dependent XS-sets for 3D core simulations are being generated mainly by well validated, fast running and user-friendly lattice codes such as Casco and Helios. At research institutions and universities, alternative tools to the commercial ones with full access to the source code as well as moderate cost are urgently needed. The Scale system is being developed and improved for lattice physics calculations of real core loading of Light Water Reactors (LWR). It represents a promising alternative to the commercial lattice codes. At Karlsruhe Institute of Technology (Kit) a computational route based on Scale/Triton/Newt for BWR core loading is under development. The generated XS-data sets have to be transformed in PMAXS-format for use in the reactor dynamic code PARCS. This task is performed by the module GenPMAXS being developed and tested at the Michigan University. To verify the computational route, a BWR fuel assembly depletion problem was calculated by PARCS and compared to the CASMO results. Since the SCALE/TRITON XS-file does actually not contain all required neutronic data, FORTRAN routines have been developed to incorporate the missing data e.g. the yields of Iodine, Xenon and Promethium into the XS-data sets in the PMAXS-format. The comparison of the results obtained with PARCS (using the corrected PMAXS file) and CASMO for the depletion problem exhibited a good agreement. Consequently, this approach was followed for the generation of a complete XS-set for a real BWR core to be used in subsequent transient analysis. Then 3D neutronic and thermal hydraulic core model were elaborated for a TRACE/PARCS analysis. The thermal hydraulic model is based on the 3D VESSEL

  18. 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

  19. BWR spent fuel transport and storage system for KKL: TN trademark 52L, TN trademark 97L, TN trademark 24 BHL

    Energy Technology Data Exchange (ETDEWEB)

    Sicard, D.; Verdier, A. [COGEMA Logistics (AREVA Group) (France); Monsigny, P.A. [NOK/KKL (Switzerland)

    2004-07-01

    The LEIBSTADT (KKL) nuclear power plant in Switzerland has opted to ship spent fuel to a central facility called ZWILAG for interim storage. In the mid-nineties, COGEMA LOGISTICS was contracted by KKL for the supply of the TN trademark a52L and TN trademark 97L transport and storage casks for BWR fuel types. In 2003, KKL also ordered from COGEMA LOGISTICS the supply of six TNae24 BHL transport and storage casks. This paper shows how all the three cask designs have responded to the KKL needs to ship and store BWR spent fuel. In addition, it highlights the already significant operational feedback of the TN trademark 52L and TN trademark 97L casks by the KKL and ZWILAG operators.

  20. RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - models and correlations

    Energy Technology Data Exchange (ETDEWEB)

    Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.; Neymotin, L.Y.

    1998-03-01

    This document describes the major modifications and improvements made to the modeling of the RAMONA-3B/MOD0 code since 1981, when the code description and assessment report was completed. The new version of the code is RAMONA-4B. RAMONA-4B is a systems transient code for application to different versions of Boiling Water Reactors (BWR) such as the current BWR, the Advanced Boiling Water Reactor (ABWR), and the Simplified Boiling Water Reactor (SBWR). This code uses a three-dimensional neutron kinetics model coupled with a multichannel, non-equilibrium, drift-flux, two-phase flow formulation of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients and instability issues. Chapter 1 is an overview of the code`s capabilities and limitations; Chapter 2 discusses the neutron kinetics modeling and the implementation of reactivity edits. Chapter 3 is an overview of the heat conduction calculations. Chapter 4 presents modifications to the thermal-hydraulics model of the vessel, recirculation loop, steam separators, boron transport, and SBWR specific components. Chapter 5 describes modeling of the plant control and safety systems. Chapter 6 presents and modeling of Balance of Plant (BOP). Chapter 7 describes the mechanistic containment model in the code. The content of this report is complementary to the RAMONA-3B code description and assessment document. 53 refs., 81 figs., 13 tabs.

  1. RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - models and correlations

    Energy Technology Data Exchange (ETDEWEB)

    Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.; Neymotin, L.Y.

    1998-03-01

    This document describes the major modifications and improvements made to the modeling of the RAMONA-3B/MOD0 code since 1981, when the code description and assessment report was completed. The new version of the code is RAMONA-4B. RAMONA-4B is a systems transient code for application to different versions of Boiling Water Reactors (BWR) such as the current BWR, the Advanced Boiling Water Reactor (ABWR), and the Simplified Boiling Water Reactor (SBWR). This code uses a three-dimensional neutron kinetics model coupled with a multichannel, non-equilibrium, drift-flux, two-phase flow formulation of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients and instability issues. Chapter 1 is an overview of the code`s capabilities and limitations; Chapter 2 discusses the neutron kinetics modeling and the implementation of reactivity edits. Chapter 3 is an overview of the heat conduction calculations. Chapter 4 presents modifications to the thermal-hydraulics model of the vessel, recirculation loop, steam separators, boron transport, and SBWR specific components. Chapter 5 describes modeling of the plant control and safety systems. Chapter 6 presents and modeling of Balance of Plant (BOP). Chapter 7 describes the mechanistic containment model in the code. The content of this report is complementary to the RAMONA-3B code description and assessment document. 53 refs., 81 figs., 13 tabs.

  2. Planning guidance for nuclear-power-plant decontamination. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Munson, L.F.; Divine, J.R.; Martin, J.B.

    1983-06-01

    Direct and indirect costs of decontamination are considered in the benefit-cost analysis. A generic form of the benefit-cost ratio is evaluated in monetary and nonmonetary terms, and values of dollar per man-rem are cited. Federal and state agencies that may have jurisiction over various aspects of decontamination and waste disposal activities are identified. Methods of decontamination, their general effectiveness, and the advantages and disadvantages of each are outlined. Dilute or concentrated chemical solutions are usually used in-situ to dissolve the contamination layer and a thin layer of the underlying substrate. Electrochemical techniques are generally limited to components but show high decontamination effectiveness with uniform corrosion. Mechanical agents are particularly appropriate for certain out-of-system surfaces and disassembled parts. These processes are catagorized and specific concerns are discussed. The treatment, storage, and disposal or discharge or discharge of liquid, gaseous, and solid wastes generated during the decontamination process are discussed. Radioactive and other hazardous chemical wastes are considered. The monitoring, treatment, and control of radioactive and nonradioactive effluents, from both routine operations and possible accidents, are discussed. Protecting the health and safety of personnel onsite during decontamination is of prime importance and should be considered in each facet of the decontamination process. The radiation protection philosophy of reducing exposure to levels as low as reasonably achievable should be stressed. These issues are discussed.

  3. 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)

  4. Simulation of Thermal Stratification in BWR Suppression Pools with One Dimensional Modeling Method

    Energy Technology Data Exchange (ETDEWEB)

    Haihua Zhao; Ling Zou; Hongbin Zhang

    2014-01-01

    The suppression pool in a boiling water reactor (BWR) plant not only is the major heat sink within the containment system, but also provides the major emergency cooling water for the reactor core. In several accident scenarios, such as a loss-of-coolant accident and extended station blackout, thermal stratification tends to form in the pool after the initial rapid venting stage. Accurately predicting the pool stratification phenomenon is important because it affects the peak containment pressure; the pool temperature distribution also affects the NPSHa (available net positive suction head) and therefore the performance of the Emergency Core Cooling System and Reactor Core Isolation Cooling System pumps that draw cooling water back to the core. Current safety analysis codes use zero dimensional (0-D) lumped parameter models to calculate the energy and mass balance in the pool; therefore, they have large uncertainties in the prediction of scenarios in which stratification and mixing are important. While three-dimensional (3-D) computational fluid dynamics (CFD) methods can be used to analyze realistic 3-D configurations, these methods normally require very fine grid resolution to resolve thin substructures such as jets and wall boundaries, resulting in a long simulation time. For mixing in stably stratified large enclosures, the BMIX++ code (Berkeley mechanistic MIXing code in C++) has been developed to implement a highly efficient analysis method for stratification where the ambient fluid volume is represented by one-dimensional (1-D) transient partial differential equations and substructures (such as free or wall jets) are modeled with 1-D integral models. This allows very large reductions in computational effort compared to multi-dimensional CFD modeling. One heat-up experiment performed at the Finland POOLEX facility, which was designed to study phenomena relevant to Nordic design BWR suppression pool including thermal stratification and mixing, is used for

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-09-01

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

  6. 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

  7. Cofrentes EOC16B poolside measurements of channels from the three BWR vendors

    Energy Technology Data Exchange (ETDEWEB)

    Sedano, Pablo J. Garcia; Ayuela, Javier Iglesias [Iberdrola Ingenieria Construccion SAU, Veronica Anaures (Mexico); Albendea, Manuel [Iberdrola Generation S. A., Plaza Euskadi, 5 48009 BILBAO (Spain)

    2008-10-15

    As part of the EPRI Fuel Reliability Program, a fuel channel focus group was formed in 2002 to initiate measurements on irradiated BWR fuel channels. Fuel channels from GNF and AREVA have been measured in campaigns performed during 2004{approx}2007. Fuel channels designed and supplied by Westinghouse were of particular interest since no measurement information had been previously taken on modern Westinghouse channels operating on conventional loading pattern cycles, either in European or U.S. plants. Conventional loading pattern cycles are more susceptible to experience shadow corrosion induced bow since the fresh bundles are exposed to control blade influence early in life. During summer of 2007 extensive poolside measurements of a total of 180 fuel channels (24 SVEA-96 +/L, 68 SVEA-96 Optima-2, 36 GE-12, 42 GE-14 and 10 ATRIUM-10XP) have been performed by Westinghouse at Cofrentes NPP (Spanish BWR-6 operating on 24 month cycle strategy). This campaign has been co-sponsored by EPRI, Iberdrola and Westinghouse Sweden. Channels covering a range of exposure and control blade history were selected in order to determine the dependency of the channel deformation with those parameters. Channels with the most limiting conditions of exposure and control blade history were included. Channel bow, bulge and twist have been measured and fast neutron fluence calculations have been performed in order to determine the effects of neutron fluence gradient and shadow corrosion on the total channel deformation. Additionally channel oxide measurements have been performed on 20 channels from the three fuel vendors.The results indicate that channel bow and bulge remained at anticipated levels with no indication of significant channel bow due to shadow corrosion phenomenon. Destructive metallographic evaluations of samples taken from one cycle Westinghouse channels with high control blade exposure are underway at Studsvik hot cell facilities. These examinations will provide additional

  8. Prony's method application for BWR instabilities characterization

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-04-01

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

  9. 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.

  10. Diffusion bonded matrix of HGMF applied for BWR condensate water purification

    Energy Technology Data Exchange (ETDEWEB)

    Soda, Fumitaka; Yukawa, Takao; Ito, Kazuyuki

    1984-03-01

    A high Gradient Magnetic Filter (HGMF) applied to the purification of power plant primary water has recently attracted much attention. In the application of HGMF to the water treatment of power plants, especially nuclear power plants, reliabillties of matrix (filtering medium) as well as removal performance for cruds (insoluble corrosion products) are considered to be important factors. To satisfy these factors, a new filtering medium named Diffision Bonded Matrix (DBM) has been developed and the test results are reported. Filtering efficiency and mechanical stiffness of DBM were examined using HGMF pilot test units consisting of 160 mm diameters x 240 mm length filter. The filtering velocity and the magnetic flux density used in this test were 800 m/h 5 kG, respectively. The filtering efficiencies and of 85-100% were obtained for artificial cruds for DBM. The DBM indicated slightly better filtering efficiency than for conventional wool matrix under the same filtering and matrix conditions. The DBM kept its original mechanical properties and very few pieces of fibers were broken off while the conventional wool matrix lost its volume elasticities and the considerable amount of fibers was broken off during the test operation. The results described here demonstrated the applicability of DBM for treatment of BWR primary water by High Gradient Magnetic Filter.

  11. 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)

  12. A detailed BWR recirculation loop model for RELAP

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-01-15

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

  13. Environmental mitigation for SCC initiation of BWR core internals by hydrogen injection during start-up

    Energy Technology Data Exchange (ETDEWEB)

    Dozaki, K.; Abe, A.; Nagata, N.; Takiguchi, H. [The Japan Atomic Power Co. (Japan)

    2004-07-01

    Hydrogen injection into the reactor water has been applied to many BWR power stations. Since hydrogen injected accelerates recombination of oxidant generated by water radiolysis, oxidant concentration, such as dissolved oxygen concentration in reactor water can be reduced. As the result of the reduction of oxidant concentration, Electrochemical Corrosion Potential (ECP) at the surface of structural material can be lowered. Lowered ECP moderates Stress Corrosion Cracking (SCC) sensitivity of structural materials, such as stainless steels. As usual, hydrogen injection system begins to work after the plant start-up is finished, when the condition of normal operation is established. Accordingly, Hydrogen Water Chemistry (HWC) does not cover all the period of plant operation. As far as SCC crack growth is considered, loss of HWC during plant start-up does not result in significant crack growth, because of duration of plant start-up is much shorter than that of plant normal operation, when HWC condition is being satisfied. However, the reactor water environment and load conditions during a plant start-up may contribute to the initiation of SCC. It is estimated that the core internals are subjected to the strain rate that may cause susceptibility to SCC initiation during start-up. Dissolved oxygen (DO) and hydrogen peroxide (H{sub 2}O{sub 2}) has a peak, and ECP is in high levels during start-up. Therefore it is beneficial to perform hydrogen injection during start-up as well in order to suppress SCC initiation. We call it HWC During Start-up (HDS) here. (orig.)

  14. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-10-15

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

  18. BWR: Development and Validation of KERENA reactor; Les REB: Developpement et validation du reacteur KERENA

    Energy Technology Data Exchange (ETDEWEB)

    Diercks, F.; Fuchs, M. [E.ON Kernkraft GmbH (Germany); Erve, M.; Pasler, D. [AREVA (Germany)

    2010-07-01

    KERENA is an advanced boiling water reactor, combining AREVA's and E.ON's expertise. A project was launched to customize the final basic design for this advanced nuclear power plant having a net power output of about 1, 250 MW, a net efficiency of about 37% and a design service life of 60 years. The development takes into account the technical and accumulated operating experience of the project partners. The plant safety concept is based on an optimized combination of a reduced number of proven active safety systems and passive safety systems, utilizing basic laws of physics, such as gravity, enabling them to function without electrical power supplies or activation by powered instrumentation and control systems. Control of a postulated core melt accident is assured with considerable safety margins thanks to passive flooding of the containment for in-vessel melt retention. All passive safety systems are validated in an experimental test program at AREVA, using 1:1 scale test facilities (INKA test facility Karlstein). The KERENA boiling water reactor is compliant with international nuclear codes and standards, and is also designed to withstand the effects of an aircraft crash involving a military aircraft or a large passenger airline. The safety level of the KERENA reactor has been able to be significantly increased compared to existing BWR plants. The advantages of the new safety concept are: -) Reduced susceptibility of safety systems to failures; -) Larger safety margins; -) Good plant behavior in the event of accidents due to the fact that conditions change at a slower rate; -) Grace periods of several days after an accident before operator intervention is required; -) Significantly reduced impact of operator error on reactor safety; -) No need for large-scale emergency response actions such as temporary evacuation or relocation of the neighboring population following a core melt accident. (A.C.)

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

    Science.gov (United States)

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

    2016-12-01

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

  20. Description and assessment of RAMONA-3B Mod. 0 Cycle 4: a computer code with three-dimensional neutron kinetics for BWR system transients

    Energy Technology Data Exchange (ETDEWEB)

    Wulff, W; Cheng, H S; Diamond, D J; Khatib-Rahbar, M

    1984-01-01

    This report documents the physical models and the numerical methods employed in the BWR systems code RAMONA-3B. The RAMONA-3B code simulates three-dimensional neutron kinetics and multichannel core hydraulics of nonhomogeneous, nonequilibrium two-phase flows. RAMONA-3B is programmed to calculate the steady and transient conditions in the main steam supply system for normal and abnormal operational transients, including the performances of plant control and protection systems. Presented are code capabilities and limitations, models and solution techniques, the results of development code assessment and suggestions for improving the code in the future.

  1. Investigation of valve failure problems in LWR power plants

    Energy Technology Data Exchange (ETDEWEB)

    None

    1980-04-01

    An analysis of component failures from information in the computerized Nuclear Safety Information Center (NSIC) data bank shows that for both PWR and BWR plants the component category most responsible for approximately 19.3% of light water reactor (LWR) power plant shutdowns. This investigation by Burns and Roe, Inc. shows that the greatest cause of shutdowns in LWRs due to valve failures is leakage from valve stem packing. Both BWR plants and PWR plants have stem leakage problems (BWRs, 21% and PWRs, 34%).

  2. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-12-01

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

  4. Thermal utilization opportunities with a small-to-medium sized BWR

    Energy Technology Data Exchange (ETDEWEB)

    Konkin, D.; Simonson, C.J.; Dalai, A.K.; Tanino, K.; Guo, H., E-mail: doug.konkin@usask.ca [Univ. of Saskatchewan, Saskatchewan (Canada); Nishida, K.; Mochida, T. [Hitachi-GE Nuclear Energy, Ltd., Ibaraki (Japan); Ikegawa, T.; Kito, K. [Hitachi, Ltd., Ibaraki (Japan); Knudsen, R. [LeanOptions Consulting, Inc., Regina, Saskatchewan (Canada); Aikman, A. [SNC-Lavalin, Saskatoon, Saskatchewan (Canada); Humphries, R. [AMEC, Toronto, Ontario (Canada)

    2014-07-01

    Hitachi-GE Nuclear Energy Ltd. (Hitachi-GE) has developed a conceptual design for a Double MS: Modular Simplified & Medium Small Reactor (DMS) under the sponsorship of The Japan Atomic Power Company. Recent efforts have yielded enhancements for improved safety and reactor core performance. The DMS is an innovative small-to-medium sized Boiling Water Reactor (BWR), which, based only on electricity generation, has been estimated to almost overcome economy of scale concerns when compared to proven conventional Advanced Boiling Water Reactor (ABWR) technologies. In order to make the DMS more attractive, the University of Saskatchewan (U of S), Hitachi-GE and Hitachi Ltd. (Hitachi) have collaborated on a joint research and development (R&D) initiative to study the utilization of heat and steam from the Balance of Plant (BOP) associated with the DMS for thermal utilization (TU) applications. In this paper, the advanced features of the DMS and the individual projects of the R&D program will be described. (author)

  5. Duty and Distance

    NARCIS (Netherlands)

    C. Binder (C.); C. Heilmann (Conrad)

    2017-01-01

    markdownabstractEver since the publication of Peter Singer’s article ‘‘Famine, Affluence, and Morality’’ has the question of whether the (geographical) distance to people in need affects our moral duties towards them been a hotly debated issue. Does geographical distance affect our moral duties?

  6. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Campos Gonzalez, Rina Margarita

    2007-07-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, William BJ J [ORNL

    2016-01-01

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

  9. 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.

  10. Influence of the wet-well nodalization of a BWR3 Mark I on the containment thermal-hydraulic response during an SBO accident

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, Luis E., E-mail: luisen.herranz@ciemat.es; Fontanet, Joan; Fernández, Elena; López, Claudia

    2015-12-15

    Highlights: • Analysis of SBO sequences in BWR3 Mark I containments. • Multiple-mesh nodalization allows pool stratification set up. • Mass, momentum and energy exchanges between nodes play a key role. • Validation/verification against a scaled-down database required to credit meshing schemes. - Abstract: In the field of severe accidents simulation one of the most challenging issues is nodalization. This paper explores the effect of the wet well modeling on significant variables describing the sequence evolution. The code used for the study has been MELCOR 2.1 and the scenario chosen has been a prolonged SBO occurring in a BWR3 Mark I. The results indicate that some significant magnitudes show a moderate scatter depending on WW nodalization (i.e., core uncovery, RPV failure, hydrogen production), whereas the SP thermal state might display outstanding deviations, which sometimes affect significantly key variables like containment pressure. The difficulties and uncertainties around defining a suitable WW nodalization have been highlighted and the need to properly balance the entire plant meshing has been stressed. Even though a number of noding schemes has been explored, the results discussion underlines the importance of having a deep understanding of the potential phenomena governing the scenario and of mastering the code facilities to better model it. Some insights into WW nodalization in MELCOR 2.1 have been gained for the specific scenario (i.e., a prolonged SBO in a BWR3 Mark I) explored: a single node assumption might underestimate PCV pressurization; a loose coupling of water and gas exchanges in the WW nodalization would be preferred if the drift flux model is chosen for momentum exchange in the flow pathways between WW nodes; the potential of some axial thermal stratification in the pool should be taken into account when noding the WW; sensitivity analyses on physically supported WW nodalization schemes should be conducted and focused on key

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  12. 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)

  13. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    H. Wang

    1997-01-23

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

  15. Minor Actinides Loading Optimization for Proliferation Resistant Fuel Design - BWR

    Energy Technology Data Exchange (ETDEWEB)

    G. S. Chang; Hongbin Zhang

    2009-09-01

    One approach to address the United States Nuclear Power (NP) 2010 program for the advanced light water reactor (LWR) (Gen-III+) intermediate-term spent fuel disposal need is to reduce spent fuel storage volume while enhancing proliferation resistance. One proposed solution includes increasing burnup of the discharged spent fuel and mixing minor actinide (MA) transuranic nuclides (237Np and 241Am) in the high burnup fuel. Thus, we can reduce the spent fuel volume while increasing the proliferation resistance by increasing the isotopic ratio of 238Pu/Pu. For future advanced nuclear systems, MAs are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. MAs play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. A typical boiling water reactor (BWR) fuel unit lattice cell model with UO2 fuel pins will be used to investigate the effectiveness of adding MAs (237Np and/or 241Am) to enhance proliferation resistance and improve fuel cycle performance for the intermediate-term goal of future nuclear energy systems. However, adding MAs will increase plutonium production in the discharged spent fuel. In this work, the Monte-Carlo coupling with ORIGEN-2.2 (MCWO) method was used to optimize the MA loading in the UO2 fuel such that the discharged spent fuel demonstrates enhanced proliferation resistance, while minimizing plutonium production. The axial averaged MA transmutation characteristics at different burnup were compared and their impact on neutronics criticality and the ratio of 238Pu/Pu discussed.

  16. Numerical simulation of boron injection in a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Tinoco, Hernan, E-mail: htb@forsmark.vattenfall.s [Forsmarks Kraftgrupp AB, SE-742 03 Osthammar (Sweden); Buchwald, Przemyslaw [Reactor Technology, Royal Institute of Technology, SE-100 44 Stockholm (Sweden); Frid, Wiktor, E-mail: wiktor@reactor.sci.kth.s [Reactor Technology, Royal Institute of Technology, SE-100 44 Stockholm (Sweden)

    2010-02-15

    The present study constitutes a first step to understand the process of boron injection, transport and mixing in a BWR. It consists of transient CFD simulations of boron injection in a model of the downcomer of Forsmark's Unit 3 containing about 6 million elements. The two cases studied are unintentional start of boron injection under normal operation and loss of offsite power with partial ATWS leaving 10% of the core power uncontrolled. The flow conditions of the second case are defined by means of an analysis with RELAP5, assuming boron injection start directly after the first ECCS injection. Recent publications show that meaningful conservative results may be obtained for boron or thermal mixing in PWRs with grids as coarse as that utilized here, provided that higher order discretization schemes are used to minimize numerical diffusion. The obtained results indicate an apparently strong influence of the scenario in the behavior of the injection process. The normal operation simulation shows that virtually all boron solution flows down to the Main Recirculation Pump inlet located directly below the boron inlet nozzle. The loss of offsite power simulation shows initially a spread of the boron solution over the entire sectional area of the lower part of the downcomer filled with colder water. This remaining effect of the ECCS injection lasts until all this water has left the downcomer. Above this region, the boron injection jet develops in a vertical streak, eventually resembling the injection of the normal operation scenario. Due to the initial spread, this boron injection will probably cause larger temporal and spatial concentration variations in the core. In both cases, these variations may cause reactivity transients and fuel damage due to local power escalation. To settle this issue, an analysis using an extended model containing the downcomer, the MRPs and the Lower Plenum will be carried out. Also, the simulation time will be extended to a scale of

  17. 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)

  18. 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)

  19. Detailed investigation of radiolytic gas accumulation in BWR piping with the CMFD-Code helios

    Energy Technology Data Exchange (ETDEWEB)

    Tilman Diesselhorst [Framatome ANP GmbH, NGPS1, P.O.Box 3220, D-91058 Erlangen (Germany); Vladimir Stevanovic [University of Belgrade, 27 Marta 80, 11000 Belgrade (Yugoslavia)

    2005-07-01

    Full text of publication follows: In consequence of the incidents with hydrogen reaction in the Brunsbuettel and Hamaoka NPP's the accumulation of radiolytic gases in piping and components of BWR plants was checked using large scale temperature monitoring. But the more situations in the systems had been looked at, the more questions came up. It had to be realized that not in any concrete situation the possibility of critical gas accumulation can clearly be predicted due to many parameters of influence and conditions, partly acting contradictorily. Therefore the computational multi-fluid dynamic code HELIOS was developed to get a tool for the prediction of radiolytic gas accumulation allowing a detailed insight in the complex thermo-hydraulic and physic-chemical processes. Standard CFD codes generally give no satisfying and reliable results in this field. In the HELIOS code the processes of steam inflow into the modeled piping, convection and diffusion are taken into account, as well as the drainage of condensate film. The conservation equations of mass, energy and momentum are three-dimensionally formulated, separately for steam, gases and condensate. Considered are condensation and heat transfer in the fluid and the pipe wall and isolation, the influence of the gas fraction on the efficiency of condensation included. The effects of absorption and degassing between gas volume and liquid film are also considered. Nearly any piping arrangement can be modeled consisting of vertical, horizontal and inclined sections and bends of different diameter. Various conditions of heat losses can be considered. Results are the time dependent fields of gas/vapor fraction and condensate and the respective temperatures and flow velocities, as well as the accumulation rate of radiolytic gas. The code is verified in its formulations and validated in its applicability for relevant piping systems by comparing the results with analytical solutions, results of quasi one

  20. Optimal estimate of the coolant flow in the assemblies of a BWR of natural circulation in real time; Estimacion optima del flujo de refrigerante en los ensambles de un BWR de circulacion natural en tiempo real

    Energy Technology Data Exchange (ETDEWEB)

    Valle H, J.; Morales S, J. B. [UNAM, Facultad de Ingenieria, Division de Estudios de Posgrado, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, Col. Progreso, 62550 Jiutepec, Morelos (Mexico); Espinosa P, G., E-mail: julfi_jg@yahoo.com.mx [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)

    2012-10-15

    The present work exposes the design and the implementation of an advanced controller that allows estimating the coolant flow in the fuel assemblies of a BWR reactor of natural circulation in real time. To be able to reduce the penalizations that are established in the calculations of the operation limits due to the magnitude of the uncertainties in the coolant flows of a natural circulation reactor, is the objective of this research. In this work the construction of the optimal controller that allows estimating the coolant flows in a fuel channels group of the reactor is shown, as well as the operation of this applied to a reduced order model that simulates the dynamics of a natural circulation reactor. The controller design required of an estimator of the valuation variables not directly of the plant and of the estimates use of the local distributions of the coolant flow. The controller construction of the estimator was based mathematically in the filter Kalman whose algorithm allows to be carried out an advanced control of the system. To prove the estimator operation was development a simplified model that reproduces the basic dynamics of the flowing coolant in the reactor, which works as observer of the system, this model is coupled by means of the estimator controller to a detail model of the plant. The results are presented by means of graphics of the interest variables and the estimate flow, and they are documented in the chart that is presented at the end of this article. (Author)

  1. Simulation of the flow obstruction of a jet pump in a BWR reactor with the code RELAP/SCDAPSIM; Simulacion de la obstruccion de flujo de una bomba jet en un reactor BWR con el codigo RELAP/SCDAPSIM

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas V, J.; Filio L, C., E-mail: jaime.cardenas@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose M. Barragan 779, Col. Narvarte, 03020 Ciudad de Mexico (Mexico)

    2016-09-15

    This work simulates the flow obstruction of a jet pump in one of the recirculation loops of a nuclear power plant with a reactor of type BWR at 100% of operating power, in order to analyze the behavior of the total flow of the refrigerant passing through the reactor core, the total flow in each recirculation loop of the reactor, together with the 10 jet pumps of each loop. The behavior of the power and the reactivity insertion due to the change of the refrigerant flow pattern is also analyzed. The simulation was carried out using the RELAP/SCDAPSIM version 3.5 code, using a reactor model with 10 jet pumps in each recirculation loop and a core consisting of 6 radial zones and 25 axial zones. The scenario postulates the flow obstruction in a jet pump in a recirculation loop A when the reactor operates at 100% rated power, causing a change in the total flow of refrigerant in the reactor core, leading to a decrease in power. Once the reactor conditions are established to its new power, the operator tries to recover the nominal power using the flow control valve of the recirculation loop A, opening stepwise as a strategy to safely recover the reactor power. In this analysis is assumed that the intention of the nuclear plant operator is to maintain the operation of the reactor during the established cycle. (Author)

  2. 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.

  3. 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

  4. 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.

  5. [Ethics, power and duty].

    Science.gov (United States)

    Gândara, Manuela

    2008-01-01

    Nursing care is a humane action which bears with it an ethical dimension which is revealed by the focus of the care provided. Among the different options for action, a nurse will choose that which contributes to the care of a patient without harming that patient. The care choice made is the result of a conscientious, deliberate decision-making process which presupposes a recognition of what one plans to do, what one is capable of doing and what can accomplish. Freedom to choose which care option is applied imposes on a nurse the duty to act according to principles and duties which rule the profession; a nurse's power to act transforms to a power of the duty to provide treatment.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-05-15

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

  7. 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)

  8. Duty and Distance

    NARCIS (Netherlands)

    C. Binder (C.); C. Heilmann (Conrad)

    2017-01-01

    markdownabstractEver since the publication of Peter Singer’s article ‘‘Famine, Affluence, and Morality’’ has the question of whether the (geographical) distance to people in need affects our moral duties towards them been a hotly debated issue. Does geographical distance affect our moral

  9. O plantão noturno em anestesia reduz a latência ao sono El plantón nocturno en anestesia reduce la latencia al sueño Short sleep latency in residents after a period on duty in anesthesia

    Directory of Open Access Journals (Sweden)

    Lígia Andrade da Silva Telles Mathias

    2004-10-01

    ñana, después 24 horas de trabajo, sin dormir, sin plantón en los 3 días anteriores (M2; a las 13 horas de la tarde, después de 30 horas de trabajo, sin dormir, sin plantón en los 3 días anteriores (M3. En todas esas situaciones fue realizado electroencefalograma (EEG continuo, en sala apropiada para registro de los señales de sueño, evaluándose la latencia del sueño (LS. RESULTADOS: Se Verificó reducción significativa de la LS entre los residentes, después de 24 ó 30 horas de plantón sin dormir. Entre los praticantes que tuvieron noche de sueño normal en la víspera del examen, 36,4% presentaron LS en nivel considerado patológico. CONCLUSIONES: La jornada de plantón de 24 ó 30 horas lleva a valores de LS menores que 5 minutos, considerados patológicos, reflejando la fatiga extrema de residentes de Anestesiologia. Pode ser importante la reglamentación del número de horas de descanso pos-plantón.BACKGROUND AND OBJECTIVES: Physicians in general, and anesthesiologists in particular, have long working hours. Residents of Anesthesiology may present significant fatigue and stress. This study aimed at investigating first and second year residents’ sleep latency after a period on duty. METHODS: Participated in this study 11 residents in different situations: at 7:00 am, after a normal night sleep (> 7 h, without on duty period in the last 3 days (M1; at 7:00 am, after 24h of night work, without on duty period in the last 3 days (M2; and at 1:00 pm after 30h of work without on duty period in the last 3 days (M3. Continuous EEG was performed for all situations in adequate room to record sleep signals. Sleep latency (SL was evaluated. RESULTS: There has been significant shorter SL among residents after 24 or 30 hours without sleep. From residents after a normal night sleep the day before the evaluation, 36.4% presented pathological SL levels. CONCLUSIONS: Periods on duty for 24 or 30 hours lead to SL values below 5 minutes, which are considered pathologic and

  10. Resolving conflicts of duty in fiduciary relationships

    National Research Council Canada - National Science Library

    Laby, Arthur B

    2004-01-01

    While duties of loyalty generally do not conflict with other duties of loyalty, and while conflicting duties of care typically only raise issues of competing resources, the fiduciary's duty of loyalty...

  11. 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.

  12. 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

  13. 19 CFR 151.65 - Duties.

    Science.gov (United States)

    2010-04-01

    ... revenue will be properly protected. Liquidated duties shall be based upon the port director's final... 19 Customs Duties 2 2010-04-01 2010-04-01 false Duties. 151.65 Section 151.65 Customs Duties U.S...) EXAMINATION, SAMPLING, AND TESTING OF MERCHANDISE Wool and Hair § 151.65 Duties. Duties on wool or...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-07-01

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

  15. Analysis of Heat Balance on Innovative-Simplified Nuclear Power Plant Using Multi-Stage Steam Injectors

    Science.gov (United States)

    Goto, Shoji; Ohmori, Shuichi; Mori, Michitsugu

    The total space and weight of the feedwater heaters in a nuclear power plant (NPP) can be reduced by replacing low-pressure feedwater heaters with high-efficiency steam injectors (SIs). The SI works as a direct heat exchanger between feedwater from condensers and steam extracted from turbines. It can attain pressures higher than the supplied steam pressure. The maintenance cost is lower than that of the current feedwater heater because of its simplified system without movable parts. In this paper, we explain the observed mechanisms of the SI experimentally and the analysis of the computational fluid dynamics (CFD). We then describe mainly the analysis of the heat balance and plant efficiency of the innovative-simplified NPP, which adapted to the boiling water reactor (BWR) with the high-efficiency SI. The plant efficiencies of this innovative-simplified BWR with SI are compared with those of a 1100MWe-class BWR. The SI model is adopted in the heat balance simulator as a simplified model. The results show that the plant efficiencies of the innovate-simplified BWR with SI are almost equal to those of the original BWR. They show that the plant efficiency would be slightly higher if the low-pressure steam, which is extracted from the low-pressure turbine, is used because the first-stage of the SI uses very low pressure.

  16. Euthanasia and physicians' moral duties.

    Science.gov (United States)

    Seay, Gary

    2005-10-01

    Opponents of euthanasia sometimes argue that it is incompatible with the purpose of medicine, since physicians have an unconditional duty never to intentionally cause death. But it is not clear how such a duty could ever actually be unconditional, if due consideration is given to the moral weight of countervailing duties equally fundamental to medicine. Whether physicians' moral duties are understood as correlative with patients' moral rights or construed noncorrelatively, a doctor's obligation to abstain from intentional killing cannot be more than a defeasible duty.

  17. Estimating boiling water reactor decommissioning costs: A user`s manual for the BWR Cost Estimating Computer Program (CECP) software. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Bierschbach, M.C. [Pacific Northwest National Lab., Richland, WA (United States)

    1996-06-01

    Nuclear power plant licensees are required to submit to the US Nuclear Regulatory Commission (NRC) for review their decommissioning cost estimates. This user`s manual and the accompanying Cost Estimating Computer Program (CECP) software provide a cost-calculating methodology to the NRC staff that will assist them in assessing the adequacy of the licensee submittals. The CECP, designed to be used on a personal computer, provides estimates for the cost of decommissioning boiling water reactor (BWR) power stations to the point of license termination. Such cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial costs; and manpower costs. In addition to costs, the CECP also calculates burial volumes, person-hours, crew-hours, and exposure person-hours associated with decommissioning.

  18. Estimating boiling water reactor decommissioning costs. A user`s manual for the BWR Cost Estimating Computer Program (CECP) software: Draft report for comment

    Energy Technology Data Exchange (ETDEWEB)

    Bierschbach, M.C. [Pacific Northwest Lab., Richland, WA (United States)

    1994-12-01

    With the issuance of the Decommissioning Rule (July 27, 1988), nuclear power plant licensees are required to submit to the U.S. Regulatory Commission (NRC) for review, decommissioning plans and cost estimates. This user`s manual and the accompanying Cost Estimating Computer Program (CECP) software provide a cost-calculating methodology to the NRC staff that will assist them in assessing the adequacy of the licensee submittals. The CECP, designed to be used on a personal computer, provides estimates for the cost of decommissioning BWR power stations to the point of license termination. Such cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial costs; and manpower costs. In addition to costs, the CECP also calculates burial volumes, person-hours, crew-hours, and exposure person-hours associated with decommissioning.

  19. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-01

    The objective of this report is to identify issues and support documentation and identify and detail existing research on spent fuel dry storage; provide information to support potential R&D for the UKABWR (United Kingdom Advanced Boiling Water Reactor) Spent Fuel Interim Storage (SFIS) Pre-Construction Safety Report; and support development of answers to questions developed by the regulator. Where there are gaps or insufficient data, Oak Ridge National Laboratory (ORNL) has summarized the research planned to provide the necessary data along with the schedule for the research, if known. Spent nuclear fuel (SNF) from nuclear power plants has historically been stored on site (wet) in spent fuel pools pending ultimate disposition. Nuclear power users (countries, utilities, vendors) are developing a suite of options and set of supporting analyses that will enable future informed choices about how best to manage these materials. As part of that effort, they are beginning to lay the groundwork for implementing longer-term interim storage of the SNF and the Greater Than Class C (CTCC) waste (dry). Deploying dry storage will require a number of technical issues to be addressed. For the past 4-5 years, ORNL has been supporting the U.S. Department of Energy (DOE) in identifying these key technical issues, managing the collection of data to be used in issue resolution, and identifying gaps in the needed data. During this effort, ORNL subject matter experts (SMEs) have become expert in understanding what information is publicly available and what gaps in data remain. To ensure the safety of the spent fuel under normal and frequent conditions of wet and subsequent dry storage, intact fuel must be shown to: 1.Maintain fuel cladding integrity; 2.Maintain its geometry for cooling, shielding, and subcriticality; 3.Maintain retrievability, and damaged fuel with pinhole or hairline cracks must be shown not to degrade further. Where PWR (pressurized water reactor) information is

  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. Full scale stability and void fraction measurements for the ATRIUM trademark 10XM BWR fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Wehle, Franz; Velten, Roger; Kronenberg, Juris; Beisiegel, Achim [AREVA NP GmbH, Erlangen (Germany); Pruitt, D.W.; Greene, K.R. [AREVA NP Inc., Lynchburg, VA (United States); Farawila, Y.M. [Farawila et al., Inc., Richland, WA (United States)

    2011-07-01

    This paper describes recent advances in BWR fuel testing at AREVA NP's KATHY loop including stability and void fraction measurements. The stability tests for the ATRIUM trademark 10XM bundle with corner PLFR's were expanded in scope compared with previous campaigns to include simulated reactivity and power feedback essentially reproducing BWR operational environment. The oscillation magnitude was allowed to grow to explore inlet flow reversal and cyclical dryout and rewetting. The void fraction measurements employed a gamma ray computed tomography technique that reveals not only the average but the detailed sub-channel void distribution, and the range of measured void fraction has been expanded to higher values than was previously attained. With the completion of the required licensing tests and stability performance demonstration, the ATRIUM trademark 10XM is available and fully qualified for reload supply. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Costa, A. L.; Pereira, C.; Da Silva, C. A. M.; Veloso, M. A. F. [Departamento de Engenharia Nuclear, Universidade Federal de Minas Gerais, Av. Antonio Carlos, 6627, CEP 31270-90, Belo Horizonte, MG (Brazil); Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores/CNPq (Brazil)

    2009-07-01

    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

  7. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Tautges, T.J.

    1993-10-01

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

  9. Characterization of corrosion layers on irradiated and non-irradiated surfaces in BWR conditions

    Energy Technology Data Exchange (ETDEWEB)

    Kysela, J.; Balek, V.; Zmitko, M.; Brozova, A.; Burda, J. [Nuclear Research Inst., Rez (Czech Republic); Hoffmann, H.; Ruehle, W. [VGB Essen (Germany); Bezdicka, P. [Institute of Inorganic Chemistry, ASCR, Rez (Czech Republic)

    2002-07-01

    Stress corrosion cracking of low-alloyed steel 22NiMoCr37 is evaluated with the goal to determine crack growth rate in irradiated steel under conditions simulating closely conditions of BWR RPV under operation. For the experiment, in pile BWR experimental loop has been built at Nuclear Research Institute, Rez. During the experiment, specimens are loaded by cyclic and constant load. Crack growth is monitored by means of potential drop measurement and COD. Corrosion layers formed on specimens in reactor water loop exposed to BWR primary water chemistry and radiation were studied. Two sets of specimens were placed in loop channels. One set of specimens was situated in reactor conditions and the second set out of reactor, other parameters like water chemistry (e.g. concentration of hydrogen, oxygen and conductivity), temperature and flow rate were identical. By means of this an effect of radiation could be studied. The differences in chemical composition, structure and microstructure of corrosion products were characterized by SEM and X-ray powder diffractometry. The differences in microstructure of corrosion layer formed under different conditions were observed. (authors)

  10. 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)

  11. Last experiences on ID BWR shroud inspection and the new developments to examine the below core plate areas

    Energy Technology Data Exchange (ETDEWEB)

    Alberdi, R.; Gonzalez, E.; Willke, A.; Yague, L. [TECNATOM SA, Madrid (Spain)

    2001-07-01

    In recent years, the owners of BWR type nuclear power plants have had to address new inspection requirements relating to the core shroud inside the reactor vessel, the aim of which is to contain the fuel assemblies and provide support for the structures located in the upper part of the reactor. The shroud consists of a cylinder measuring some 40-50 mm in thickness, manufactured from various sections of AISI-304 stainless steel and INCONEL, joined by vertical and circumferential welds. The appearance of unstable cracks in these welds would directly affect the structural integrity of the component and the safety of the plant. As regards access to the core shroud and to the surface to be examined, two alternatives might be considered: inspection from outside the component, moving along the so-called annulus between the reactor vessel wall and the component (OD inspection), or from the interior (ID inspection). With a view to addressing this problem, Tecnatom has in recent years launched several projects, grouped under the generic name TEIDE, in order to develop scanners and NDT techniques achieving the maximum inspection coverage of this component. The decision was taken to perform ID inspections, mainly because this type of scanners were not available at that time, and which provide the 4 following advantages. 1) Maximum inspected weld length. This avoids interference with the jet pumps and the systems present in the annulus and affecting OD inspections. Besides, the repairs performed on in-service core shrouds in all cases imply the addition of new fixed elements on their outer surface, since the fuel assembly space must be left free. 2) Reduction of inspection times and of unforeseen events: maintenance of planning schedules, reduction of personnel doses, reduced critical path time. 3) High inspection accuracy and repeatability. 4) Simplification of equipment positioning work (similar to the installation of fuel assemblies). As regards inspection techniques, the

  12. Analysis by the Monte Carlo method of doses around the pool of storage of the control rods irradiated in a BWR reactor; Analisis mediante el metodo de Monte Carlo de las dosis alrededor de la piscina de almacenamiento de las barras de control irradiadas en un reactror BWR

    Energy Technology Data Exchange (ETDEWEB)

    Rodenas, J.; Gallardo, S.

    2011-07-01

    The control rods of a boiling water reactor (BWR) are subject to a neutron flux and thus become activated during their stay in the reactor core. Activation occurs especially in the stainless steel components and impurities. The activity generated results in a dose around the bar, while it le unimportant in the reactor, but to be taken into account when removed f ron it. The bars drawn are stored on hangers placed in the storage pools of spent fuel f ron the plant. Each hanger 12 accommodates control rods and are arranged so that at least three meters of water abode the heads of the control rods. The dose received by potentially exposed workers who are in the vicinity of the storage must be calculated to ensure adequate protection of the came. This dose can be decreased significantly by changing the arrangement of the bars on hangers.

  13. Reactor scram experience for shutdown system reliability analysis. [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Edison, G.E.; Pugliese, S.L.; Sacramo, R.F.

    1976-06-01

    Scram experience in a number of operating light water reactors has been reviewed. The date and reactor power of each scram was compiled from monthly operating reports and personal communications with the operating plant personnel. The average scram frequency from ''significant'' power (defined as P/sub trip//P/sub max greater than/ approximately 20 percent) was determined as a function of operating life. This relationship was then used to estimate the total number of reactor trips from above approximately 20 percent of full power expected to occur during the life of a nuclear power plant. The shape of the scram frequency vs. operating life curve resembles a typical reliability bathtub curve (failure rate vs. time), but without a rising ''wearout'' phase due to the lack of operating data near the end of plant design life. For this case the failures are represented by ''bugs'' in the plant system design, construction, and operation which lead to scram. The number of scrams would appear to level out at an average of around three per year; the standard deviations from the mean value indicate an uncertainty of about 50 percent. The total number of scrams from significant power that could be expected in a plant designed for a 40-year life would be about 130 if no wearout phase develops near the end of life.

  14. PVT modeling of reservoir fluids using PC-SAFT EoS and Soave-BWR EoS

    DEFF Research Database (Denmark)

    Yan, Wei; Varzandeh, Farhad; Stenby, Erling Halfdan

    2015-01-01

    non-cubic EoS models, such as the Perturbed Chain Statistical Associating Fluid Theory (PC-SAFT) EoS and the Soave modified Benedict-Webb-Rubin (Soave-BWR) EoS, may partly replace the roles of these classical cubic models in the upstream oil industry. Here, we attempt to make a comparative study...... of non-cubic models (PC-SAFT and Soave-BWR) and cubic models (SRK and PR) in several important aspects related to PVT modeling of reservoir fluids, including density description for typical pure components in reservoir fluids, description of binary VLE, prediction of multicomponent phase envelopes...... and Soave-BWR. For PC-SAFT, new correlations for estimating its model parameters in heptanes plus are developed. The results reveal that the non-cubic models are clearly advantageous in density calculation of pure components. For binary VLE and multicomponent phase envelopes, the results are similar...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  16. Simplified system for the pressure control of a Nucleo electric central of the BWR type; Sistema simplificado para el control de presion de una central Nucleoelectrica del tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez J, J. [FI-UNAM, DEPFI Campus Morelos, Jiutepec, Morelos (Mexico)

    2003-07-01

    One of the main preoccupations of the electric power generator stations is the appropriate operation of the same ones. The operators must be qualified to respond in an adequate way and to be able to take to these power stations to an optimal, sure and stable operation condition under any circumstance. The Laboratory of Analysis in Nuclear Reactors Engineering (LAIRN) of the Engineering Faculty of UNAM (Fl) in collaboration with the International Atomic Energy Agency (IAEA), it develops an interactive classroom simulator in which simulations of the phenomena which take place in a nuclear power station are executed. The classroom simulator bases its operation on specialized nuclear codes feeding interactive graphic unfolding with those that it is possible to make a monitoring, supervision and control of the behavior of the power station under any operation regime, either in normal operation, transitory events or postulated accident sequence. The development of this classroom simulator includes a modular and re configurable structure. Due to it is indispensable to count with a higher inter activity with the system it is included the simulation of the control system of the plant and inside the same, one of those more important it is the reactor pressure control system. The present work describes the conceptual design and the used methodology for the development and implementation in the simulator of a simplified model of the pressure control system for a BWR generic central. The reach of the development will allow to accomplish the necessary tests to demonstrate that this has an adequate performance according to the carried out simplifications. (Author)

  17. Crack growth tests on a ferritic reactor pressure vessel steel under the simultaneous influence of simulated BWR coolant and irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmann, H. [VGB PowerTech e.V., Essen (Germany); Huettner, F. [Hamburgische Electricitaets-Werke AG, Hamburg (Germany); Ilg, U. [EnBW Kraftwerke AG, Philippsburg (Germany); Wachter, O. [E.ON Kernkraft GmbH, Hannover(Germany); Widera, M. [RWE Power AG, Essen (Germany); Brozova, A.; Ernestova, M.; Kysela, J.; Vsolak, R. [Nuclear Research Institute Rez plc (Czech Republic)

    2004-07-01

    Crack growth tests under constant load with initial in-situ cycling were performed on the low alloy reactor pressure vessel (RPV) steel 22 NiMoCr 3 7 (A 508 Cl. 2) with the goal to determine crack growth rates of irradiated and non-irradiated steel under the simultaneous influence of simulated BWR coolant and irradiation. The tests were performed under conditions as near as possible to operational conditions in a commercial BWR reactor. The research results are summarized and are compared with international data. (orig.)

  18. 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)

  19. 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)

  20. 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)

  1. 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

  2. IMPACT OF CPO EXPORT DUTIES ON MALAYSIAN PALM OIL INDUSTRY

    Directory of Open Access Journals (Sweden)

    Ibragimov Abdulla

    2014-01-01

    Full Text Available In January 2013, Malaysia reduced the export duty structure to be in line with the Indonesia’s duty structure. Both countries export crude and processed palm oil. Since Malaysia and Indonesia are close competitors and they compete in the same market, a change in export duty rate in one country will affect the other. Indonesia, as the world’s biggest palm oil producer, has drastically widened the values between the crude palm oil and refined palm oil export taxes since October 2011, to encourage more downstream investments and production of refined palm oil products. Under the revised export duty structure, crude palm oil and crude palm kernel oil are cheaper for downstream activities in Indonesia. The new structure is expected to reduce Malaysia’s competitiveness in the world market as its export duty is relatively higher. A high export duty results in high price of crude palm oil which is the raw material for processed palm oil. The research questions are: (i What are the likely future trends of crude palm oil exports under the new crude palm oil export duties? Will it increase, reduce or stabilize? (ii What are the likely future trends of processed palm oil exports? Will it increase exponentially, stabilize or reduce? To answer these questions, a system dynamics model was developed for the Malaysian palm oil. Application of the system dynamics model provides a framework to understand the feedback structure and how changes in variables impact the behavior of the palm oil industry. This research suggests that with low crude palm oil export duties crude palm oil domestic price, profitability of plantation owners, immature crop, mature crop, total planted area, production and exports of crude palm oil increase, however exports of processed palm oil decrease.

  3. Assessment of the mechanical performance of the Westinghouse BWR control rod CR 99 at high depletion levels

    Energy Technology Data Exchange (ETDEWEB)

    Seltborg, P.; Jinnestrand, M. [Westinghouse Electric Sweden AB, SE-721 63 Vaesteraas (Sweden)

    2012-07-01

    A long-term program assessing the mechanical performance of the Westinghouse BWR control rod CR 99 at high depletion levels has been performed. The scope of the program has mainly been based on the operation of four CR 99 Generation 2 control rods in demanding positions during 6 and 7 cycles in the Leibstadt Nuclear Power Plant (KKL) and on the detailed visual inspections and blade wing thickness measurements that were performed after the rods were discharged. By correlating statistically the blade wing thickness measurements to the appearance of irradiation-assisted stress corrosion cracking (IASCC), the probability of IASCC appearance as function of the blade wing swelling was estimated. In order to correlate the IASCC probability of a CR 99 to its depletion, the {sup 10}B depletion of the studied rods was calculated in detail on a local level with the stochastic Monte Carlo code MCNP in combination with the Westinghouse nodal code system PHOENIX4/POLCA7. Using this information coupled to the blade wing measurement data, a finite element model describing the blade wing swelling of an arbitrary CR 99 design as function of {sup 10}B depletion could then be generated. In the final step, these relationships were used to quantify the probability of IASCC appearance as function of the {sup 10}B depletion of the CR 99 Generations 2 and 3. Applying this detailed mapping of the CR 99 behavior at high depletion levels and using an on-line core monitoring system with explicit {sup 10}B depletion tracking capabilities will enable a reliable prediction of the probability for IASCC appearance, thus enhancing the optimized design and the sound operation of the CR 99 control rod. Another important outcome of the program was that it was clearly shown that no significant amount of boron leakage did occur through any of the detected IASCC cracks, despite the very high depletion levels achieved. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  5. LOCA steam condensation loads in BWR Mark II pressure suppression containment system

    Energy Technology Data Exchange (ETDEWEB)

    Kukita, Y.; Namatame, K.; Takeshita, I.; Shiba, M.

    1987-06-01

    Hydrodynamic loads induced in the BWR Mark II pressure suppression containment system during a loss-of-coolant accident (LOCA) were investigated using a large scale test facility. The maximum-bounding loading conditions on the pressure suppression pool-boundary structures were defined by conducting experiments for a wide range of parameters. The maximum-bounding loads occurred when steam, with air concentration less than 1% in weight, was injected at moderate rates (approx. = 30 kg/m/sup 2/.s) into a low-temperature (below 310 K) pool. Such conditions are most likely to be encountered during LOCAs with intermediate break sizes.

  6. LOCA air-injection loads in BWR Mark II pressure suppression containment systems

    Energy Technology Data Exchange (ETDEWEB)

    Kukita, Y.; Shiba, M. (Japan Atomic Energy Research Inst., Tokai, Ibaraki); Namatame, K. (Institute of Nuclear Safety, Tokyo (Japan))

    1984-02-01

    Large-scale blowdown tests were conducted to investigate the thermal-hydrodynamic response of a boiling-water reactor (BWR) Mark II pressure suppression containment system to a postulated loss-of-coolant accident. This paper presents the test results on the early blowdown transients, where air in the drywell is injected into the pressure suppression pool and induces various hydrodynamic loads onto the containment pressure boundary and internal structures. The test data are compared to predictions by analytical models used for the licensing evaluation of the hydrodynamic loads to assess these models.

  7. Effect of non-heterogeneous wetwell boundaries on pressure suppression system response. [BWR

    Energy Technology Data Exchange (ETDEWEB)

    McCauley, E.W.; Holman, G.S.; Namatame, K.; Kukita, Y.; Shiba, M.

    1980-08-29

    The Full-Scale Mark II CRT (Containment Response Test) Program is in progress at the Tokai-Mura Establishment of the Japan Atomic Energy Research Institute (JAERI). The primary objective of the on-going CRT Program is to provide a data base for evaluation of the pressure suppression pool (wetwell) hydrodynamic loads associated with a postulated loss-of-coolant accident (LOCA) in the BWR Mark II containment system. The test facility is 1/18 of full scale in volume and has a wetwell which is a full-scale geometric replica of one 20/sup 0/-sector of a reference 1100MWe Mark II.

  8. A Mechanistic Approach for the Prediction of Critical Power in BWR Fuel Bundles

    Science.gov (United States)

    Chandraker, Dinesh Kumar; Vijayan, Pallipattu Krishnan; Sinha, Ratan Kumar; Aritomi, Masanori

    The critical power corresponding to the Critical Heat Flux (CHF) or dryout condition is an important design parameter for the evaluation of safety margins in a nuclear fuel bundle. The empirical approaches for the prediction of CHF in a rod bundle are highly geometric specific and proprietary in nature. The critical power experiments are very expensive and technically challenging owing to the stringent simulation requirements for the rod bundle tests involving radial and axial power profiles. In view of this, the mechanistic approach has gained momentum in the thermal hydraulic community. The Liquid Film Dryout (LFD) in an annular flow is the mechanism of CHF under BWR conditions and the dryout modeling has been found to predict the CHF quite accurately for a tubular geometry. The successful extension of the mechanistic model of dryout to the rod bundle application is vital for the evaluation of critical power in the rod bundle. The present work proposes the uniform film flow approach around the rod by analyzing individual film of the subchannel bounded by rods with different heat fluxes resulting in different film flow rates around a rod and subsequently distributing the varying film flow rates of a rod to arrive at the uniform film flow rate as it has been found that the liquid film has a strong tendency to be uniform around the rod. The FIDOM-Rod code developed for the dryout prediction in BWR assemblies provides detailed solution of the multiple liquid films in a subchannel. The approach of uniform film flow rate around the rod simplifies the liquid film cross flow modeling and was found to provide dryout prediction with a good accuracy when compared with the experimental data of 16, 19 and 37 rod bundles under BWR conditions. The critical power has been predicted for a newly designed 54 rod bundle of the Advanced Heavy Water Reactor (AHWR). The selected constitutive models for the droplet entrainment and deposition rates validated for the dryout in tube were

  9. 40 CFR 86.1818-12 - Greenhouse gas emission standards for light-duty vehicles, light-duty trucks, and medium-duty...

    Science.gov (United States)

    2010-07-01

    ... light-duty vehicles, light-duty trucks, and medium-duty passenger vehicles. 86.1818-12 Section 86.1818... vehicles, light-duty trucks, and medium-duty passenger vehicles. (a) Applicability. This section contains... vehicles, light-duty trucks, and medium-duty passenger vehicles. Manufacturers that qualify as a small...

  10. Component failures that lead to reactor scrams. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Burns, E. T.; Wilson, R. J.; Lim, E. Y.

    1980-04-01

    This report summarizes the operating experience scram data compiled from 35 operating US light water reactors (LWRs) to identify the principal components/systems related to reactor scrams. The data base utilized to identify the scram causes is developed from a EPRI-utility sponsored survey conducted by SAI coupled with recent data from the USNRC Gray Books. The reactor population considered in this evaluation is limited to 23 PWRs and 12 BWRs because of the limited scope of the program. The population includes all the US NSSS vendors. It is judged that this population accurately characterizes the component-related scrams in LWRs over the first 10 years of plant operation.

  11. Effect of sensitization and cold work on stress corrosion susceptibility of austenitic stainless steels in BWR and PWR conditions

    Energy Technology Data Exchange (ETDEWEB)

    Haenninen, H.; Aho-Mantila, I.

    1981-05-01

    The influence of metallurgical variables on stress corrosion cracking of austenitic stainless steels, in particular AISI 304 and OX18H10T, has been examined both in O2-enriched BWR-conditions (8 ppm O2) and in typical PWR-conditions.

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

    Energy Technology Data Exchange (ETDEWEB)

    Behringer, K.; Hennig, D

    2002-11-01

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

  13. High-fidelity multiphysics simulation of BWR assembly with coupled TORT-TD/CTF

    Energy Technology Data Exchange (ETDEWEB)

    Magedanz, J. [Dept. of Mechanical and Nuclear Engineering, Pennsylvania State Univ., Reber Building, Univ. Park, PA 16802 (United States); Perin, Y. [Gesellschaft fuer Anlagen- und Reaktorsicherheit GRS mbH, Forschungsinstitute, D-85748 Garching (Germany); Avramova, M. [Dept. of Mechanical and Nuclear Engineering, Pennsylvania State Univ., Reber Building, Univ. Park, PA 16802 (United States); Pautz, A.; Puente-Espel, F.; Seubert, A.; Sureda, A.; Velkov, K.; Zwermann, W. [Gesellschaft fuer Anlagen- und Reaktorsicherheit GRS mbH, Forschungsinstitute, D-85748 Garching (Germany)

    2012-07-01

    This paper describes the application of the coupled codes TORT-TD and CTF to the pin-by-pin modeling of a BWR fuel assembly with thermal-hydraulic feedback. TORT-TD, developed at GRS, is a time-dependent three dimensional discrete ordinates code. CTF is the PSU's improved version of the subchannel code COBRA-TF, which uses a two-fluid, three-field model to represent two-phase flow with entrained droplets, and is commonly applied to evaluate LWR safety margins. The coupled codes system TORT-TD/CTF, already applied to several PWR cases involving MOX, was adapted from PWR to BWR applications. The purpose of the research described in this paper is to verify the coupling for modeling two-phase flow at the pin cell level. Using an ATRIUM-10 assembly, the system's steady-state capabilities were tested on two cases: one without control blade insertion and another with partially inserted blades. The influence of the neutron absorber on local axial and radial parameters is presented. The description of an inlet flow reduction transient is an example for the time-dependent capability of the coupled system. (authors)

  14. Investigation of control rod worth and nuclear end of life of BWR control rods

    Energy Technology Data Exchange (ETDEWEB)

    Magnusson, Per

    2008-01-15

    This work has investigated the Control Rod Worth (CRW) and Nuclear End of Life (NEOL) values for BWR control rods. A study of how different parameters affect NEOL was performed with the transport code PHOENIX4. It was found that NEOL, expressed in terms of {sup 10}B depletion, can be generalized beyond the conditions for which the rod is depleted, such as different power densities and void fractions, the corresponding variation in the NEOL will be about 0.2-0.4% {sup 10}B. It was also found that NEOL results for different fuel types and different fuel enrichments have a variation of about 2-3% in {sup 10}B depletion. A comparative study on NHOL and CRW was made between PHOENIX4 and the stochastic Monte Carlo code MCNP. It was found that there is a significant difference, both due to differences in the codes and to limitations in the geometrical modeling in PHOENIX4. Since MCNP is considered more physically correct, a methodology was developed to calculate the nuclear end of life of BWR control rods with MCNP. The advantages of the methodology are that it does not require other codes to perform the depletion of the absorber material, it can describe control rods of any design and it can deplete the control rod absorber material without burning the fuel. The disadvantage of the method is that is it time-consuming.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-08-15

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

  16. Identification and initial assessment of candidate BWR late-phase in-vessel accident management strategies

    Energy Technology Data Exchange (ETDEWEB)

    Hodge, S.A.

    1991-04-15

    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.

  17. Structural mechanics program: progress in 1981. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Tagart, S.W. Jr.; Marston, T.U.; Nickell, R.E.; Norris, D.M.

    1982-10-01

    The goal of the EPRI Structural Mechanics Program is to improve nuclear plant reliability and availability. The program is directed toward characterization of materials, evaluation and analysis of flaws, and application and technology transfer. There are fourteen topics involving more than forty separate contracts. The largest efforts are: (1) the continuation of projects aimed at developing a valid radiation embrittlement data base for evaluating the fracture toughness of irradiated pressure vessel steels; (2) the development of weld repair procedures for reactor pressure vessels as alternatives to the half bead repair method; (3) the development of simplified design methodology for the prediction of crack initiation, stable crack growth, and instability of ductile material in the presence of flaws; and (4) evaluation of the thermal anneal remedy for reactor pressure vessel irradiation damage. The significant progress made in 1981 in this program is reviewed and the interrelationships of the projects are discussed.

  18. ''Last experiences on ID BWR shroud inspection and the new developments to examine the below core plate areas''

    Energy Technology Data Exchange (ETDEWEB)

    Alberdi, R.; Willke, A.; Gonzalez, E.; Yague, L

    2001-07-01

    In recent years, the owners of BWR type nuclear power plants have had to address new inspection requirements relating to the core shroud inside the reactor vessel, the aim of which is to contain the fuel assemblies and provide support for the structures located in the upper part of the reactor. The shroud consists of a cylinder measuring some 40-50 mm in thickness, manufactured from various sections of AISI-304 stainless steel and INCONEL, joined by vertical and circumferential welds. The appearance of unstable cracks in these welds would directly affect the structural integrity of the component and the safety of the plant. As regards access to the core shroud and to the surface to be examined, two alternatives might be considered: inspection from outside the component, moving along the so-called annulus between the reactor vessel wall and the component (OD inspection), or from the interior (ID inspection). With a view to addressing this problem, Tecnatom has in recent years launched several projects, grouped under the generic name TEIDE, in order to develop scanners and NDT techniques achieving the maximum inspection coverage of this component. As regards inspection techniques, the decision was taken to carry out acquisition simultaneously using both ultrasonics (UT) and eddy currents (ET). (author)

  19. Neutron flux calculation and fluence in the encircling of the core and vessel of a reactor BWR; Calculo de flujo neutronico y fluencia en la envolvente del nucleo y la vasija de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Martinez C, E.

    2011-07-01

    One of the main objectives related to the safety of any nuclear power plant, including the nuclear power plant of Laguna Verde is to ensure the structural integrity of reactor pressure vessel. To identify and quantify the damage caused by neutron irradiation in the vessel of any nuclear reactor, it is necessary to know both the neutron flux and the neutron fluence that the vessel has been receiving during its operation lifetime, and that the damage observed by mechanical testing are products of microstructural effects induced by neutron irradiation; therefore, it is important the study and prediction of the neutron flux in order to have a better understanding of the damage that these materials are receiving. The calculation here described uses the DORT code, which solves the neutron transport equation in discrete ordinates in two dimensions (x-y, r-{theta} and r-z), according to a regulatory guide, it should make an approximation of the neutron flux in three dimensions by the so called synthesis method. It is called in that way because it achieves a representation of 3 Dimensional neutron flux combining or summarizing the fluxes calculated by DORT r-{theta}, r-z and r. This work presents the application of synthesis method, according to Regulatory Guide 1190, to determine the 3 Dimensional fluxes in internal BWR reactor using three different spatial meshes. The results of the neutron flux and fluence, using three different meshes in the directions r, {theta} and z were compared with results reported in the literature obtaining a difference not larger than 9.61%, neutron flux reached its maximum, 1.58 E + 12 n/cm{sup 2}s, at a height H 4 (239.07 cm) and angle 32.236{sup o} in the core shroud and 4.00 E + 09 n/cm{sup 2}s at a height H 4 and angle 35.27{sup o} in the inner wall of the reactor vessel, positions that are consistent to within {+-}10% over the ones reported in the literature. (Author)

  20. Product audit for heavy duty diesel engines in production environment

    Science.gov (United States)

    Suh, Sanghoon; Beresford, Jim

    2005-09-01

    A product audit at manufacturing plants has become more important due to the customer's requirements on product quality. Noise and vibration performance have been a primary concern for gas engines and small size diesel engines. Lately, more interest has been shown by truck manufacturers about engine noise for heavy duty diesel application. It has been regarded that acoustic measurements requires dedicated measurement environment for detailed study. This case study shows that acoustic measurements can be performed at performance cell without any dedicated acoustic treatment at the manufacturing plant to identify some of the noise characteristics with proper preparation. Order tracking and loudness were used to identify two different characteristics related to front gear train in heavy duty diesel engines. In addition, the coordination between technical organization and manufacturing plant for the data acquisition and analysis is discussed.

  1. Possibilities with OHWC. Development and application of ECP-simulation in Swedish BWRs; Moejligheter med OHWC. Utveckling och tillaempning av ECP-simulering i svenska BWR

    Energy Technology Data Exchange (ETDEWEB)

    Lundgren, K. [ALARA Engineering, Skultuna (Sweden); Wikmark, G. [Advanced Nuclear Technology, Uppsala (Sweden)

    2000-02-01

    Hydrogen injection (HWC) to boiling water reactors has been used for two decades in Sweden, in order to reduce the impact of pipe cracking. The effect of HWC is to establish a sufficiently reducing environment in the systems to protect and hence mitigate the growth of existing stress corrosion cracks. Some disadvantages of HWC have been identified. One is the transitional increase of the dose rate of the main steam lines by up to seven times, another the corrosion release of systems with carbon steel components as a result of the reducing chemistry. In some cases, especially in the USA, an elevated activity build-up has been observed in a few plants in connection to the application of HWC. There is also a fear for increased hydrogen pick-up in fuel cladding and fuel channels by HWC operation. The hydrogen pick-up is already today in many cases limiting for fuel life. The objective of the current work has been to investigate the conditions by application of so called Optimised HWC. This implies a HWC operation with lower hydrogen addition rates than normally used. For this purpose, a computer model in order to simulate the radiolysis chemistry and the ECP (electrochemical corrosion potentials) in BWR systems has been developed. A previously developed radiolysis code, BwrChem, as well as a hydrogen peroxide decomposition code for piping, PEROX, have hence been equipped with ECP calculation modules. The ECP calculation algorithms have been based on fundamental electrochemical theory. The new model has been applied to simulate the radiolysis conditions in a large number of locations in typical BWRs. For the simulation, the external mechanical pump plant Barsebaeck-1 and the internal pump plant Forsmark-1 have been used. A wide range of hydrogen injection rates, down to 0. 1 ppm in the feed water, have been studied. The electrochemical model based on fundamental theory required adequate fundamental parameters. Significant effort has been used to scrutinise and evaluate

  2. Supervision Duty of School Principals

    Directory of Open Access Journals (Sweden)

    Kürşat YILMAZ

    2009-04-01

    Full Text Available Supervision by school administrators is becoming more and more important. The change in the roles ofschool administrators has a great effect on that increase. At present, school administrators are consideredmore than as technical directors, but as instructional leaders. This increased the importance of schooladministrators’ expected supervision acts. In this respect, the aim of this study is to make a conceptualanalysis about school administrators’ supervision duties. For this reason, a literature review related withsupervision and contemporary supervision approaches was done, and the official documents concerningsupervision were examined. As a result, it can be said that school administrators’ supervision duties havebecome very important. And these duties must certainly be carried out by school administrators.

  3. 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)

  4. 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)

  5. 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)

  6. Experimental simulation of the water cooling of corium spread over the floor of a BWR containment

    Energy Technology Data Exchange (ETDEWEB)

    Morage, F.; Lahey, R.T. Jr.; Podowski, M.Z. [Rensselaer Polytechnic Institute, Troy, NY (United States)

    1995-09-01

    This paper is concerned with an experimental investigation of the cooling effect of water collected on the surface of corium released onto the floor of a BWR drywell. In the present experiments, the actual reactor materials were replaced by simulant materials. Specifically, the results are shown for Freon-11 film boiling over liquid Wood`s metal spread above a solid porous surface through which argon gas was injected. An analysis of the obtained experimental data revealed that the actual film boiling heat transfer between a molten pool of corium and the water above the pool should be more efficient than predicted by using standard correlations for boiling over solid surfaces. This effect will be further augmented by the gas released due to the ablation of concrete floor beneath the corium and percolating towards its upper surface and into through the water layer above.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-15

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

  8. Revised uranium--plutonium cycle PWR and BWR models for the ORIGEN computer code

    Energy Technology Data Exchange (ETDEWEB)

    Croff, A. G.; Bjerke, M. A.; Morrison, G. W.; Petrie, L. M.

    1978-09-01

    Reactor physics calculations and literature searches have been conducted, leading to the creation of revised enriched-uranium and enriched-uranium/mixed-oxide-fueled PWR and BWR reactor models for the ORIGEN computer code. These ORIGEN reactor models are based on cross sections that have been taken directly from the reactor physics codes and eliminate the need to make adjustments in uncorrected cross sections in order to obtain correct depletion results. Revised values of the ORIGEN flux parameters THERM, RES, and FAST were calculated along with new parameters related to the activation of fuel-assembly structural materials not located in the active fuel zone. Recommended fuel and structural material masses and compositions are presented. A summary of the new ORIGEN reactor models is given.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1985-04-01

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

  10. Performance of the primary containment of a BWR during a severe accident whit the code RELAP/SCDAPSIM; Comportamiento del contenedor primario de un reactor BWR durante un accidente severo con el codigo RELAP/SCDAPSIM

    Energy Technology Data Exchange (ETDEWEB)

    Castillo G, F.

    2015-07-01

    In this thesis work, it was developed a model of the vacuum breaker valves and down comers for a BWR Mark II primary containment for the code RELAP/SCDAPSIM Mod. 3.4. This code was used to simulate a Station Blackout (Sbo) that evolves to a severe accident scenario. To accomplish this task, the vacuum breaker valves and down comers were included in a simplified model of the primary containment that includes both wet well and dry well, which was coupled with a model of the Nuclear Steam Supply System (NSSS), in order to study the behavior of the primary containment during the evolution of the accident scenario. In the analysis of the results of the simulation, the behavior of the wet well and dry well during the event was particularly monitored, by analyzing the evolution of temperature and pressure profiles in such volumes, this to determine the impact of the inclusion of the breaker vacuum valves and down comers. The results show that the effect of this extension of the model is that more conservative results are obtained, i.e., higher pressures are reached in both wet well and dry well than when it is used a containment model that does not include neither the vacuum valves nor the down comers. The most relevant results obtained show that the Rcic alone is able to keep the core fully covered, but even in such a case, it evaporates about 15% of the initial inventory of liquid water in the Pressure Suppression Pool (Psp). When the Rcic operation is lost, 20% more of the liquid water inventory in the Psp is further reduced within four to twelve hours (approximately), time at which the simulation crashed. Besides, there is a significant increase of pressure in the containment. As the accident evolves, the pressure in the containment continues increasing, but there is still considerable margin to reach the design pressure of the containment. At the end of the simulation, the results show a gauge pressure value of 224,550 Pa in the Psp and 187,482 Pa in the wet well

  11. Cause-Effect relationship of the Laguna Verde BWR power instability by empirical mode decomposition; Relacion efecto-causa de la inestabilidad de potencia del BWR de Laguna Verde por descomposicion modal empirica

    Energy Technology Data Exchange (ETDEWEB)

    Blazquez, J.; Ruiz, J.; Castillo, R. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)

    2008-07-01

    The signals coming from natural phenomena are in essence non lineal and not stationary. A recent development, well-known as Empirical Mode Decomposition (EMD) it presents a novel focus that allows to represent in adaptative form non stationary signals as a sum of components of half zero. These components denominated Intrinsic Mode Functions (IMF) they help to the analysis of the frequency composition of unidimensional signals. The use of the EMD followed by the Hilbert transform of the IMFs it allows to carry out an analysis in time-frequency of the non lineal and not stationary data. This technique is known as the Hilbert Huang Transform (HHT). In this work a power instability event occurred in January 24, 1995 in the unit I of the nuclear power station of Laguna Verde (Mexico), corresponding to a BWR/5 is analyzed. When a Nuclear Plant suffers a power instability event, it is required obligatorily to explain to the Regulator Organism the effects and the causes of the event. The effects are described simply; not in vain there is a registration of signals in the Process Computer of where the required information is extracted. But the causes are not always immediate and easy for to identify. The power instability can happen during the start, when the refrigeration flow is relatively low in front to the power. By reason of that the reactivity coefficient by holes is negative, the power oscillates with a very defined frequency, generally of the order of 0.5 Hz. If the oscillations increase progressively of amplitude, we are in an instability event. It is interesting to include in the report the instant in that the began instability and the actions of the operator before and after the same one. As the actions are registered, the investigation is focused toward the instant of the beginning to be able to identify them. In this work the power signal in five empiric ways of Hilbert-Huang and a residual breaks down. The instability is only reflected in the way of smaller

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

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

  14. Statistical Safety Evaluation of BWR Turbine Trip Scenario Using Coupled Neutron Kinetics and Thermal Hydraulics Analysis Code SKETCH-INS/TRACE5.0

    Science.gov (United States)

    Ichikawa, Ryoko; Masuhara, Yasuhiro; Kasahara, Fumio

    The Best Estimate Plus Uncertainty (BEPU) method has been prepared for the regulatory cross-check analysis at Japan Nuclear Energy Safety Organization (JNES) on base of the three-dimensional neutron-kinetics/thermal- hydraulics coupled code SKETCH-INS/TRACE5.0. In the preparation, TRACE5.0 is verified against the large-scale thermal-hydraulic tests carried out with NUPEC facility. These tests were focused on the pressure drop of steam-liquid two phase flow and void fraction distribution. From the comparison of the experimental data with other codes (RELAP5/MOD3.3 and TRAC-BF1), TRACE5.0 was judged better than other codes. It was confirmed that TRACE5.0 has high reliability for thermal hydraulics behavior and are used as a best-estimate code for the statistical safety evaluation. Next, the coupled code SKETCH-INS/TRACE5.0 was applied to turbine trip tests performed at the Peach Bottom-2 BWR4 Plant. The turbine trip event shows the rapid power peak due to the voids collapse with the pressure increase. The analyzed peak value of core power is better simulated than the previous version SKETCH-INS/TRAC-BF1. And the statistical safety evaluation using SKETCH-INS/TRACE5.0 was applied to the loss of load transient for examining the influence of the choice of sampling method.

  15. Study of intermediate configurations during the fuel reload in BWRs; Estudio de configuraciones intermedias durante la recarga de combustible en BWR's

    Energy Technology Data Exchange (ETDEWEB)

    Fuentes M, L.; Castillo M, J. A.; Ortiz S, J. J.; Perusquia del C, R. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Jacinto C, S., E-mail: luis.fuentes@inin.gob.mx [Universidad Autonoma del Estado de Yucatan, Calle 60 No. 491-A por 57, 97000 Merida, Yucatan (Mexico)

    2012-10-15

    The criticality state of the core of a boiling water reactor (BWR) was evaluated, during the reload process for the intermediate states between the load pattern of cycle end and the beginning of the next, using the information of the load pattern of the operation cycles 13 and 14 of Unit 1 of the nuclear power plant of Laguna Verde. For this evaluation the codes CASMO-4 and Simulate-3 for conditions of the core in cold were used. The strategy consisted on moving assemblies with 4 burned cycles of the reactor core. Later on were re situated the remaining assemblies, placing them in the positions to occupy in the next operation cycle. Finally, was carried out the assemblies load of fresh fuel. In each realized change, it was observing the behavior of the k-effective value that is the parameter used to evaluate the criticality state of each state of the core change. In a second stage, was designed a program that builds in automatic way each one of the intermediate cores and also analyzes the criticality state of the reactor core after each withdrawal, re situated and load of fuel assemblies. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  17. THE INDEFENSIBLE DUTY TO DEFEND

    National Research Council Canada - National Science Library

    Neal Devins; Saikrishna Prakash

    2012-01-01

    .... First, the duties to enforce and defend lack any sound basis in the Constitution. Hence, while President Obama is right to refuse to defend the Defense of Marriage Act, he is wrong to continue to enforce a law he believes is unconstitutional...

  18. Implementation of a methodology to perform the uncertainty and sensitivity analysis of the control rod drop in a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Reyes F, M. del C.

    2015-07-01

    A methodology to perform uncertainty and sensitivity analysis for the cross sections used in a Trace/PARCS coupled model for a control rod drop transient of a BWR-5 reactor was implemented with the neutronics code PARCS. A model of the nuclear reactor detailing all assemblies located in the core was developed. However, the thermohydraulic model designed in Trace was a simple model, where one channel representing all the types of assemblies located in the core, it was located inside a simple vessel model and boundary conditions were established. The thermohydraulic model was coupled with the neutronics model, first for the steady state and then a Control Rod Drop (CRD) transient was performed, in order to carry out the uncertainty and sensitivity analysis. To perform the analysis of the cross sections used in the Trace/PARCS coupled model during the transient, Probability Density Functions (PDFs) were generated for the 22 parameters cross sections selected from the neutronics parameters that PARCS requires, thus obtaining 100 different cases for the Trace/PARCS coupled model, each with a database of different cross sections. All these cases were executed with the coupled model, therefore obtaining 100 different outputs for the CRD transient with special emphasis on 4 responses per output: 1) The reactivity, 2) the percentage of rated power, 3) the average fuel temperature and 4) the average coolant density. For each response during the transient an uncertainty analysis was performed in which the corresponding uncertainty bands were generated. With this analysis it is possible to observe the results ranges of the responses chose by varying the uncertainty parameters selected. This is very useful and important for maintaining the safety in the nuclear power plants, also to verify if the uncertainty band is within of safety margins. The sensitivity analysis complements the uncertainty analysis identifying the parameter or parameters with the most influence on the

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-12-01

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

  20. VIPRE-W / MEFISTO-T - A mechanistic tool for transient prediction of dryout in BWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Adamsson, C., E-mail: carl.adamsson@psi.ch [Westinhouse Electric Sweden, Vasteras (Sweden); Paul Scherrer Institut, Villigen (Switzerland); Le Corre, J-M., E-mail: lecorrjm@westinghouse.com [Westinhouse Electric Sweden, Vasteras (Sweden)

    2011-07-01

    The VIPRE-W/MEFISTO-T code package constitutes a simplified approach to sub-channel film-flow analysis whereby the transport equations for the liquid films are decoupled from each other. The approach allows fast and robust simulation with high axial resolution of realistic BWR transients. It has previously been shown that a steady-state version of the model agrees well with dryout measurements in full-scale fuel assembly mock-ups performed at the Westinghouse FRIGG loop. In this paper, we present validation of the transient version of the code with around 300 transient dryout experiments from the same loop. The transients involve realistic variations of flow and power and three different axial power distributions at conditions typical for BWR operation. The results from the film-flow analysis show high precision in the dryout prediction but a hitherto unexplained bias that reduces the accuracy. (author)

  1. Duties: Legal? Moral? Religious? or Social?

    Science.gov (United States)

    Blum, Ann

    1990-01-01

    Presents activities in which students are asked to (1) identify sources of duties affecting individual behavior; (2) define and give examples of legal, as well as social, religious and moral duties; (3) and compare social, religious, moral, and legal duties and discuss their relationships. (DB)

  2. Stakes and Solutions for current and up-coming Licensing Challenges in PWR and BWR Reload and Safety Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Curca-Tiving, F.; Opel, S.

    2014-07-01

    Regulatory requirements for reloads and safety analyses are evolving: New safety criteria, requests for enlarged qualification databases, statistical applications, uncertainty propagation... In order to address these challenges and access more predictable licensing processes, AREVA implements a consistent code and methodology suite for PWR and BWR core design and safety analysis, based on a first principles modeling with an extremely broad international verification and validation data base. (Author)

  3. Evaluation of the thermal-mechanical performance of fuel rods of a BWR during a power ramp using the FUELSIM code; Evaluacion del desempeno termomecanico de barras de combustible de un reactor BWR durante una rampa de potencia utilizando el codigo FUELSIM

    Energy Technology Data Exchange (ETDEWEB)

    Pantoja C, R.

    2010-07-01

    To avoid the risk to environment due to release of radioactive material, because of occurrence of an accident, it is the priority of the design and performance of the diverse systems of safety of a commercial nuclear power plant. The safety of nuclear power plants requires, therefore, monitoring those parameters having some direct or indirect effect on safety. The thermal limits are values set for those parameters considered having most impact on the safe operation of a nuclear power reactor. Some thermal limits monitoring requires the thermal-mechanical analysis of the rods containing the nuclear fuel. The fuel rod thermal-mechanical behavior under irradiation is a complex process in which there exists a great deal of interrelated physical and chemical phenomena, so that the fuel rod performance analysis in the core of a nuclear power reactor is generally accomplished by using computer codes, which integrate several of the phenomena that are expected to occur during the lifetime of the fuel rod in the core. The main application of the thermal-mechanical analysis codes is the prediction of occurrence of conditions and/or phenomena that could lead to the deterioration or even mechanical failure of the fuel rod cladding, as, for example, the pellet-cladding interaction. In the operation of a nuclear power reactor, fuel preconditioning operations refer to the operational procedures employed to reduce the fuel rod failure probability due to fuel-cladding interaction, specially during reactor startup. Preconditioning simulations are therefore necessary to determine in advance limit values for the power that can be generated in a fuel rod, and thus avoiding any rod damage. In this work, a first analysis of the thermal-mechanical performance of typical fuel rods used in nuclear reactors of the type BWR 5/6, as those two nuclear reactors in Laguna Verde, Veracruz, is performed. This study includes two types of fuel rods: one from a fuel assembly design with an array 8 x 8

  4. 3D simulation of a core operation cycle of a BWR using Serpent; Simulacion 3D de un ciclo de operacion del nucleo de un BWR usando SERPENT

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

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

  7. 19 CFR 141.1 - Liability of importer for duties.

    Science.gov (United States)

    2010-04-01

    ... 19 Customs Duties 2 2010-04-01 2010-04-01 false Liability of importer for duties. 141.1 Section... Merchandise § 141.1 Liability of importer for duties. (a) Time duties accrue. Duties and the liability for... for by law. (b) Payment of duties—(1) Personal debt of importer. The liability for duties,...

  8. The new duty of care for nuclear power plant operators in Sec. 9a subpara. 2a AtG; Zur neuen Sorgepflicht der Kraftwerksbetreiber gem. paragraph 9a Abs. 2a AtG

    Energy Technology Data Exchange (ETDEWEB)

    Posser, Herbert [Freshfields Bruckhaus Deringer LLP, Duesseldorf (Germany)

    2014-07-15

    The new stipulation in Sec. 9a subpara. 2a AtG - pursuant to which operators of nuclear power plants are no longer entitled to use the interim storage facility in Gorleben for radioactive waste stemming from the reprocessing plants in Sellafield and La Hague, but have to establish further capacities in their own facilities for spent nuclear fuels at the site of the power plants - is illegal under constitutional law. It imposes an unproportional burden on the plant operators as well as on GNS, and infringes property rights without pursuing a legitimate purpose. (orig.)

  9. Heavy Duty Vehicle Futures Analysis.

    Energy Technology Data Exchange (ETDEWEB)

    Askin, Amanda Christine; Barter, Garrett.; West, Todd H.; Manley, Dawn Kataoka

    2014-05-01

    This report describes work performed for an Early Career Research and Development project. This project developed a heavy-duty vehicle (HDV) sector model to assess the factors influencing alternative fuel and efficiency technology adoption. This model builds on a Sandia light duty vehicle sector model and provides a platform for assessing potential impacts of technological advancements developed at the Combustion Research Facility. Alternative fuel and technology adoption modeling is typically developed around a small set of scenarios. This HDV sector model segments the HDV sector and parameterizes input values, such as fuel prices, efficiencies, and vehicle costs. This parameterization enables sensitivity and trade space analyses to identify the inputs that are most associated with outputs of interest, such as diesel consumption and greenhouse gas emissions. Thus this analysis tool enables identification of the most significant HDV sector drivers that can be used to support energy security and climate change goals.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-01

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

  11. Advanced water processing system (AWPS), including advanced filtration system (AFS) and advanced ion selective system (AISS) for improved utility (PWR/BWR) water processing performance

    Energy Technology Data Exchange (ETDEWEB)

    Denton, Mark S. [ATG, Inc.(United States); Vance, Jene N. [V and V, Inc. (United States)

    1999-07-01

    The advanced water processing system (AWPS) has the potential for wide spread success on a worldwide scale in both PWR and BWRs. The AWPS incorporates the advanced features (patent pending) of advanced filtration and advanced ion selective technologies (patented). Typical problems encountered in current filtration systems include: (1) poor effluent quality, (2) short run lengths on filters, (3) frequent filter change-outs/backwashes, (4) large waste volumes, and (5) failed filter cartridges. The advanced filtration system (AFS) features reduced waste production per million gallons of water processed, cleaner water for recycle or release to the environment, filter element volume 100 times less than that of competitive filters, and a far lower capital cost compared to systems with similar performance. The AWPS should be of interest to plants that are upgrading, or to new plants to lower both their capital and operating costs, as well as total curie discharge levels. In addition, the AWPS will function in non-nuclear, as well as nuclear, applications of water purification, specially where pre coat filtration/ion exchange or reverse osmosis (RO) is being applied to process water with high concentrations of colloidal contaminants. Pilot testing has been successfully completed in the U. S. at the Byron (PWR), LaSalle (BWR), and Dresden(BWR) nuclear plants for Commonwealth Edison, and the Bruce several spent filters in a High Integrated Container these bench- and pilot-scale demonstrations will be presented herein. Full-scale designs or systems have been shipped to these locations. In all cases, the testing demonstrated: (1) longer run lengths (300,000 gallons between backwashes--a 100 fold improvement), (2) recoverability of cartridge filters after backwash (cartridge lives of approximately 6 months to a year--a 5 to 10 fold improvement in filter life), (3) large removal efficiencies for colloidal particles (reduced discharge curies), and (4) reduced waste volumes

  12. Knowledge and abilities catalog for nuclear power plant operators: Boiling water reactors, Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-08-01

    The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWRs) (NUREG-1123, Revision 1) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog along with the Operator Licensing Examiner Standards (NUREG-1021) and the Examiner`s Handbook for Developing Operator Licensing Written Examinations (NUREG/BR-0122), will cover the topics listed under Title 10, Code of Federal Regulations, Part 55 (10 CFR 55). The BWR Catalog contains approximately 7,000 knowledge and ability (K/A) statements for ROs and SROs at BWRs. The catalog is organized into six major sections: Organization of the Catalog, Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Functions, Emergency and Abnormal Plant Evolutions, Components, and Theory. Revision 1 to the BWR Catalog represents a modification in form and content of the original catalog. The K/As were linked to their applicable 10 CFR 55 item numbers. SRO level K/As were identified by 10 CFR 55.43 item numbers. The plant-wide generic and system generic K/As were combined in one section with approximately one hundred new K/As. Component Cooling Water and Instrument Air Systems were added to the Systems Section. Finally, High Containment Hydrogen Concentration and Plant Fire On Site evolutions added to the Emergency and Abnormal Plant Evolutions section.

  13. 3D modeling of missing pellet surface defects in BWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, B.W., E-mail: Benjamin.Spencer@inl.gov; Williamson, R.L.; Stafford, D.S.; Novascone, S.R.; Hales, J.D.; Pastore, G.

    2016-10-15

    Highlights: • A global/local analysis procedure for missing pellet surface defects is proposed. • This is applied to defective BWR fuel under blade withdrawal and high power ramp conditions. • Sensitivity of the cladding response to key model parameters is studied. - Abstract: One of the important roles of cladding in light water reactor fuel rods is to prevent the release of fission products. To that end, it is essential that the cladding maintain its integrity under a variety of thermal and mechanical loading conditions. Local geometric irregularities in fuel pellets caused by manufacturing defects known as missing pellet surfaces (MPS) can in some circumstances lead to elevated cladding stresses that are sufficiently high to cause cladding failure. Accurate modeling of these defects can help prevent these types of failures. The BISON nuclear fuel performance code developed at Idaho National Laboratory can be used to simulate the global thermo-mechanical fuel rod behavior, as well as the local response of regions of interest, in either 2D or 3D. In either case, a full set of models to represent the thermal and mechanical properties of the fuel, cladding and plenum gas is employed. A procedure for coupling 2D full-length fuel rod models to detailed 3D models of the region of the rod containing a MPS defect is detailed here. The global and local model each contain appropriate physics and behavior models for nuclear fuel. This procedure is demonstrated on a simulation of a boiling water reactor (BWR) fuel rod containing a pellet with an MPS defect, subjected to a variety of transient events, including a control blade withdrawal and a ramp to high power. The importance of modeling the local defect using a 3D model is highlighted by comparing 3D and 2D representations of the defective pellet region. Parametric studies demonstrate the effects of the choice of gaseous swelling model and of the depth and geometry of the MPS defect on the response of the cladding

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-06-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Fuentes C, P

    2003-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-07-01

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

  17. High-speed simulation of transients in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

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

    1984-01-01

    A combination of advanced modeling techniques and modern, special-purpose peripheral minicomputer technology is presented which affords realistic predictions of plant transient and severe off-normal events in LWR power plants through on-line simulations at a speed ten times greater than actual process speeds. Results are shown for a BWR plant simulation. Results are shown to demonstrate computing capacity, accuracy, and speed. Simulation speeds have been achieved which are 110 times larger than those of a CDC-7600 mainframe computer or ten times greater than real-time speed.

  18. Decay Heat Calculations for PWR and BWR Assemblies Fueled with Uranium and Plutonium Mixed Oxide Fuel using SCALE

    Energy Technology Data Exchange (ETDEWEB)

    Ade, Brian J [ORNL; Gauld, Ian C [ORNL

    2011-10-01

    In currently operating commercial nuclear power plants (NPP), there are two main types of nuclear fuel, low enriched uranium (LEU) fuel, and mixed-oxide uranium-plutonium (MOX) fuel. The LEU fuel is made of pure uranium dioxide (UO{sub 2} or UOX) and has been the fuel of choice in commercial light water reactors (LWRs) for a number of years. Naturally occurring uranium contains a mixture of different uranium isotopes, primarily, {sup 235}U and {sup 238}U. {sup 235}U is a fissile isotope, and will readily undergo a fission reaction upon interaction with a thermal neutron. {sup 235}U has an isotopic concentration of 0.71% in naturally occurring uranium. For most reactors to maintain a fission chain reaction, the natural isotopic concentration of {sup 235}U must be increased (enriched) to a level greater than 0.71%. Modern nuclear reactor fuel assemblies contain a number of fuel pins potentially having different {sup 235}U enrichments varying from {approx}2.0% to {approx}5% enriched in {sup 235}U. Currently in the United States (US), all commercial nuclear power plants use UO{sub 2} fuel. In the rest of the world, UO{sub 2} fuel is still commonly used, but MOX fuel is also used in a number of reactors. MOX fuel contains a mixture of both UO{sub 2} and PuO{sub 2}. Because the plutonium provides the fissile content of the fuel, the uranium used in MOX is either natural or depleted uranium. PuO{sub 2} is added to effectively replace the fissile content of {sup 235}U so that the level of fissile content is sufficiently high to maintain the chain reaction in an LWR. Both reactor-grade and weapons-grade plutonium contains a number of fissile and non-fissile plutonium isotopes, with the fraction of fissile and non-fissile plutonium isotopes being dependent on the source of the plutonium. While only RG plutonium is currently used in MOX, there is the possibility that WG plutonium from dismantled weapons will be used to make MOX for use in US reactors. Reactor-grade plutonium

  19. Void Reactivity Coefficient Analysis during Void Fraction Changes in Innovative BWR Assemblies

    Directory of Open Access Journals (Sweden)

    Andrius Slavickas

    2015-01-01

    Full Text Available The study of the void reactivity variation in innovative BWR fuel assemblies is presented in this paper. The innovative assemblies are loaded with high enrichment fresh UO2 and MOX fuels. UO2 fuel enrichment is increased above existing design limitations for LWR fuels (>5%. MOX fuel enrichment with fissile Pu content is established to achieve the same burnup level as that of high enrichment UO2 fuel. For the numerical analysis, the TRITON functional module of SCALE 6.1 code with the 238-group ENDF/B-VI cross section data library was applied. The investigation of the void reactivity feedback is performed in the entire 0–100% void fraction range. Higher values of void reactivity coefficient for assembly loaded with MOX fuel are found in comparison with values for assembly loaded with UO2 fuel. Moreover, coefficient values for MOX fuel are positive over 75% void fraction. The variation of the void reactivity coefficient is explained by the results of the decomposition analysis based on four-factor formula and neutron absorption reactions for main isotopes. Additionally, the impact of the moderation enhancement on the void reactivity coefficient was investigated for the innovative assembly with MOX fuel.

  20. Decomposition Analysis of Void Reactivity Coefficient for Innovative and Modified BWR Assemblies

    Directory of Open Access Journals (Sweden)

    Andrius Slavickas

    2014-01-01

    Full Text Available The decomposition analysis of void reactivity coefficient for innovative BWR assemblies is presented in this paper. The innovative assemblies were loaded with high enrichment UO2 and MOX fuels. Additionally the impact of the moderation enhancement on the void reactivity coefficient through a full fuel burnup discharge interval was investigated for the innovative assembly with MOX fuel. For the numerical analysis the TRITON functional module of SCALE code with ENDF/B-VI cross section library was applied. The obtained results indicate the influence of the most important isotopes to the void reactivity behaviour over a fuel burnup interval of 70 GWd/t for both UO2 and MOX fuels. From the neutronic safety concern positive void reactivity coefficient values are observed for MOX fuel at the beginning of the fuel irradiation cycle. For extra-moderated assembly designs, implementing 8 and 12 water holes, the neutron spectrum softening is achieved and consequently the lower void reactivity values. Variations in void reactivity coefficient values are explained by fulfilled decomposition analysis based on neutrons absorption reactions for separate isotopes.

  1. Environmentally assisted cracking behavior of dissimilar metal weldments in simulated BWR coolant environments

    Science.gov (United States)

    Huang, J. Y.; Chiang, M. F.; Jeng, S. L.; Huang, J. S.; Kuo, R. C.

    2013-01-01

    The environmentally assisted cracking behavior of dissimilar metal (DM) welds, including Alloy 52-A 508 and Alloy 82-A508, under simulated BWR coolant conditions was studied. Effects of postweld heat treatment and sulfur content of the base metal on the corrosion fatigue and SCC growth rates of DM welds were evaluated. The crack growth rates for the DM weld heat-treated at 621 °C for 24 h were observed to be faster than those for the as-welded. But the DM weld heat-treated at 621 °C for 8 h + 400 °C for 200 h showed better SCC resistance than the as-welded. The longer the heat treatment at 621 °C, the higher the chromium carbides density along the grain boundary was observed. Sulfur could diffuse out of the base metal and segregate along the grain boundaries of the dilution zone, leading to weakening the grain boundary strength and the SCC resistance of the Alloy 52-A508 weld.

  2. Generic BWR-4 degraded core in-vessel study. Status report

    Energy Technology Data Exchange (ETDEWEB)

    1984-11-01

    Original intent of this project was to produce a phenomenological study of the in-vessel degradation which occurs during the TQUX and TQUV sequences for a generic BWR-4 from the initiation of the FSAR Chapter 15 operational transient through core debris bed formation to the failure of the primary pressure boundary. Bounding calculations were to be performed for the two high pressure and low pressure non-LOCA scenarios to assess the uncertainties in the current state of knowledge regarding the source terms for containment integrity studies. Source terms as such were defined in terms of hydrogen generation, unreacted metal, and coolant inventroy, and in terms of the form, sequencing and mode of dispersal through the primary vessel boundary. Fission product release was not to be considered as part of this study. Premature termination of the project, however, led to the dicontinuation of work on an as is basis. Work on the in-core phase from the point of scram to core debris bed formation was largely completed. A preliminary scoping calculation on the debris bed phase had been initiated. This report documents the status of the study at termination.

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

    Energy Technology Data Exchange (ETDEWEB)

    Sebrell, W.

    1983-07-01

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

  4. Analysis of the documents about the core envelopment of nuclear reactor at the Laguna Verde U-1 power plant; Analisis de documentos de los materiales de la envolvente del nucleo del reactor nuclear de la CLV U-1

    Energy Technology Data Exchange (ETDEWEB)

    Zamora R, L.; Medina F, A. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    1999-07-01

    The degradation of internal components at BWR type reactors is an important subject to consider in the performance availability of the power plant. The Wuergassen nuclear reactor license was confiscated due to the presence of cracking in the core envelopment. In consequence it is necessary carrying out a detailed study with the purpose to avoid these problems in the future. This report presents a review and analysis of documents and technical information referring to the core envelopment of a BWR/5/6 and the Laguna Verde Unit 1 nuclear reactor in Mexico. In this document are presented design data, documents about fabrication processes, and manufacturing of core envelopment. (Author)

  5. Fitness for duty in the nuclear power industry

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, N.; Moore, C.; Grant, T.; Fleming, T.; Hunt, P.; Martin, R.; Murphy, S.; Hauth, J.; Wilson, R.; Bittner, A.; Bramwell, A.; Macaulay, J.; Olson, J.; Terrill, E.; Toquam, J. (Battelle Human Affairs Research Center, Seattle, WA (United States))

    1991-09-01

    This report presents an overview of the NRC licensees' implementation of the FFD program during the first full year of the program's operation and provides new information on a variety of FFD technical issues. The purpose of this document is to contribute to appropriate changes to the rule, to the inspection process, and to other NRC activities. It describes the characteristics of licensee programs, discusses the results of NRC inspections, updates technical information covered in previous reports, and identifies lessons learned during the first year. Overall, the experience of the first full year of licensees' FFD program operations indicates that licensees have functioning fitness for duty programs devoted to the NRC rule's performance objectives of achieving drug-free workplaces in which nuclear power plant personnel are not impaired as they perform their duties. 96 refs., 14 tabs.

  6. Individual rights versus societal duties.

    Science.gov (United States)

    Vermeersch, E

    1999-10-29

    In 'bioethics', the rights to self-determination and to informed consent of the patient are prerequisites to every medical decision: paternalism is no longer a justifiable attitude. Hence, it seems that compulsory vaccination is an unacceptable praxis. Even John Stuart Mill. however, took into account other values: e.g. the duty not to harm others. This article is dedicated to the analysis of the historical development of these values and to their relevance for the ethics of vaccination. The acceptability of coercion is upheld, but no clear-cut answers are given in general: in every case the pros and cons of coercion are to be weighed carefully against each other.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-15

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

  8. The Duty to Rescue and Investigators' Obligations.

    Science.gov (United States)

    MacKay, Douglas; Rulli, Tina

    We examine current applications of the moral duty to rescue to justify clinical investigators' duties of ancillary care and standard of care to subjects in resource-poor settings. These applications fail to explain why investigators possess obligations to research participants, in particular, and not to people in need, in general. Further, these applications fail to recognize the normative significance of the institutional role of the investigators. We offer a positive account of the duty to rescue for investigators as institutional agents, with duties to populations rather than merely individuals.

  9. Development of mathematical models for the aero derivative and heavy duty gas turbines; Desenvolvimento de modelos matematicos para as turbinas a gas aeroderivativas e heavy duty

    Energy Technology Data Exchange (ETDEWEB)

    Freire, Marcelo; Mendes, Pedro Paulo de C.; Ferreira, Claudio; Passaro, Mauricio Campos; Gomes, Leonardo Vinicius [Escola Federal de Engenharia de Itajuba, MG (Brazil). Dept. de Eletronica]. E-mails: freire_marcelo@hotmail.com; ppaulo@iee.efei.br; claudio@iee.efei.br; mcpassaro@uol.com.br; leonardo@iee.efei.br

    2002-07-01

    This paper develops, implements and simulates simplified mathematical models of multiple shafts, aero derivatives and heavy-duty gas turbines, aiming the subsides for studies of power systems dynamic behaviour. These components are fundamental to an approximated evaluation of the National Integrated System after the new thermoelectric plants are incorporated.

  10. Optimization of analysis best-estimate of a fuel element BWR with Code STAR-CCM+; Optimizacion del analisis best-estimate de un elemento combustible BWR con el codigo STAR-CCM+

    Energy Technology Data Exchange (ETDEWEB)

    Morgado Canada, E.; Concejal Barmejo, A.; Jimenez Varas, G.; Solar Martinez, A.

    2014-07-01

    The objective of the project is the evaluation of the code STAR-CCM +, as well as the establishment of guidelines and standardized procedures for the discretization of the area of study and the selection of physical models suitable for the simulation of BWR fuel. For this purpose several of BFBT experiments have simulated [1] provide a data base for the development of experiments for measuring distribution of fractions of holes to changes in power in order to find the most appropriate models for the simulation of the problem. (Author)

  11. 19 CFR 159.38 - Rates for estimated duties.

    Science.gov (United States)

    2010-04-01

    ... duties. For purposes of calculating estimated duties, the port director shall use the rate or rates... 19 Customs Duties 2 2010-04-01 2010-04-01 false Rates for estimated duties. 159.38 Section 159.38 Customs Duties U.S. CUSTOMS AND BORDER PROTECTION, DEPARTMENT OF HOMELAND SECURITY; DEPARTMENT OF...

  12. Comparison study of the thermal mechanical performance of fuel rods during BWR fuel preconditioning operations using the computer codes FUELSIM and FEMAXI-V

    Energy Technology Data Exchange (ETDEWEB)

    Pantoja C, R. [IPN, Escuela Superior de Fisica y Matematicas, Departamento de Ingenieria Nuclear, Av. Instituto Politecnico Nacional s/n, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Ortiz V, J.; Castillo D, R., E-mail: rafael.pantoja10@yahoo.com.m [ININ, Departamento de Sistemas Nucleares, Carretera Mexico-Toluca s/n, Ocoyoacac 52750, Estado de Mexico (Mexico)

    2010-10-15

    The safety of nuclear power plants requires monitoring those parameters having some direct or indirect effect on safety. The thermal limits are values set for those parameters considered having most impact on the safe operation of a nuclear power reactor. Some thermal limits monitoring requires the thermal-mechanical analysis of the rods containing the nuclear fuel. The fuel rod thermal-mechanical behaviour under irradiation is a complex process in which there exists a great deal of interrelated physical and chemical phenomena, so that the fuel rod performance analysis in the core of a nuclear power reactor is generally accomplished by using computer codes, which integrate several of the phenomena that are expected to occur during the lifetime of the fuel rod in the core. In the operation of a nuclear power reactor, pre-conditioning simulations are necessary to determine in advance limit values for the power that can be generated in a fuel rod during any power ramp, and mainly during reactor startup, and thus avoiding any rod damage. In this work, a first analysis of the thermal-mechanical performance of typical fuel rods used in nuclear reactors of the type BWR is performed. This study includes two types of fuel rods: one from a fuel assembly design with array 8 x 8, and the other one from a 10 x 10 fuel assembly design, and a comparison of the thermal-mechanical performance between the two different rod designs is performed. The performance simulations were performed by the code FUELSIM, and compared against results previously obtained from similar simulation with the code FEMAXI-V. (Author)

  13. Modeling and validation of a mechanistic tool (MEFISTO) for the prediction of critical power in BWR fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Adamsson, Carl, E-mail: carl.adamsson@psi.ch [Westinghouse Electric Sweden, SE-721 63, Vaesteras (Sweden); Le Corre, Jean-Marie, E-mail: lecorrjm@westinghouse.com [Westinghouse Electric Sweden, SE-721 63, Vaesteras (Sweden)

    2011-08-15

    Highlights: > The MEFISTO code efficiently and accurately predicts the dryout event in a BWR fuel bundle, using a mechanistic model. > A hybrid approach between a fast and robust sub-channel analysis and a three-field two-phase analysis is adopted. > MEFISTO modeling approach, calibration, CPU usage, sensitivity, trend analysis and performance evaluation are presented. > The calibration parameters and process were carefully selected to preserve the mechanistic nature of the code. > The code dryout prediction performance is near the level of fuel-specific empirical dryout correlations. - Abstract: Westinghouse is currently developing the MEFISTO code with the main goal to achieve fast, robust, practical and reliable prediction of steady-state dryout Critical Power in Boiling Water Reactor (BWR) fuel bundle based on a mechanistic approach. A computationally efficient simulation scheme was used to achieve this goal, where the code resolves all relevant field (drop, steam and multi-film) mass balance equations, within the annular flow region, at the sub-channel level while relying on a fast and robust two-phase (liquid/steam) sub-channel solution to provide the cross-flow information. The MEFISTO code can hence provide highly detailed solution of the multi-film flow in BWR fuel bundle while enhancing flexibility and reducing the computer time by an order of magnitude as compared to a standard three-field sub-channel analysis approach. Models for the numerical computation of the one-dimensional field flowrate distributions in an open channel (e.g. a sub-channel), including the numerical treatment of field cross-flows, part-length rods, spacers grids and post-dryout conditions are presented in this paper. The MEFISTO code is then applied to dryout prediction in BWR fuel bundle using VIPRE-W as a fast and robust two-phase sub-channel driver code. The dryout power is numerically predicted by iterating on the bundle power so that the minimum film flowrate in the bundle

  14. Forgiveness and the Limits of Duty

    NARCIS (Netherlands)

    Archer, Alfred

    2017-01-01

    Can there be a duty to forgive those who have wronged us? According to a popular view amongst philosophers working on forgiveness the answer is no. Forgiveness, it is claimed, is always elective. This view is rejected by Gamlund (2010a; 2010b) who argues that duties to forgive do exist and then

  15. Health versus harm: euthanasia and physicians' duties.

    Science.gov (United States)

    Garcia, J L A

    2007-01-01

    This essay rebuts Gary Seay's efforts to show that committing euthanasia need not conflict with a physician's professional duties. First, I try to show how his misunderstanding of the correlativity of rights and duties and his discussion of the foundation of moral rights undermine his case. Second, I show aspects of physicians' professional duties that clash with euthanasia, and that attempts to avoid this clash lead to absurdities. For professional duties are best understood as deriving from professional virtues and the commitments and purposes with which the professional as such ought to act, and there is no plausible way in which her death can be seen as advancing the patient's medical welfare. Third, I argue against Prof. Seay's assumption that apparent conflicts among professional duties must be resolved through "balancing" and argue that, while the physician's duty to extend life is continuous with her duty to protect health, any duty to relieve pain is subordinate to these. Finally, I show that what is morally determinative here, as throughout the moral life, is the agent's intention and that Prof. Seay's implicitly preferred consequentialism threatens not only to distort moral thinking but would altogether undermine the medical (and any other) profession and its internal ethics.

  16. 45 CFR 46.403 - IRB duties.

    Science.gov (United States)

    2010-10-01

    ... 45 Public Welfare 1 2010-10-01 2010-10-01 false IRB duties. 46.403 Section 46.403 Public Welfare DEPARTMENT OF HEALTH AND HUMAN SERVICES GENERAL ADMINISTRATION PROTECTION OF HUMAN SUBJECTS Additional Protections for Children Involved as Subjects in Research § 46.403 IRB duties. In addition to...

  17. Exit and the duty to admit

    National Research Council Canada - National Science Library

    Lenard, Patti Tamara

    2015-01-01

    .... I argue that receiving states are duty-bound to act in ways that enable migrants to exercise their right to exit. In particular, I argue that receiving states have a perfect duty to collectivize the process by which needy migrants can exercise the right to exit.

  18. Forgiveness and the Limits of Duty

    NARCIS (Netherlands)

    Archer, Alfred

    2017-01-01

    Can there be a duty to forgive those who have wronged us? According to a popular view amongst philosophers working on forgiveness the answer is no. Forgiveness, it is claimed, is always elective. This view is rejected by Gamlund (2010a; 2010b) who argues that duties to forgive do exist and then prov

  19. Seven Legal Duties of a Coach.

    Science.gov (United States)

    Figone, Albert J.

    1989-01-01

    This article identifies seven legal duties of coaches and discusses the practical application of each. These duties relate to supervision, planning, warning of risks, providing a safe environment, evaluating players for injuries and incapacities, fairly matching players, and first aid and emergency procedures. (IAH)

  20. 7 CFR 1205.332 - Duties.

    Science.gov (United States)

    2010-01-01

    ... the Secretary, a program of research, advertising, and sales promotion projects, together with a... AND ORDERS; MISCELLANEOUS COMMODITIES), DEPARTMENT OF AGRICULTURE COTTON RESEARCH AND PROMOTION Cotton Research and Promotion Order Cotton Board § 1205.332 Duties. The Board shall have the following duties:...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

  5. Duty ratio of cooperative molecular motors.

    Science.gov (United States)

    Dharan, Nadiv; Farago, Oded

    2012-02-01

    Molecular motors are found throughout the cells of the human body and have many different and important roles. These micromachines move along filament tracks and have the ability to convert chemical energy into mechanical work that powers cellular motility. Different types of motors are characterized by different duty ratios, which is the fraction of time that a motor is attached to its filament. In the case of myosin II (a nonprocessive molecular machine with a low duty ratio), cooperativity between several motors is essential to induce motion along its actin filament track. In this work we use statistical mechanical tools to calculate the duty ratio of cooperative molecular motors. The model suggests that the effective duty ratio of nonprocessive motors that work in cooperation is lower than the duty ratio of the individual motors. The origin of this effect is the elastic tension that develops in the filament which is relieved when motors detach from the track.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-03-01

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

  8. 75 FR 5964 - Certain Polyester Staple Fiber From Taiwan: Preliminary Results of Antidumping Duty...

    Science.gov (United States)

    2010-02-05

    ... instruct U.S. Customs and Border Protection (CBP) to assess antidumping duties on all appropriate entries... to the order may be coated, usually with a silicon or other finish, or not coated. PSF is generally..., for the following movement expenses: inland freight from the plant to the port of exportation,...

  9. [Patients' rights--doctors' duties].

    Science.gov (United States)

    Jaeger, L; Bertram, E; Grate, S; Mischkowsky, T; Paul, D; Probst, J; Scala, E; Wbllenweber, H D

    2015-06-01

    On 26 February 2013 the new "Law on Patients' Rights" (hereinafter also the "Law") became effective. This Law strengthens patients' rights vis-à-vis the insurdnce company and also regulates patients' rights regarding their relation to the doctor. This has consequences for the laws on medical liability all doctors must consider. The doctor's performance is and remains a service and such service does not hold any guarantee of success. Nevertheless, this Law primarily reads as a "law on the duties of physicians". To duly take into account these duties and to avoid mistakes and misinterpretation of the Law, the Ethics Committee of the Consortium of Osteosynthesis Trauma Germany (AOTRAUMA-D) has drafted comments on the Law. Brief summaries of its effects are to be found at the end of the respective comment under the heading "Consequences for Practice". The text of the law was influenced particularly by case law, as continuously developed by the German Federal Court of Justice ("BGH"). The implementation of the Law on Patients' Rights was effected by the newly inserted sections 630a to 630h of the German Civil Code (the "BGB"), which are analysed below. The following comments are addressed to physicians only and do not deal with the specific requirements and particularities of the other medical professions such as physiotherapy, midwifery and others so on. Special attention should be paid to the comments on the newly inserted Duty to inform, which has to be fullfilled prior to any diagnostic or therapeutic procedure (sec. 630c para 2 sentence 1 BGB). Under certain conditions the doctor also has to inform the patient about the circumstances that lead to the presumed occurance of a therapeutic or diagnostic malpractice (sec. 630c para. 2 sentence 2 BGB), based on the manifestation of an undesired event or an undesired outcome. As before, the patient's valid consent to any procedure (sec. 630d BGB) is directly linked to the comprehensive and timely provision of information

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

    Energy Technology Data Exchange (ETDEWEB)

    Nilsson, Lars [Lentek, Nykoeping (Sweden)

    2006-05-15

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

  11. Heavy duty complete extension slides

    Science.gov (United States)

    Bueno, José Ignacio; Vázquez, Javier

    2001-09-01

    The selection from available commercial market of a set of slides to be used in an habitable pressurised module in space, to draw a 660 mm box out of a rack, up to a completely extracted position in a safely supported configuration, seems in principle not to be a complicated task. That was the first approach taken in the design process of the telescopic guides of the Crew Work Bench (CWB) included in the Fluid Science Laboratory (FSL), part of "ESA Microgravity Facilities for Columbus" within the Columbus Orbital Facility (COF) of the International Space Station (ISS). Nevertheless, common space compatible requirements such as materials, specific environmental loads, available envelope, total weight, etc., can make the selection of telescopic slides from commercial market unfeasible. A specific development to design space compatible telescopic slides for the CWB was undertaken. A set of heavy duty space compatible telescopic slides were designed, manufactured and tested. They should be operative in both, 1-g environment and in orbit, and additionally should withstand an inadvertent astronaut kick or bump of 556 N in any direction.

  12. 7 CFR 1210.328 - Duties.

    Science.gov (United States)

    2010-01-01

    ... Regulations of the Department of Agriculture (Continued) AGRICULTURAL MARKETING SERVICE (MARKETING AGREEMENTS AND ORDERS; MISCELLANEOUS COMMODITIES), DEPARTMENT OF AGRICULTURE WATERMELON RESEARCH AND PROMOTION PLAN Watermelon Research and Promotion Plan National Watermelon Promotion Board § 1210.328 Duties....

  13. 7 CFR 1207.328 - Duties.

    Science.gov (United States)

    2010-01-01

    ... Regulations of the Department of Agriculture (Continued) AGRICULTURAL MARKETING SERVICE (MARKETING AGREEMENTS AND ORDERS; MISCELLANEOUS COMMODITIES), DEPARTMENT OF AGRICULTURE POTATO RESEARCH AND PROMOTION PLAN Potato Research and Promotion Plan National Potato Promotion Board § 1207.328 Duties. The Board...

  14. Ethics, pandemics, and the duty to treat.

    Science.gov (United States)

    Malm, Heidi; May, Thomas; Francis, Leslie P; Omer, Saad B; Salmon, Daniel A; Hood, Robert

    2008-08-01

    Numerous grounds have been offered for the view that healthcare workers have a duty to treat, including expressed consent, implied consent, special training, reciprocity (also called the social contract view), and professional oaths and codes. Quite often, however, these grounds are simply asserted without being adequately defended or without the defenses being critically evaluated. This essay aims to help remedy that problem by providing a critical examination of the strengths and weaknesses of each of these five grounds for asserting that healthcare workers have a duty to treat, especially as that duty would arise in the context of an infectious disease pandemic. Ultimately, it argues that none of the defenses is currently sufficient to ground the kind of duty that would be needed in a pandemic. It concludes by sketching some practical recommendations in that regard.

  15. State duties of protection and fundamental rights

    Directory of Open Access Journals (Sweden)

    C Starck

    2000-05-01

    Full Text Available Duties of protection are duties of the state to protect certain legal interests of its citizens. They cover the interests of life, health, freedom and property and also protect some other interests and certain constitutionally recognised institutions. State duties of protection must be considered in connection with fundamental rights. The foundations of modern constitutionalism and attendant procedures are essential to develop guidelines for a constructive critique of the jurisprudence of the Constitutional Court. This is done with reference to the recent history of France, Germany and England. The historical excursus reveals that a single theory underlies the variety of constitutional states. The development of the constitutional state gave rise to the significance of the preservation of freedom through the maintenance of law and the separation of powers. This has given rise to various legal devices, based also in part on experience with moderate rule and earlier theories of the imperium limitatum.A textual analysis of the German Basic Law is undertaken to determine whether and how the duties of protection are expressly created. Furthermore, the duties that have been discovered in the Basic Law by the Federal Constitutional Court are considered. These duties include the protection of human life and health, personal freedom, the right to autonomous development of one's personality, freedom of science, research and teaching, marriage and the family, children, mothers, professional freedom, property and the protection of German nationals against foreign states. Finally the justification of such duties and the constitutional control of the manner of protection are considered.In a final section a critique of relevant constitutional jurisprudence is undertaken. It is argued that claims to protection cannot be directly binding law. They presuppose legislation. If statutory protection is connected with infringements of third-party fundamental rights

  16. 19 CFR 10.43 - Duty-free status.

    Science.gov (United States)

    2010-04-01

    ... § 10.43 Duty-free status. (a) The port director may, at his discretion, require appropriate proof of... 19 Customs Duties 1 2010-04-01 2010-04-01 false Duty-free status. 10.43 Section 10.43 Customs Duties U.S. CUSTOMS AND BORDER PROTECTION, DEPARTMENT OF HOMELAND SECURITY; DEPARTMENT OF THE...

  17. 19 CFR 151.22 - Estimated duties on raw sugar.

    Science.gov (United States)

    2010-04-01

    ... 19 Customs Duties 2 2010-04-01 2010-04-01 false Estimated duties on raw sugar. 151.22 Section 151... THE TREASURY (CONTINUED) EXAMINATION, SAMPLING, AND TESTING OF MERCHANDISE Sugars, Sirups, and Molasses § 151.22 Estimated duties on raw sugar. Estimated duties shall be taken on raw sugar, as defined...

  18. 19 CFR 12.62 - Enforcement; duties of Customs officers.

    Science.gov (United States)

    2010-04-01

    ... 19 Customs Duties 1 2010-04-01 2010-04-01 false Enforcement; duties of Customs officers. 12.62 Section 12.62 Customs Duties U.S. CUSTOMS AND BORDER PROTECTION, DEPARTMENT OF HOMELAND SECURITY...; duties of Customs officers. (a) In accordance with the authority contained in sections 10 and 12 of...

  19. Simulation of a scenario of total loss of external and internal power (Sbo) for different vent pressures of the containment of a BWR-5; Simulacion de un escenario de perdida total de potencia externa e interna (SBO) para distintas presiones de venteo de la contencion de un reactor BWR-5

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas V, J.; Mugica R, C. A.; Godinez S, V., E-mail: Jaime.cardenas@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

    2014-10-15

    The simulation of a Station Black Out (Sbo) was realized with intervention of the vent containment by means of a rigid vent coming from the dry-well and that discharges directly to the atmosphere, with the MELCOR code version 2.1. This scenario was carried out for a BWR-5 and containment type Mark II, with a thermal power of 2317 MWt similar to the reactor of nuclear power plant of Laguna Verde. For this scenario was considered as only available system for coolant injection to the reactor to the Reactor Core Isolation Cooling (Rcic), which remained operating 4 hours with batteries bank. The Security and Relief Valves (SR V) were considered functional (by simplicity) and that they mechanically do not exceed their capacity to liberate pressure due to the performances in their safety way. The operator maneuver to perform the SR V and to de pressurize the vessel until the pressure (13 kg/cm{sup 2}) to operate the low pressure systems was modeled. The results cover approximately 48 hours (172000 seconds), time in which was observed the behavior of the level and pressure in the vessel. Also the scenario evolution was analyzed to different vent pressures of the primary containment (2.0, 3.0, 4.5, 6.0, and 10.0 kg/cm{sup 2}), the temperature profiles of the dry-well, the hydrogen accumulation in the containment, the radio-nuclides liberation through rigid vent to the atmosphere and the inventory of these. In this work an analysis of the pressure behavior in the primary containment is presented, with the purpose of minimizing liberated fission products to the environment. (Author)

  20. 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

  1. PSI contribution to the CASTOC round robin on EAC of low-alloy RPV steels under BWR conditions

    Energy Technology Data Exchange (ETDEWEB)

    Seifert, H.P.; Ritter, S

    2001-08-01

    Within the CASTOC-project (5th EU FW programme), the environmentally-assisted crack growth (EAC) behaviour of low-alloy reactor pressure vessel (RPV) steels is experimentally investigated under simulated transient and steady-state BWR power operation conditions by 6 European laboratories. The present report contains a summary of the PSI contribution to the Working Package 1 (WP1) of this project. WP1 is an interlaboratory round robin EAC test in simulated BWR/NWC environment under cyclic and static loading conditions. The round robin shall demonstrate the applicability of the used advanced test technique and establishes the technical basis for the decision of test conditions in the other working packages. In the first part of the report, the PSI testing facility/measurement instruments and the applied test and evaluation procedure are discussed in detail. In the second part, the exact test conditions and test results with detailed post-test fractographical evaluation in the SEM are presented. The test results are compared with other PSI results, literature data and nuclear codes. Stable and stationary test conditions within the specified range could be achieved in the PSI test during the whole conditioning and experimental phase. The cyclic crack growth rate results agree well with recent PSI results at a higher dissolved oxygen content of 8 ppm and are slightly below the 'high-sulphur line' of the PLEDGE-model. The crack growth rates are significantly above the ASME XI 'wet' curve. Compared to fatigue crack growth rates in air under otherwise identical test conditions, the effect of the high-temperature water environment resulted in an acceleration of crack growth by a factor of 150-250 under these low-cyclic loading conditions. The test results at constant load confirm the extremely low susceptibility to SCC crack growth under static load at 288 {sup o}C observed in tests at MPA, PSI and in a European Round Robin. They agree well with the RPV

  2. Towards a Duty of Care for Biodiversity

    Science.gov (United States)

    Earl, G.; Curtis, A.; Allan, C.

    2010-04-01

    The decline in biodiversity is a worldwide phenomenon, with current rates of species extinction more dramatic than any previously recorded. Habitat loss has been identified as the major cause of biodiversity decline. In this article we suggest that a statutory duty of care would complement the current mix of policy options for biodiversity conservation. Obstacles hindering the introduction of a statutory duty of care include linguistic ambiguity about the terms ‘duty of care’ and ‘stewardship’ and how they are applied in a natural resource management context, and the absence of a mechanism to guide its implementation. Drawing on international literature and key informant interviews we have articulated characteristics of duty of care to reduce linguistic ambiguity, and developed a framework for implementing a duty of care for biodiversity at the regional scale. The framework draws on key elements of the common law ‘duty of care’, the concepts of ‘taking reasonable care’ and ‘avoiding foreseeable harm’, in its logic. Core elements of the framework include desired outcomes for biodiversity, supported by current recommended practices. The focus on outcomes provides opportunities for the development of innovative management practices. The framework incorporates multiple pathways for the redress of non-compliance including tiered negative sanctions, and positive measures to encourage compliance. Importantly, the framework addresses the need for change and adaptation that is a necessary part of biodiversity management.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-07-01

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

  9. Trace Code Validation for BWR Spray Cooling Injection and CCFL Condition Based on GÖTA Facility Experiments

    Directory of Open Access Journals (Sweden)

    Stefano Racca

    2012-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-01-01

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

  11. Description of a disposition line on the stress corrosion cracking behaviour of ferritic reactor pressure vessel steels under BWR-conditions; Beschreibung einer einhuellenden Risswachstumskurve zum Spannungsrisskorrosionsverhalten von ferritischen Reaktordruckbehaelter (RDB)-Staehlen unter Siedewasserreaktor (SWR)-Bedingungen

    Energy Technology Data Exchange (ETDEWEB)

    Bruemmer, G. [HEW, Hamburg (Germany); Hoffmann, H. [VGB-GS, Essen (Germany); Ilg, U. [EnBW Kraftwerke AG, Philippsburg (Germany); Wachter, O. [E.ON-Kernkraft, Hannover (Germany); Widera, M. [RWE Power, Essen (Germany); Roth, A. [Framatome ANP GmbH, Erlangen (Germany)

    2002-07-01

    The inner surface of the reactor pressure vessel of BWR reactors is lined with a welded, corrosion-resistant steel liner. In an assumed case of liner rupture down to the low-alloy ferritic base material, an integrity assessment of the pressure vesssel in consideration of the effects of reactor coolant is of utmost importance, and research in this field has been going on for more than ten years now. An analysis of the available data shows that it is now possible to describe a disposition line on the stress corrosion cracking behaviour of ferritic reactor pressure vessel steels in BWR conditions. Crack growth rates of a stress intensity factor corresponding to a T/4 wall defect (i.e. 25 percent of the wall thickness) are technically not relevant. This scientific finding is supported by measurements of about 450 reactor operation years of all German LWR reactor plants, none of which showed crack initiation in the reactor pressure vessel. [German] Die mediumberuehrte Innenoberflaeche des Reaktordruckbehaelters (RDB) von Siedewasserreaktoren (SWR) ist mit einer korrosionsbestaendigen austenitischen Schweissplattierung versehen. Fuer den unterstellten Fall einer bis auf den niedriglegierten, ferritischen Grundwerkstoff durchgerissenen Pattierung ist fuer die Beurteilung der Integritaet des RDB unter Beruecksichtigung der Einwirkung des Reaktorkuehlmittels die Klaerung der Frage eines korrosionsgestuetzten Risswachstums von grosser Bedeutung. Dieses Thema ist daher bereits seit mehr als 10 Jahren Gegenstand umfangreicher Forschungsaktivitaeten. Ende der 80er- und Anfang der 90er-Jahre wurden fuer ferritische RDB-Staehle von SWR-Anlagen Risswachstumsgeschwindigkeiten veroeffentlicht, die binnen weniger als einem Jahr zum Durchriss der drucktragenden Wand eines RDB gefuehrt haetten. Daraufhin wurden internationale Forschungsaktivitaeten zur Ermittlung zuverlaessiger und reproduzierbarer Risswachstumsdaten initiiert, deren Ergebnisse zusammenfassend dargestellt werden. Die

  12. Analysis of the microstructural evolution of the damage by neutron irradiation in the pressure vessel of a nuclear power reactor BWR; Analisis de la evolucion microestructural del dano por irradiacion neutronica en la vasija de presion de un reactor nuclear de potencia BWR

    Energy Technology Data Exchange (ETDEWEB)

    Moranchel y R, M.

    2012-07-01

    Nuclear reactor pressure vessel type BWR, installed in Mexico and in many other countries, are made of an alloy of low carbon steel. The American Society for Testing and Materials (Astm) classifies this alloy as A533-B, class 1. Both the vessel and other internal structures are continuously exposed to the neutron flux from the reactions of fission in nuclear fuel. A large number of neutrons reach the vessel and penetrate certain depth depending on their energy. Its penetration in the neutron collides with the nuclei of the atoms out of their positions in the crystal lattice of steel, producing vacancies, interstitial, segregations, among other defects, capable of affecting its mechanical properties. Analyze the micro-structural damage to the vessel due to neutron irradiation, is essential for reasons of integrity of this enclosure and safety of any nuclear power plant. The objective of this thesis work is theoretical and experimentally determine the microstructural damage of a type nuclear reactor vessel steel BWR, due to neutron radiation from the reactor core, using microscopic and spectroscopic techniques as well as Monte Carlo simulation. Microscopy Optical, Scanning Electron Microscopy, Transmission Electron Microscopy, Energy Dispersion of X-rays Spectrometry and X-rays Diffractometry were the techniques used in this research. These techniques helped in the characterization of both the basis of design of pressure vessel steel and steel irradiated, after eight years of neutron irradiation on the vessel, allowing know the surface morphology and crystal structures of the previous steel and post-irradiation, analyze the change in the microstructure of the steel vessel, morphological damage to surface level in an irradiated sample, among which are cavities in the order of microns produced by Atomic displacements due to the impact of neutronic, above all in the first layers of thickness of the vessel, the effect of swelling, regions of greater damage and Atomic

  13. Analysis of BWR instabilities coupled with 3D code RELAP5 / PARCSv2.7. Application to the event happened in Oskarshamn-2 in 1999; Analisis de inestabilidades en BWR con el codigo acoplado 3D RELAP5/PARCSv2.7. Aplicacion al evento sucedido en Oskarshamn-2 en1999

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-07-01

    In this work, part of our works in the frame of the OECD/NEA Oskarshamn-2 (O{sub 2}) BWR Stability Benchmark for Coupled Code Calculations and Uncertainty Analysis in Modelling are shown. The objective is to simulate the instability event registered in February 1999 at the Swedish NPP Oskarshamn-2 with the coupled code RELAP5/PARCSv2.7. (Author)

  14. Study and characterization of noble metal deposits on similar rusty surfaces to those of the reactor U-1 type BWR of nuclear power station of Laguna Verde; Estudio y caracterizacion de depositos de metales nobles sobre superficies oxidadas similares a las del reactor de la Central de Laguna Verde (CNLV) U1 del tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Flores S, V. H.

    2011-07-01

    In the present investigation work, were determined the parameters to simulate the conditions of internal oxidation reactor circulation pipes of the nuclear power plant of Laguna Verde in Veracruz. We used 304l stainless steel cylinders with two faces prepared with abrasive paper of No. 600, with the finality to obtain similar surface to the internal circulation piping nuclear reactor. Oxides was formed within an autoclave (Autoclave MEX-02 unit B), which is a device that simulates the working conditions of the nuclear reactor, but without radiation generated by the fission reaction within the reactor. The oxidation conditions were a temperature of 280 C and pressure of 8 MPa, similar conditions to the reactor operating in nuclear power plant of Laguna Verde in Veracruz, Mexico (BWR conditions), with an average conductivity of 4.58 ms / cm and 2352 ppb oxygen to simulate normal water chemistry NWC. Were obtained deposits of noble metal oxides formed on 304l stainless steel samples, in a 250 ml autoclave at a temperature range of 180 to 200 C. The elements that were used to deposit platinum-rhodium (Pt-Rh) with aqueous Na{sub 2}Pt (OH){sub 6} and Na{sub 3}Rh (NO{sub 2}){sub 6}, Silver (Ag) with an aqueous solution of AgNO{sub 3}, zirconium (Zr) with aqueous Zr O (NO{sub 3}) and ZrO{sub 2}, and zinc (Zn) in aqueous solution of Zn (NO{sub 3}){sub 2} under conditions of normal water chemistry. Also there was the oxidation of 304l stainless steel specimens in normal water chemistry with a solution of Zinc (Zn) (NWC + Zn). Oxidation of the specimens in water chemistry with a solution of zinc (Zn + NWC) was prepared in two ways: within the MEX-02 autoclave unit A in a solution of zinc and a flask at constant temperature in zinc solution. The oxides formed and deposits were characterized by scanning electron microscopy, energy dispersive X-ray analysis, elemental field analysis and X-ray diffraction. By other hand was evaluated the electrochemical behavior of the oxides

  15. Duty-ratio of cooperative molecular motors

    CERN Document Server

    Dharan, Nadiv

    2012-01-01

    Molecular motors are found throughout the cells of the human body, and have many different and important roles. These micro-machines move along filament tracks, and have the ability to convert chemical energy into mechanical work that powers cellular motility. Different types of motors are characterized by different duty-ratios, which is the fraction of time that a motor is attached to its filament. In the case of myosin II - a non-processive molecular machine with a low duty ratio - cooperativity between several motors is essential to induce motion along its actin filament track. In this work we use statistical mechanical tools to calculate the duty ratio of cooperative molecular motors. The model suggests that the effective duty ratio of non-processive motors that work in cooperation is lower than the duty ratio of the individual motors. The origin of this effect is the elastic tension that develops in the filament which is relieved when motors detach from the track.

  16. Development of Sirius facility that simulates void-reactivity feedback, and regional and core-wide stability estimation of natural circulation BWR

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-01

    The SIRIUS facility was designed and constructed for highly accurate simulation of core-wide and regional instabilities of the BWR. A real-time simulation was performed in the digital controller for modal point kinetics of reactor neutronics and fuel-rod conduction on the basis of measured void fractions in reactor core sections of the thermal-hydraulic loop. Stability experiments were conducted for a wide range of fluid conditions, power distributions, and fuel rod thermal conductivity time constants, including the normal operating conditions of a typical natural circulation BWR. The results showed that there is a sufficiently wide stability margin under normal operating conditions, even when void-reactivity feedback is taken into account. (author)

  17. Effect of sensitization and cold work on stress corrosion susceptibility of austenitic stainless steels in boiling water reactor (BWR) and pressurized water reactor (PWR) conditions

    Energy Technology Data Exchange (ETDEWEB)

    Haenninen, H.; Aho-Mantila, I.

    1981-05-01

    The influence of metallurgical variables on stress corrosion cracking of austenitic stainless steels, in particular AISI 304 and OX18H10T, was examined in O/sub 2/ enriched BWR conditions (8 ppm O/sub 2/) and in typical PWR conditions. Cracking susceptibility in BWR conditions is especially sensitive to alpha martensite content and sensitization. Cracking in alpha martensite compounds is intergranular and transgranular and it can not be related to sensitization. Sensitization induces cracking only in creviced conditions (double U-bend specimens) in AISI 304 steels. In creviced conditions OX18H10T steel exhibits cracking in solution annealed, stabilized and sensitized conditions. The sensitized material is most susceptible. Cracking in solution annealed and stabilized OX18H10T steel is intergranular and transgranular. In PWR conditions (O/sub 2/ content 2 ppb) no cracking is observed. (ESA)

  18. Robots cut risks and costs in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Roman, H.T. (PSE and G (US))

    1991-07-01

    The electric utility industry, the third largest user of computers in the United States, is realizing the value of a relatively new computer-related technology; robotics. Since its use in the cleanup of Three Mile Island, 44 utility companies have used robotic devices to save radiation exposure to their employees and to achieve measurable and often significant cost savings. This paper reports on the use of robots to reduce risk and cost in electric utilities especially in PWR and BWR nuclear power plants.

  19. Experience of Areva in fuel services for PWR and BWR; Experiencia de Areva en servicios de combustible para PWR y BWR

    Energy Technology Data Exchange (ETDEWEB)

    Morales, I.

    2015-07-01

    AREVA being an integrated supplier of fuel assemblies has included in its strategy to develop services and solutions to customers who desire to improve the performance and safety of their fuel. These services go beyond the simple 'after sale' services that can be expected from a fuel supplier: The portfolio of AREVA includes a wide variety of services, from scientific calculations to fuel handling services in a nuclear power plant. AREVA is committed to collaborate and to propose best-in-class solutions that really make the difference for the customer, based on 40 years of Fuel design and manufacturing experience. (Author)

  20. Crack growth behaviour of low-alloy steels for pressure boundary components under transient light water reactor operating conditions - CASTOC, Part I: BWR/NWC conditions

    Energy Technology Data Exchange (ETDEWEB)

    Ritter, S.; Seifert, H.P. [Paul Scherrer Institute, PSI, Villigen (Switzerland); Devrient, B.; Roth, A. [Framatome ANP GmbH, Erlangen (Germany); Ehrnsten, U. [VTT Industrial Systems, Espoo (Finland); Ernestova, M.; Zamboch, M. [Nuclear Research Institute, NRI, Rez (Czech Republic); Foehl, J.; Weissenberg, T. [Staatliche Materialpruefungsanstalt, MPA, Stuttgart (Germany); Gomez-Briceno, D.; Lapena, J. [Centro de Investigaciones Energeticas Medioambientales y Tecnologicas, CIEMAT, Madrid (Spain)

    2004-07-01

    One of the ageing phenomena of pressure boundary components of light water reactors (LWR) is environmentally-assisted cracking (EAC). The project CASTOC (5. Framework Programme of the EU) was launched September 2000 with six European partners and terminated August 2003. It was focused in particular on the EAC behaviour of low-alloy steels (LAS) and to some extent to weld metal, heat affected zone and the influence of an austenitic cladding. The main objective was directed to the clarification of EAC crack growth behaviour/mechanism of LAS in high-temperature water under steady-state power operation (constant load) and transient operating conditions (e.g., start-up/shut-down, transients in water chemistry and load). Autoclave tests were performed with Western and Russian type reactor pressure vessel steels under simulated boiling water reactor (BWR)/normal water chemistry (NWC) and pressurised water reactor (VVER) conditions. The investigations were performed with fracture mechanics specimens of different sizes and geometries. The applied loading comprised cyclic loads, static loads and load spectra where the static load was periodically interrupted by partial unloading. With regard to water chemistry, the oxygen content (VVER) and impurities of sulphate and chlorides (BWR) were varied beyond allowable limits for continuous operation. The current paper summarises the most important crack growth results obtained under simulated BWR/NWC conditions. The results are discussed in the context of the current crack growth rate curves in the corresponding nuclear codes. (authors)

  1. Development and Assessment of CFD Models Including a Supplemental Program Code for Analyzing Buoyancy-Driven Flows Through BWR Fuel Assemblies in SFP Complete LOCA Scenarios

    Science.gov (United States)

    Artnak, Edward Joseph, III

    This work seeks to illustrate the potential benefits afforded by implementing aspects of fluid dynamics, especially the latest computational fluid dynamics (CFD) modeling approach, through numerical experimentation and the traditional discipline of physical experimentation to improve the calibration of the severe reactor accident analysis code, MELCOR, in one of several spent fuel pool (SFP) complete loss-ofcoolant accident (LOCA) scenarios. While the scope of experimental work performed by Sandia National Laboratories (SNL) extends well beyond that which is reasonably addressed by our allotted resources and computational time in accordance with initial project allocations to complete the report, these simulated case trials produced a significant array of supplementary high-fidelity solutions and hydraulic flow-field data in support of SNL research objectives. Results contained herein show FLUENT CFD model representations of a 9x9 BWR fuel assembly in conditions corresponding to a complete loss-of-coolant accident scenario. In addition to the CFD model developments, a MATLAB based controlvolume model was constructed to independently assess the 9x9 BWR fuel assembly under similar accident scenarios. The data produced from this work show that FLUENT CFD models are capable of resolving complex flow fields within a BWR fuel assembly in the realm of buoyancy-induced mass flow rates and that characteristic hydraulic parameters from such CFD simulations (or physical experiments) are reasonably employed in corresponding constitutive correlations for developing simplified numerical models of comparable solution accuracy.

  2. CFD analysis for the hydrogen transport in the primary contention of a BWR using the codes OpenFOAM and Gas-Flow; Analisis CFD para el transporte de hidrogeno en la contencion primaria de un reactor BWR usando los codigos OpenFOAM y GasFlow

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez P, D. A.

    2014-07-01

    The accidents in Unit 2 of the Three Mile Island Nuclear Power Plant (NPP) in the United States (March 28{sup th}, 1979), the one in Unit 4 of the NPP Chernobyl in Ukraine (April 26{sup th}, 1986) and the explosions in some units of Fukushima NPP in Japan (March 11{sup th}, 2011) boosted the investigations on severe accidents with core damage and, in particular, the threat to the ultimate barrier by an eventual explosion from uncontrolled Hydrogen combustion within the containment was considered of particular relevance. Research programs for analyzing Hydrogen behavior and control during this kind of accidents were early initiated by research and regulatory bodies. Assessment on Hydrogen behavior once it has been postulated to be released on the containment system can be divided into two phases, in the first one, transport and the concentrations of the gas mixtures and steam in each volume or area comprised between the structures of the containment are calculated, in the second one, the propagation of the detonation of the Hydrogen is calculated if there are the conditions to occur. Currently, there are computer programs that can be used in one, or both stages of computation, and they are based on one of the two solution methods in current use, one of them are integrated codes (e.g. MELCOR), which consists in assuming the containment as a network composed of hydraulic tanks or nodes on which the balance equations of mass and energy have to be solved, the network is connected by ducts or connections where the momentum balance equation arise. This methodology relies on the use of semi-empirical relationships and the criteria used to define a geometric pattern, are subjective. The second method, which is having relevance due to the large computing power of modern computers, is the numerical solution of the three-dimensional Navier-Stokes equations in complex geometries. This method of solution is known as Computational Fluid Dynamics (CFD), and offers the advantage of

  3. Simulation of the operational monitoring of a BWR with Simulate-3; Simulacion del seguimiento operacional de un reactor BWR con Simulate-3

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez F, J. O.; Martin del Campo M, C.; Fuentes M, L.; Francois L, J. L., E-mail: ace.jo.cu@gmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico)

    2015-09-15

    This work was developed in order to describe the methodology for calculating the fuel burned of nuclear power reactors throughout the duration of their operating cycle and for each fuel reload. In other words, simulate and give monitoring to the main operation parameters of sequential way along its operation cycles. For this particular case, the operational monitoring of five consecutive cycles of a reactor was realized using the information reported by their processes computer. The simulation was performed with the Simulate-3 software and the results were compared with those of the process computer. The goal is to get the fuel burned, cycle after cycle for obtain the state conditions of the reactor needed for the fuel reload analyses, stability studies and transients analysis, and the development of a methodology that allows to manage and resolve similar cases for future fuel cycles of the nuclear power plant and explore the various options offered by the simulator. (Author)

  4. Multivariable analysis of a failure event of pressure regulator in a BWR; Analisis multivariable de un evento de falla del regulador de presion en un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Castillo D, R.; Ortiz V, J. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Calleros M, G. [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Carretera Cardel-Nautla, Km. 43.5, Veracruz (Mexico)], e-mail: rogelio.castillo@inin.gob.mx

    2009-10-15

    The boiling water reactors can experiment three types of instabilities: one caused by the controllers failure of plant, another renowned instability by reactivity and the last knew as thermal hydraulics instability. An event of pressure regulator failure of electro-hydraulic control of Unit 1 of nuclear power plant of Laguna Verde was analyzed, which caused power oscillations that were increasing their magnitude in the time course. The event has been analyzed using the Fourier transformation in short time for time-frequency analysis and for the frequency domain be employment the power spectral density. Both techniques reported a resonance to oscillation frequency of 0.055 Hz in the power spectrum, this frequency is of observed order of magnitude when fail the reactor control systems. However, these analysis did not allow to study the interrelation of event signals. Of the previous studies, were obtained power spectral densities containing picks and valleys related with the dynamic behaviour of reactor, which includes the control systems performance. For a pick or present valley to a specific frequency in the power spectrum for one of previous variables, can determine the influence of other variables on the pick or valley by relative contribution of power. This method was established in a developed program of name Noise, which uses a multivariable autoregressive model to obtain the autoregressive coefficients, and starting from them the relative contribution of power is determined. Basically two important results were obtained, the first is related with the influence of feed water flow on the other variables to the frequency of 0.055 Hz, the second is related with the instability by reactivity and confirms that this way was not excited during the event. (Author)

  5. Prediction of the local power factor in BWR fuel cells by means of a multilayer neural network; Prediccion del factor local de potencia en celdas de combustible BWR mediante una red neuronal multicapas

    Energy Technology Data Exchange (ETDEWEB)

    Montes, J.L.; Ortiz, J.J.; Perusquia C, R. [ININ, Carretera Mexico-Toluca s/n, 52750 La Marquesa, Ocoyoacac, Estado de Mexico (Mexico); Francois, J.L.; Martin del Campo M, C. [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)]. e-mail: jlmt@nuclear.inin.mx

    2007-07-01

    To the beginning of a new operation cycle in a BWR reactor the reactivity of this it increases by means of the introduction of fresh fuel, the one denominated reload fuel. The problem of the definition of the characteristics of this reload fuel represents a combinatory optimization problem that requires significantly a great quantity of CPU time for their determination. This situation has motivated to study the possibility to substitute the Helios code, the one which is used to generate the new cells of the reload fuel parameters, by an artificial neuronal network, with the purpose of predicting the parameters of the fuel reload cell of a BWR reactor. In this work the results of the one training of a multilayer neuronal net that can predict the local power factor (LPPF) in such fuel cells are presented. The prediction of the LPPF is carried out in those condition of beginning of the life of the cell (0.0 MWD/T, to 40% of holes in the one moderator, temperature of 793 K in the fuel and a moderator temperature of 560 K. The cells considered in the present study consist of an arrangement of 10x10 bars, of those which 92 contains U{sup 235}, some of these bars also contain a concentration of Gd{sub 2}O{sub 3} and 8 of them contain only water. The axial location inside the one assembles of recharge of these cells it is exactly up of the cells that contain natural uranium in the base of the reactor core. The training of the neuronal net is carried out by means of a retro-propagation algorithm that uses a space of training formed starting from previous evaluations of cells by means of the Helios code. They are also presented the results of the application of the neuronal net found for the prediction of the LPPF of some cells used in the real operation of the Unit One of the Laguna Verde Nuclear Power station. (Author)

  6. Design and optimization of a fuel reload of BWR with plutonium and minor actinides; Diseno y optimizacion de una recarga de combustible de BWR con plutonio y actinidos menores

    Energy Technology Data Exchange (ETDEWEB)

    Guzman A, J. R.; Francois L, J. L.; Martin del Campo M, C.; Palomera P, M. A. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, Jiutepec, Morelos 62550 (Mexico)]. e-mail: maestro_juan_rafael@hotmail.com

    2008-07-01

    In this work is designed and optimized a pattern of fuel reload of a boiling water reactor (BWR), whose fuel is compound of uranium coming from the enrichment lines, plutonium and minor actinides (neptunium, americium, curium); obtained of the spent fuel recycling of reactors type BWR. This work is divided in two stages: in the first stage a reload pattern designs with and equilibrium cycle is reached, where the reload lot is invariant cycle to cycle. This reload pattern is gotten adjusting the plutonium content of the assembly for to reach the length of the wished cycle. Furthermore, it is necessary to increase the concentration of boron-10 in the control rods and to introduce gadolinium in some fuel rods of the assembly, in order to satisfy the margin approach of out. Some reactor parameters are presented: the axial profile of power average of the reactor core, and the axial and radial distribution of the fraction of holes, for the one reload pattern in balance. For the design of reload pattern codes HELIOS and CM-PRESTO are used. In the second stage an optimization technique based on genetic algorithms is used, along with certain obtained heuristic rules of the engineer experience, with the intention of optimizing the reload pattern obtained in the first stage. The objective function looks for to maximize the length of the reactor cycle, at the same time as that they are satisfied their limits related to the power and the reactor reactivity. Certain heuristic rules are applied in order to satisfy the recommendations of the fuel management: the strategy of the control cells core, the strategy of reload pattern of low leakage, and the symmetry of a quarter of nucleus. For the evaluation of the parameters that take part in the objective function it simulates the reactor using code CM-PRESTO. Using the technique of optimization of the genetic algorithms an energy of the cycle of 10834.5 MW d/tHM is obtained, which represents 5.5% of extra energy with respect to the

  7. Effect of a Sulphate Transient on the EAC Crack Growth Behaviour of Low-Alloy RPV Steels under Simulated BWR Operating Conditions (CASTOC WP 3, PSI Test 1)

    Energy Technology Data Exchange (ETDEWEB)

    Ritter, S.; Seifert, H. P

    2002-03-01

    Within the CASTOC-project (5th EU FW programme), the environmentally-assisted crack (EAC) growth behaviour of low-alloy reactor pressure vessel (RPV) steels is experimentally investigated under simulated transient and steady-state boiling water reactor (BWR) power operation conditions by six European laboratories. Within WP 3 of this project, the Paul Scherrer Institut (PSI) investigates the effect of water chemistry transients on the EAC crack growth behaviour under periodical partial unloading (PPU) conditions. The present report is a summary of the first PSI test of WP 3 with a Na{sub 2}SO{sub 4} transient. In the first part of the report, the theoretical background on crack growth mechanisms, crack chemistry, mass transport and water chemistry transients as well as a brief literature survey on other water chemistry transient investigations is given. Furthermore, the experimental equipment and test procedure is presented, followed by a summary of the results of PSI test 1 of WP 3. Finally the results are discussed in detail and compared to literature data. In the first part of the experiment, an actively growing EAC crack was generated by PPU in oxygenated high-temperature, high-purity water (T = 288 {sup o}C, DO = 8 ppm, SO{sub 4}{sup 2-} < 0.6 ppb). Then a sulphate transient was applied. The duration ({approx} 300 h) and the amount of sulphate (SO{sub 4}{sup 2-} = 368 ppb) of the applied sulphate transient conservatively covered all sulphate transients, which might occur in BWR/normal water chemistry (NWC) practice. After the transient, outlet conductivity was lowered from ca. 1 {mu}S/cm to less than 0.15 {mu}S/cm within 2.6 h by a 'two-loop technique'. No accelerating effect of the sulphate transient on the EAC crack growth of both tested fracture mechanics specimens under highly oxidising BWR/NWC conditions was observed, making it impossible to deterrnine incubation or delay times. The EAC crack growth rates (CGR) before, during and after the

  8. How far does a doctor's 'duty of care' go?

    Science.gov (United States)

    Torda, A

    2005-05-01

    It is a long-standing tradition in medicine that doctors have an ethical duty to care for all patients who fall within the scope of their skill base. This duty reflects the value system of many doctors and the type of typical dedication to their craft that has long been expected and given. The modern doctor, however, may have other additional roles -- such as those of parent, researcher, business person and many others. What about the duties that accompany these other activities and what if these duties come into conflict with the duty to care for patients? How does a doctor decide how far the duty to care for patients extends? This article explores this question of duty and discusses how the notion of the traditional doctor's duty to care may need to be amended in light of the kinds of lives that doctors now lead.

  9. 7 CFR 982.39 - Duties.

    Science.gov (United States)

    2010-01-01

    ... Orders; Fruits, Vegetables, Nuts), DEPARTMENT OF AGRICULTURE HAZELNUTS GROWN IN OREGON AND WASHINGTON Order Regulating Handling Hazelnut Control Board § 982.39 Duties. The Board shall have among others the... investigate the growing, shipping and marketing conditions with respect to hazelnuts, and assemble data...

  10. Duties of care on the Internet

    NARCIS (Netherlands)

    van Eijk, N.; van Engers, T.; Wiersma, C.; Jasserand, C.; Abel, W.

    2011-01-01

    Internet Service Providers currently find themselves in the spotlight, both in a national and international context, with regard to their relationship both with governments and other private parties, on for example questions of (civil) liability. The paper focuses on duties of care as concerns the r

  11. 38 CFR 3.6 - Duty periods.

    Science.gov (United States)

    2010-07-01

    ... from an injury incurred or aggravated in line of duty or from an acute myocardial infarction, a cardiac... manner in which the travel was performed; and the immediate cause of disability or death. Whenever any... means any of the following: (i) An acute myocardial infarction. (ii) A cardiac arrest. (iii)...

  12. Mental Health Concerns: Veterans & Active Duty

    Science.gov (United States)

    ... NAMI to 741741 Find Help Living with a Mental Health Condition Family Members and Caregivers Teens and Young Adults Veterans & Active Duty Diverse Communities LGBTQ NAMI Programs Discussion Groups NAMI HelpLine Get Involved stigma free Learn how you can help replace stigma ...

  13. NGO Duties in Relation to Human Rights

    NARCIS (Netherlands)

    Philips, J.P.M.|info:eu-repo/dai/nl/298979446

    2010-01-01

    This paper investigates the moral duties that human rights NGOs, such as Amnesty International, and development NGOs, such as Oxfam, have in relation to human rights – especially in relation to the human right to a decent standard of living. The mentioned NGOs are powerful new agents on the global

  14. Machine Tool Series. Duty Task List.

    Science.gov (United States)

    Oklahoma State Dept. of Vocational and Technical Education, Stillwater. Curriculum and Instructional Materials Center.

    This task list is intended for use in planning and/or evaluating a competency-based course to prepare machine tool, drill press, grinding machine, lathe, mill, and/or power saw operators. The listing is divided into six sections, with each one outlining the tasks required to perform the duties that have been identified for the given occupation.…

  15. Duties of care on the Internet

    NARCIS (Netherlands)

    van Eijk, N.; van Engers, T.; Wiersma, C.; Jasserand, C.; Abel, W.

    2011-01-01

    Internet Service Providers currently find themselves in the spotlight, both in a national and international context, with regard to their relationship both with governments and other private parties, on for example questions of (civil) liability. The paper focuses on duties of care as concerns the

  16. The Customs Adjusted Certain Import Duties

    Institute of Scientific and Technical Information of China (English)

    2012-01-01

    In order to adapt to market development, the General Administration of Customs recently has revised the “Classification Table for Imported Articles of the People's Republic of China” and “Duty-Paid Price List for Imported Articles of the People's Republic of China”, and the new standard was put into practice on April 15th, 2012. According to the Announcement No.15, 2012, of the General Administration of Customs, the rate of duty on imports of clothing, accessories, home textiles, and others whose duty paragraph is 04000000 is adjusted to 20%, and the leather clothing and its accessories (including all kinds of leather garments and leather accessories) of duty paragraph 05000000 see the adjustment to 10%. As for the former one, the clothing includes coat, trousers, underwear, shirt/T-shirt, and other clothing; accessories include hats, scarves, headcloth, neckerchief, ties, belts, gloves, socks, handkerchiefs and so on; home textiles refer to blankets, quilts, pillows, bedspreads, sleeping bags, screens, etc.; the others are towels, bath towels, tablecloths, curtains, and carpets.

  17. NGO Duties in Relation to Human Rights

    NARCIS (Netherlands)

    Philips, J.P.M.

    2010-01-01

    This paper investigates the moral duties that human rights NGOs, such as Amnesty International, and development NGOs, such as Oxfam, have in relation to human rights – especially in relation to the human right to a decent standard of living. The mentioned NGOs are powerful new agents on the global s

  18. 7 CFR 1215.30 - Duties.

    Science.gov (United States)

    2010-01-01

    ... 7 Agriculture 10 2010-01-01 2010-01-01 false Duties. 1215.30 Section 1215.30 Agriculture Regulations of the Department of Agriculture (Continued) AGRICULTURAL MARKETING SERVICE (MARKETING AGREEMENTS... statements to be prepared in conformity with generally accepted accounting principles and to be audited by an...

  19. 7 CFR 1209.39 - Duties.

    Science.gov (United States)

    2010-01-01

    ... 7 Agriculture 10 2010-01-01 2010-01-01 false Duties. 1209.39 Section 1209.39 Agriculture Regulations of the Department of Agriculture (Continued) AGRICULTURAL MARKETING SERVICE (MARKETING AGREEMENTS... statements to be prepared in conformity with generally accepted accounting principles and to be audited by an...

  20. 7 CFR 981.39 - Duties.

    Science.gov (United States)

    2010-01-01

    ... growing, shipping, and marketing conditions with respect to almonds and to assemble data in connection... 7 Agriculture 8 2010-01-01 2010-01-01 false Duties. 981.39 Section 981.39 Agriculture Regulations of the Department of Agriculture (Continued) AGRICULTURAL MARKETING SERVICE (Marketing Agreements...

  1. 7 CFR 947.30 - Duties.

    Science.gov (United States)

    2010-01-01

    ... of the Department of Agriculture (Continued) AGRICULTURAL MARKETING SERVICE (Marketing Agreements and... duties of each such person; (e) To investigate, from time to time, and to assemble data on the growing, harvesting, shipping, and marketing conditions with respect to potatoes; (f) To keep minutes, books,...

  2. 7 CFR 958.25 - Duties.

    Science.gov (United States)

    2010-01-01

    ... of the Department of Agriculture (Continued) AGRICULTURAL MARKETING SERVICE (Marketing Agreements and... and define the duties of each such person; (e) To investigate from time to time and to assemble data on the growing, harvesting, shipping and marketing conditions with respect to onions and to engage...

  3. 7 CFR 993.36 - Duties.

    Science.gov (United States)

    2010-01-01

    ..., and assemble data on the producing, handling, shipping, and marketing conditions relative to prunes... of the Department of Agriculture (Continued) AGRICULTURAL MARKETING SERVICE (Marketing Agreements and... Regulating Handling Prune Marketing Committee § 993.36 Duties. The committee shall have, among others,...

  4. 7 CFR 925.29 - Duties.

    Science.gov (United States)

    2010-01-01

    ... investigate and assemble data on the growing, handling, and marketing conditions with respect to grapes; (i... 7 Agriculture 8 2010-01-01 2010-01-01 false Duties. 925.29 Section 925.29 Agriculture Regulations of the Department of Agriculture (Continued) AGRICULTURAL MARKETING SERVICE (Marketing Agreements...

  5. Light duty utility arm startup plan

    Energy Technology Data Exchange (ETDEWEB)

    Barnes, G.A.

    1998-09-01

    This plan details the methods and procedures necessary to ensure a safe transition in the operation of the Light Duty Utility Arm (LDUA) System. The steps identified here outline the work scope and identify responsibilities to complete startup, and turnover of the LDUA to Characterization Project Operations (CPO).

  6. Precautionary rights and duties of states

    NARCIS (Netherlands)

    Trouwborst, A.

    2006-01-01

    This study concerns the definition and implementation of the precautionary principle under general, or customary, international law. A search for patterns and common denominators in state practice resulted in the following definitions of a right and a duty which states are deemed to have under

  7. Mark II containment, supporting program report. Ramshead safety/relief valve loads methodology summary. [BWR

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, G.L.

    1977-10-01

    An overview of information describing and supporting the analytical methods for determining loads associated with safety/relief valve discharge to the suppression pool of the Mark II containment lead plants is presented. SRV discharge phenomena and associated loads, analytical methods for calculating these loads, supporting experimental information, application of the methods to lead plant assessment, and supporting programs are summarily described. The report demonstrates that, with regard to SRV discharge loads on the containment, an adequate technical basis exists for permitting the licensing assessment of the Mark II lead plants to continue on schedule.

  8. Serpent: an alternative for the nuclear fuel cells analysis of a BWR; SERPENT: una alternativa para el analisis de celdas de combustible nuclear de un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Silva A, L.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. Instituto Politecnico Nacional s/n, U.P. Adolfo Lopez Mateos, Edificio 9, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Gomez T, A. M., E-mail: lidi.s.albarran@gmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    In the last ten years the diverse research groups in nuclear engineering of the Universidad Nacional Autonoma de Mexico and Instituto Politecnico Nacional (UNAM, IPN), as of research (Instituto Nacional de Investigaciones Nucleares, ININ) as well as the personnel of the Nuclear Plant Management of the Comision Federal de Electricidad have been using the codes Helios and /or CASMO-4 in the generation of cross sections (X S) of nuclear fuel cells of the cores corresponding to the Units 1 and 2 of the nuclear power plant of Laguna Verde. Both codes belong to the Studsvik-Scandpower Company who receives the payment for the use and their respective maintenance. In recent years, the code Serpent appears among the nuclear community distributed by the OECD/Nea which does not has cost neither in its use neither in its maintenance. The code is based on the Monte Carlo method and makes use of the processing in parallel. In the Escuela Superior de Fisica y Matematicas of the IPN, the personnel has accumulated certain experience in the use of Serpent under the direction of personal of the ININ; of this experience have been obtained for diverse fuel burned, the infinite multiplication factor for three cells of nuclear fuel, without control bar and with control bar for a known thermodynamic state fixed by: a) the fuel temperature (T{sub f}), b) the moderator temperature (T{sub m}) and c) the vacuums fraction (α). Although was not realized any comparison with the X S that the codes Helios and CASMO-4 generate, the results obtained for the infinite multiplication factor show the prospective tendencies with regard to the fuel burned so much in the case in that is not present the control bar like when it is. The results are encouraging and motivate to the study group to continue with the X S generation of a core in order to build the respective library of nuclear data as a following step and this can be used for the codes PARCS, of USA NRC, DYN3D of HZDR, or others developed locally

  9. Aging and defect characterization of motor-operated valves: progress based on NPAR strategy. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Eissenberg, D.M.

    1984-01-01

    The Nuclear Plant Aging Research (NPAR) program strategy is directed at carrying out comprehensive aging assessments in order to define and resolve issues related to aging (including service wear) of electrical and mechanical components and structures at operating reactor facilities and their possible impact on plant safety. This paper describes work recently completed at Oak Ridge National Laboratory which applied the NPAR strategy to motor-operated valves (MOVs). The objective of the work was primarily to develop an understanding of the operating history and conditions and the failure modes of MOVs in nuclear plant service as a preliminary to identifying and recommending methods for trending aging degradation. A second objective was to demonstrate, using MOVs as an example, that the NPAR strategy can be applied to many electrical and mechanical components of nuclear power plants.

  10. 18 CFR 701.77 - Director-duties and responsibilities.

    Science.gov (United States)

    2010-04-01

    ... 18 Conservation of Power and Water Resources 2 2010-04-01 2010-04-01 false Director-duties and... ORGANIZATION Headquarters Organization § 701.77 Director—duties and responsibilities. The Director shall serve... their authorities; and facilitate the work of the Council and the Chairman. His duties...

  11. 78 FR 14166 - Clarification of Flight, Duty, and Rest Requirements

    Science.gov (United States)

    2013-03-05

    ... Federal Aviation Administration 14 CFR Parts 117 and 121 Clarification of Flight, Duty, and Rest... published a final rule on January 4, 2012, that amends the existing flight, duty and rest regulations... questions about the new flight, duty, and rest rule. This is a response to those questions. FOR FURTHER...

  12. 7 CFR 1160.604 - Duties of the referendum agent.

    Science.gov (United States)

    2010-01-01

    ... 7 Agriculture 9 2010-01-01 2009-01-01 true Duties of the referendum agent. 1160.604 Section 1160... Procedure for Conduct of Referenda in Connection with a Fluid Milk Promotion Order § 1160.604 Duties of the referendum agent. The referendum agent, in addition to any other duties imposed by this subpart, shall: (a...

  13. 46 CFR 531.5 - Duty to file.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 9 2010-10-01 2010-10-01 false Duty to file. 531.5 Section 531.5 Shipping FEDERAL... General Provisions § 531.5 Duty to file. (a) The duty under this part to file NSAs, amendments and notices.... (d) Registration—(1) Application. Authority to file or delegate the authority to file must...

  14. 46 CFR 530.5 - Duty to file.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 9 2010-10-01 2010-10-01 false Duty to file. 530.5 Section 530.5 Shipping FEDERAL... Provisions § 530.5 Duty to file. (a) The duty under this part to file service contracts, amendments and... conditions as the parties may agree. (c) Registration—(1) Application. Authority to file or delegate...

  15. 38 CFR 17.31 - Duty periods defined.

    Science.gov (United States)

    2010-07-01

    ... Definitions and Active Duty § 17.31 Duty periods defined. Full-time duty as a member of the Women's Army Auxiliary Corps, Women's Reserve of the Navy and Marine Corps and Women's Reserve of the Coast Guard... Patient Rights...

  16. Results of VGB research work with respect to operation of BWR pipes made of austenitic SS; Ergebnisse des VGB-Forschungsvorhabens zur Absicherung des Betriebsverhaltens austenitischer Staehle in SWR-Rohrleitungen

    Energy Technology Data Exchange (ETDEWEB)

    Kilian, R. [Siemens AG Energieerzeugung KWU, Erlangen (Germany); Bruemmer, G. [Hamburgische Electricitaets-Werke AG, Hamburg (Germany)

    1998-11-01

    The VGB research project was to examine and characterize various, operation-induced impacts on the crack formation in stabilized austenitic steels, caused by intercrystalline stress corrosion cracking as a result of sensitization after chromium depletion at the grain boundaries. The results of this project as well as available operating experience show that the measures taken so far for the future operation of the German BWR plants, for avoiding in these plants intercrystalline stress corrosion cracking, correspond to the state of the art and achieve the wanted purpose. These measures are: use of optimized material W-No. 1.4550 with reduced carbon contents; use of optimized welding techniques for reducing the heat input and the welding shrinkage (cold deformation.); optimized preparation of welding work in order to avoid shape defects during welding (eg. edge misalignment, defective mash welds); reduction of tensile stresses occurring during welding; compliance with the recent VGB water chemistry code. (orig./CB) [Deutsch] Das VGB-Forschungsvorhaben sollte verschieden gelagerte Einfluesse auf die Rissbildung im Betrieb von stabilisierten austenitischen Staehlen, verursacht durch interkristalline Spannungsrisskorrosion infolge Sensbilisierung durch Chromverarmung an den Korngrenzen, systematisch erfassen. Aus den Forschungsergebnissen dieses VGB-Programms sowie den bisher vorliegenden Betriebserfahrungen ist festzuhalten, dass die bisher durchgefuehrten Massnahmen fuer den zukuenftigen Betrieb der deutschen SWR-Anlagen zur Vermeidung von interkristalliner Spannungsrisskorrosion zielgerichtet waren und dem heutigen Wissensstand entsprechen. Diese Massnahmen sind: 1. Einsatz von optimiertem Werkstoff W.-Nr. 1.4550 mit abgesenktem Kohlenstoffgehalt; 2. Einsatz von optimierten Schweissverfahren zur Verminderung der Waermeeinbringung und zur Verringerung des Schweissschrumpfes (Kaltverformung.); 3. Durchfuehrung einer optimierten Schweissnahtvorbereitung zur Vermeidung

  17. Factors influencing helium measurements for detection of control rod failures in BWR

    Energy Technology Data Exchange (ETDEWEB)

    Larsson, I.; Sihver, L. [Div. of Nuclear Engineering, Dept. of Applied Physics, Chalmers Univ. of Technology, SE-412 96 Gothenburg (Sweden); Loner, H.; Ledergerber, G. [Kernkraftwerk Leibstadt, CH-5325 Leibstadt (Switzerland); Schnurr, B. [E.ON Kernkraft GmbH, D-84049 Essenbach (Germany)

    2012-07-01

    Much effort has been made to minimize the number and consequences of fuel failures at nuclear power plants. The consequences of control rod failures have also gained an increased attention. In this paper we introduce a system for on-line surveillance of control rod integrity which has several advantages comparing to the surveillance methods available today in boiling water reactors (BWRs). This system measures the helium released from failed control rods containing boron carbide (B4C). However, there are a number of factors that might influence measurements, which have to be taken into consideration when evaluating the measured data. These factors can be separated into two groups: 1) local adjustments, made on the sampling line connecting the detector to the off-gas system, and 2) plant operational parameters. The adjustments of the sample line conditions include variation of gas flow rate and gas pressure in the line. Plant operational factors that may influence helium measurements can vary from plant to plant. The factors studied at Leibstadt nuclear power plant (KKL) were helium impurities in injected hydrogen gas, variation of the total off-gas flow and regular water refill. In this paper we discuss these factors and their significance and present experimental results of measurements at KKL. (authors)

  18. Optimization of fuel cells for BWR using Path Re linking and flexible strategies of solution;Optimizacion de celdas de combustible para BWR empleando Path Relinking y estrategias flexibles de solucion

    Energy Technology Data Exchange (ETDEWEB)

    Castillo M, J. A.; Ortiz S, J. J.; Torres V, M.; Perusquia del Cueto, R., E-mail: alejandro.castillo@inin.gob.m [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2009-10-15

    In this work are presented the obtained preliminary results to design nuclear fuel cells for boiling water reactors (BWR) using new strategies. To carry out the cells design some of the used rules in the fuel administration were discarded and other were implemented. The above-mentioned with the idea of making a comparative analysis between the used rules and those implemented here, under the hypothesis that it can be possible to design nuclear fuel cells without using all the used rules and executing the security restrictions that are imposed in these cases. To evaluate the quality of the obtained cells it was taken into account the power pick factor and the infinite multiplication factor, in the same sense, to evaluate the proposed configurations and to obtain the mentioned parameters was used the CASMO-4 code. To optimize the design it is uses the combinatorial optimization technique named Path Re linking and the Dispersed Search as local search method. The preliminary results show that it is possible to implement new strategies for the cells design of nuclear fuel following new rules. (Author)

  19. Thermomechanical analysis of a fuel rod in a BWR reactor using the FUELSIM code; Analisis termomecanico de una barra de combustible de un reactor BWR utilizando el codigo FUELSIM

    Energy Technology Data Exchange (ETDEWEB)

    Pantoja C, R. [Escuela Superior de Fisica y Matematicas, Departamento de Ingenieria Nuclear, IPN, Av. Instituto Politecnico Nacional s/n, Col. San Pedro Zacatenco, 07738 Mexico, D. F. (Mexico); Ortiz V, J.; Araiza M, E. [ININ, Departamento de Sistemas Nucleares, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)], e-mail: rapaca78@yahoo.com.mx

    2009-10-15

    The thermomechanical behaviour of a fuel rod exposed to irradiation is a complex process in which are coupled great quantity of interrelated physical-chemical phenomena, for that analysis of rod performance in the core of a nuclear power reactor is realized generally with computation codes that integrate several phenomena expected during the time life of fuel rod in the core. An application of this type of thermomechanical codes is to predict, inside certain reliability margin, the design parameters that would be required to adjust, in order to get a better economy or rod performance, for a systematic approach to the fuel design optimization. FUELSIM is a thermomechanical code based on the models of FRAPCON code, which was developed under auspice of Nuclear Regulatory Commission of USA. FUELSIM allows iterative calculations like part of its programming structure, allowing search of extreme cases of behaviour, probabilistic analysis (or statistical), parametric analysis (or sensibility) and also can include as entrance data to the uncertainties associated with production data, code parameters and associated models. In this work is reported a first analysis of thermomechanical performance of a typical fuel rod used in a BWR 5/6. Results of maximum temperatures are presented in the fuel center and of axial deformation, for the 10 axial nodes in that the active longitude of fuel rod was divided. (Author)

  20. Comparison of results for burning with BWR reactors CASMO and SCALE 6.2 (TRITON / NEWT); Comparacion de los resultados de quemado para reactores BWR con CASMO y SCALE 6.2 (TRITON/NEWT)

    Energy Technology Data Exchange (ETDEWEB)

    Mesado, C.; Miro, R.; Barrachina, T.; Verdu, G.

    2014-07-01

    In this paper we compare the results from two codes burned, CASMO and SCALE 6.2 (TRITON). To do this, is simulated all segments corresponding to a boiling water reactor (BWR) using both codes. In addition, to account for different working points, simulations changing the instantaneous variables, these are repeated: void fractions (6 points), fuel temperature (6 points) and control rods (two points), with a total of 72 possible combinations of different instantaneous variables for each segment. After all simulations are completed for each segment, we can reorder the obtained cross sections, as SCALE CASMO both, to create a library of compositions nemtab format. This format is accepted by the neutronic code of nodal diffusion, PARCS v2.7. Finally compares the results obtained with PARCS and with the SIMULATE3 -SIMTAB methodology to level of full reactor. Also, we have made use of the KENO-VI and MCDANCOFF modules belonging to SCALE. The first is a Monte Carlo transport code with which you can validate the value of the multiplier, the second has been used to obtain values of Dancoff factor and increase the accuracy of model SCALE. (Author)

  1. Simulation of the automatic depressurization system (Ads) for a boiling water reactor (BWR) based on RELAP; Simulacion del sistema de despresurizacion automatica (ADS) para un reactor de agua en ebullicion (BWR) basado en RELAP

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez G, C.; Chavez M, C., E-mail: ces.raga@gmail.com [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2012-10-15

    The automatic depressurization system (Ads) of the boiling water reactor (BWR) like part of the emergency cooling systems is designed to liberate the vapor pressure of the reactor vessel, as well as the main vapor lines. At the present time in the Engineering Faculty, UNAM personnel works in the simulation of the Laguna Verde reactor based on the nuclear code RELAP/SCADAP and in the incorporation to the same of the emergency cooling systems. The simulation of the emergency cooling systems began with the inclusion of two hydrodynamic volumes, one source and another drain, and the incorporation of the initiation logic for each emergency system. In this work is defined and designed a simplified model of Ads of the reactor, considering a detail level based on the main elements that compose it. As tool to implement the proposed model, the RELAP code was used. The simulated main functions of Ads are centered in the quick depressurization of the reactor by means of the vapor discharge through the relief/safety valves to the suppression pool, and, in the event of break of the main vapor line, the reduction of the vessel pressure operates for that the cooling systems of the core to low pressure (Lpcs and Lpci) they can begin their operation. (Author)

  2. Nuclear fuel activity with minor actinides after their useful life in a BWR; Actividad del combustible nuclear con actinidos menores despues de su vida util en un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Martinez C, E.; Ramirez S, J. R.; Alonso V, G., E-mail: eduardo.martinez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2016-09-15

    Nuclear fuel used in nuclear power reactors has a life cycle, in which it provides energy, at the end of this cycle is withdrawn from the reactor core. This used fuel is known as spent nuclear fuel, a strong problem with this fuel is that when the fuel was irradiated in a nuclear reactor it leaves with an activity of approximately 1.229 x 10{sup 15} Bq. The aim of the transmutation of actinides from spent nuclear fuel is to reduce the activity of high level waste that must be stored in geological repositories and the lifetime of high level waste; these two achievements would reduce the number of necessary repositories, as well as the duration of storage. The present work is aimed at evaluating the activity of a nuclear fuel in which radioactive actinides could be recycled to remove most of the radioactive material, first establishing a reference of actinides production in the standard nuclear fuel of uranium at end of its burning in a BWR, and a fuel rod design containing 6% of actinides in an uranium matrix from the enrichment tails is proposed, then 4 standard uranium fuel rods are replaced by 4 actinide bars to evaluate the production and transmutation of the same, finally the reduction of actinide activity in the fuel is evaluated. (Author)

  3. Solution to the transport equation with anisotropic dispersion in a BWR type assembly using the AZTRAN code; Solucion de la ecuacion de transporte con dispersion anisotropica en un ensamble tipo BWR usando el codigo AZTRAN

    Energy Technology Data Exchange (ETDEWEB)

    Chepe P, M. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, San Rafael Atlixco No. 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico); Xolocostli M, J. V.; Gomez T, A. M. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Del Valle G, E., E-mail: liaison.web@gmail.com [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. San Pedro Zacatenco, 07730 Ciudad de Mexico (Mexico)

    2016-09-15

    Due to the current computing power, the deterministic codes for analyzing nuclear reactors that have been used for several years are becoming more relevant, since much more precise solution techniques can be used; the last century would have been very difficult, since memory and processor capacities were very limited or had high prices on the components. In this work we analyze the effect of the anisotropic dispersion of the effective dispersion section, compared to the isotropic dispersion. The anisotropy implementation was carried out in the AZTRAN transport code, which is part of the AZTLAN platform for nuclear reactors analysis (in development). The AZTRAN code solves the Boltzmann transport equation in one, two and three dimensions at steady state, using the multi-group technique for energy discretization, the RTN-0 nodal method in spatial discretization and for angular discretization the discrete ordinates without considering anisotropy originally. The effect of the anisotropy dispersion on the effective multiplication factor and the axial and radial power on a fuel assembly BWR type are analyzed. (Author)

  4. Actinides record, power calculations and activity for present isotopes in the spent fuel of a BWR; Historial de actinidos y calculos de potencia y actividad para isotopos presentes en el combustible gastado de un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Enriquez C, P.; Ramirez S, J. R.; Lucatero, M. A., E-mail: pastor.enriquez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-10-15

    The administration of spent fuel is one of the more important stages of the nuclear fuel cycle, and this has become a problem of supreme importance in countries that possess nuclear reactors. Due to this in this work, the study on the actinides record and present fission products to the discharge of the irradiated fuel in a light water reactor type BWR is shown, to quantify the power and activity that emit to the discharge and during the cooling time. The analysis was realized on a fuel assembly type 10 x 10 with an enrichment average of 3.69 wt % in U-235 and the assembly simulation assumes four cycles of operation of 18 months each one and presents an exposition of 47 G Wd/Tm to the discharge. The module OrigenArp of the Scale 6 code is the computation tool used for the assembly simulation and to obtain the results on the actinides record presents to the fuel discharge. The study covers the following points: a) Obtaining of the plutonium vector used in the fuel production of mixed oxides, and b) Power calculation and activity for present actinides to the discharge. The results presented in this work, correspond at the same time immediate of discharge (0 years) and to a cooling stage in the irradiated fuel pool (5 years). (Author)

  5. Controlling the feedwater flow in a BWR. Examples from Forsmark 2; Regleringen av matarvattenfloedet i en BWR. Med exempel fraan Forsmark 2

    Energy Technology Data Exchange (ETDEWEB)

    Bergdahl, Bengt-Goeran; Oguma, Ritsuo (GSE Power Systems AB, Nykoeping (Sweden))

    2009-03-15

    An investigation of the feedwater controller at Forsmark 2 has been performed. The investigation is based on signal analysis of measurement signals recorded during operation of the plant during different tests. The feedwater controller consists of the water level controller, the flow controller and the condenser balance controller. The overall goal of the feedwater control is to maintain constant water level (level controller) in the reactor and at the same time balance the water levels in the two condensers (condenser balance controller) to avoid that one condenser is full of water while the other one is operated with too low level. There is also a feed forward of the difference between steam flow and feedwater flow (flow controller) for each turbine system with the aim to reduce the fluctuation in reactor water level. The relation in strength between the three controllers is such that the level controller is the strongest followed by the condenser balance controller and finally the flow controller. Tests with trip of the feedwater pump and automatic start of the spare pump in each turbine system indicates a fast reduction in reactor water level that is restored after the transient in the control system. The transient in water level is stable without oscillations. However, it takes about 100 s before the reactor water level is restored. The function of the flow controller has been questioned by the authors. It does not take the action that is expected when a disturbance takes place in the difference between steam and feedwater flow. In addition to this principal weakness there is an offset in the feedwater controller output for feedwater flow 22 that reduces the contribution in flow control that is expected during the introduction of a disturbance. This offset should be adjusted during instrument maintenance of the feedwater controller. The PIP parameters for the level controller are gain factors and time constants. These have been evaluated with the aid of

  6. Radial distribution of UO{sub 2} and Gd{sub 2}O{sub 3} in fuel cells of a BWR Reactor; Distribucion radial de UO{sub 2} y Gd{sub 2}O{sub 3} en celdas de combustible de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Montes, J.L.; Ortiz, J.J.; Perusquia del C, R. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico); Francois, J.L.; Martin del Campo M, C. [Departamento de Sistemas Energeticos, Facultad de Ingenieria, UNAM, Paseo Cuauhnahuac 8532, Jiutepec, Morelos 62500 (Mexico)]. e-mail: jlmt@nuclear.inin.mx

    2008-07-01

    The fuel system that is used at the moment in a power plant based on power reactors BWR, includes as much like the one of its substantial parts to the distribution of the fissile materials like a distribution of burnt poisons within each one of the cells which they constitute the fuel assemblies, used for the energy generation. Reason why at the beginning of a new operation cycle in a reactor of this type, the reactivity of the nucleus should be compensated by the exhaustion of the assemblies that it moves away of the nucleus for their final disposition. This compensation is given by means of the introduction of the recharge fuel, starting from the UO{sub 2} enriched in U{sup 2}35, and of the Gadolinium (Gd{sub 2}O{sub 3}). The distribution of these materials not only defines the requirements of energy generation, but in certain measures also the form in that the margins will behave to the limit them thermal during the operation of the reactor. These margins must be taken into account for the safe and efficient extraction of the energy of the fuel. In this work typical fuel cells appear that are obtained by means of the use of a emulation model of an ants colony. This model allows generating from a possible inventory of values of enrichment of U{sup 2}35, as well as of concentration of Gadolinium a typical fuel cell, which consists of an arrangement of lOxlO rods, of which 92 contain U{sup 2}35, some of these rods contain a concentration of Gd{sub 2}O{sub 3} and 8 of the total contain only water. The search of each cell finishes when the value of the Local Peak Power Factor (LPPF) in the cell reaches a minimal value, or when a pre established value of iterations is reached. The cell parameters are obtained from the results of the execution of the code HELIOS, which incorporates like a part integral of the search algorithm. (Author)

  7. OECD/NEA burnup credit criticality benchmarks phase IIIA: Criticality calculations of BWR spent fuel assemblies in storage and transport

    Energy Technology Data Exchange (ETDEWEB)

    Okuno, Hiroshi; Naito, Yoshitaka [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Ando, Yoshihira [Toshiba Corp., Kawasaki, Kanagawa (Japan)

    2000-09-01

    The report describes the final results of Phase IIIA Benchmarks conducted by the Burnup Credit Criticality Calculation Working Group under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development (OECD/NEA). The benchmarks are intended to confirm the predictive capability of the current computer code and data library combinations for the neutron multiplication factor (k{sub eff}) of a layer of irradiated BWR fuel assembly array model. In total 22 benchmark problems are proposed for calculations of k{sub eff}. The effects of following parameters are investigated: cooling time, inclusion/exclusion of FP nuclides and axial burnup profile, and inclusion of axial profile of void fraction or constant void fractions during burnup. Axial profiles of fractional fission rates are further requested for five cases out of the 22 problems. Twenty-one sets of results are presented, contributed by 17 institutes from 9 countries. The relative dispersion of k{sub eff} values calculated by the participants from the mean value is almost within the band of {+-}1%{delta}k/k. The deviations from the averaged calculated fission rate profiles are found to be within {+-}5% for most cases. (author)

  8. Analysis of the FeCrAl Accident Tolerant Fuel Concept Benefits during BWR Station Blackout Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Robb, Kevin R [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    Iron-chromium-aluminum (FeCrAl) alloys are being considered for fuel concepts with enhanced accident tolerance. FeCrAl alloys have very slow oxidation kinetics and good strength at high temperatures. FeCrAl could be used for fuel cladding in light water reactors and/or as channel box material in boiling water reactors (BWRs). To estimate the potential safety gains afforded by the FeCrAl concept, the MELCOR code was used to analyze a range of postulated station blackout severe accident scenarios in a BWR/4 reactor employing FeCrAl. The simulations utilize the most recently known thermophysical properties and oxidation kinetics for FeCrAl. Overall, when compared to the traditional Zircaloy-based cladding and channel box, the FeCrAl concept provides a few extra hours of time for operators to take mitigating actions and/or for evacuations to take place. A coolable core geometry is retained longer, enhancing the ability to stabilize an accident. Finally, due to the slower oxidation kinetics, substantially less hydrogen is generated, and the generation is delayed in time. This decreases the amount of non-condensable gases in containment and the potential for deflagrations to inhibit the accident response.

  9. Radiological consequence assessments of degraded core accident scenarios derived from a generic Level 2 PSA of a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Homma, Toshimitsu; Ishikawa, Jun; Tomita, Kenichi; Muramatsu, Ken [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-12-01

    The radiological consequence assessments have been made of postulated core damage accidents with source terms derived from a generic Level 2 PSA of a BWR carried out by the Japan Atomic Energy Research Institute (JAERI). The source terms used were for the five core damage accident sequences with the drywell and wetwell failure cases, the release control case by venting of the containment and the accident termination case by the containment spray. The radiological consequences have been assessed for individual dose, collective dose, individual risk of early health effects and individual risk of late health effects by a probabilistic accident consequence assessment code, OSCAAR developed in JAERI. Following conclusions were obtained for the assumed source terms. In case of the over pressure failures of the primary containment vessel, the early fatalities can be mitigated through the implementation of early countermeasures, and the late cancer fatalities remains small. For the release control and accident termination cases, the individual and collective doses to the public can be reduced without any countermeasures due to the release reduction of the volatile radionuclides such as iodine and cesium. (author)

  10. Analyses of containment source term of BWR5 considering iodine chemistry suppression pool with THALES-2 code

    Energy Technology Data Exchange (ETDEWEB)

    Ishikawa, Jun; Moriyama, Kiyofumi [Japan Atomic Energy Agency, Ibaraki (Japan)

    2009-05-15

    After JCO criticality accident in 1999, recognized the importance of PSA application research for emergency planning and basic technical study supporting decision making in protective actions. In order to evaluate containment source term in the late phase SA, coupling of severe accident analysis code THALES-2 and kinetics of iodine chemistry code Kiche was done. And containment source term analyses were performed a typical accident sequence TQUV of BWR5/Mark-II. The lower the pH in the pool was, the more fraction of iodine were released to gas phase, as was in agreement with the known tendency. Total release fractions of all iodine species to gas phase at 40 hr were 0.1[-](pH=5), 0.01[-](pH=7), 4x10{sup -4}[-] (pH=9). I{sub 2} was dominant in released iodine to gas phase and most of released I{sub 2} was adsorbed to the wall. As the operation of the containment spray, the release of iodine tot the gas phase was enhanced due to the break of a steady state by the circulation in the containment. In future, JAEA will perform containment source term analyses for extensive accident sequences with consideration of iodine chemistry.

  11. Electrofishing power requirements in relation to duty cycle

    Science.gov (United States)

    Miranda, L.E.; Dolan, C.R.

    2004-01-01

    Under controlled laboratory conditions we measured the electrical peak power required to immobilize (i.e., narcotize or tetanize) fish of various species and sizes with duty cycles (i.e., percentage of time a field is energized) ranging from 1.5% to 100%. Electrofishing effectiveness was closely associated with duty cycle. Duty cycles of 10-50% required the least peak power to immobilize fish; peak power requirements increased gradually above 50% duty cycle and sharply below 10%. Small duty cycles can increase field strength by making possible higher instantaneous peak voltages that allow the threshold power needed to immobilize fish to radiate farther away from the electrodes. Therefore, operating within the 10-50% range of duty cycles would allow a larger radius of immobilization action than operating with higher duty cycles. This 10-50% range of duty cycles also coincided with some of the highest margins of difference between the electrical power required to narcotize and that required to tetanize fish. This observation is worthy of note because proper use of duty cycle could help reduce the mortality associated with tetany documented by some authors. Although electrofishing with intermediate duty cycles can potentially increase effectiveness of electrofishing, our results suggest that immobilization response is not fully accounted for by duty cycle because of a potential interaction between pulse frequency and duration that requires further investigation.

  12. Preliminary planning study for safety relief valve experiments in a Mark III BWR pressure suppression system

    Energy Technology Data Exchange (ETDEWEB)

    McCauley, E.W.; Holman, G.S.

    1980-04-21

    In response to a request from the Water Reactor Safety Research Division of the US NRC, a preliminary study is provided which identifies key features and consideration involved in planning a comprehensive in-plant Safety Relief Valve experimental program for a Mark III containment design. The report provides identification of program objectives, measurement system requirements, and some details quantifying expected system response. In addition, a preliminary test matrix is outlined which involves a supporting philosophy intended to enhance the usefulness of the experimental results for all members of the program team: experimentalists, analysts, and plant operator.

  13. Fitness for duty in the nuclear power industry: A review of technical issues

    Energy Technology Data Exchange (ETDEWEB)

    Moore, C.; Barnes, V.; Hauth, J.; Wilson, R.; Fawcett-Long, J.; Toquam, J.; Baker, K.; Wieringa, D.; Olson, J.; Christensen, J.

    1989-05-01

    This report presents information gathered and analyzed in support of the US Nuclear Regulatory Commission's (NRC's) efforts to develop a rule that will ensure that workers with unescorted access to protected areas of nuclear power plants are fit for duty. This report supplements information previously published in NUREG/CR-5227, Fitness for Duty in the Nuclear Power Industry: A Review of Technical Issues (Barnes et al., 1988). The primary potential fitness-for-duty concern addressed in both of these reports is impairment caused by substance abuse, although other fitness concerns are discussed. This report addresses issues pertaining to workers' use and misuse of alcohol, prescription drugs, and over-the-counter drugs as fitness-for-duty concerns; responds to several questions raised by NRC Commissioners; discusses subversion of the chemical testing process and methods of preventing such subversion; and examines concerns about the urinalysis cutoff levels used when testing for marijuana metabolites, amphetamines, and phencyclidine (PCP).

  14. Exit and the duty to admit

    Directory of Open Access Journals (Sweden)

    Patti Tamara Lenard

    2015-10-01

    Full Text Available Conventionally, it is presumed that while citizens have the right to exit the state in which they are located, no particular state (except a citizen's home state is required to admit them. Yet, this convention has produced, and continues to produce, injustice; to understand why, I focus on defining and protecting a right to exit, as distinct from the right to move in general. This analysis leads me to propose that whereas the political theoretic literature appears to have converged on a commitment to decisive asymmetry (in favor of accepting a state's right to exclude, I propose that only a weak asymmetry is justified. I argue that receiving states are duty-bound to act in ways that enable migrants to exercise their right to exit. In particular, I argue that receiving states have a perfect duty to collectivize the process by which needy migrants can exercise the right to exit.

  15. Ultraslow extraction with good duty factor

    CERN Document Server

    Cappi, R; Steinbach, C

    1980-01-01

    In the framework of antiproton physics at CERN, a new Low Energy Antiproton Ring (LEAR) is being designed. In the basic mode, it will serve as a beam stretcher giving spill times in the region of an hour. The spill phi (t) must, of course, have a good duty factor ( phi )/sup 2//( phi /sup 2/). A method employing 'stochastic extraction' has been studied theoretically and tried out at the CERN PS ( approximately 9 s flat top) where an extremely good duty factor has been achieved, showing that much longer spill times will be practicable. The pulse length can be varied within wide limits given by the ripple, the momentum acceptance and the intermodulation distortion of the amplifier chain for the noise power. In addition, another method has been found effective which uses empty buckets. These methods need no servo system and both can easily be applied to other synchrotrons. (6 refs).

  16. Melanoma in an Active Duty Marine.

    Science.gov (United States)

    Bartling, Samantha J; Rivard, Shayna C; Meyerle, Jon H

    2017-09-01

    Given that the majority of active duty service members are young and healthy, potentially malignant diagnoses such as skin cancer may be overlooked. Although melanoma accounts for only approximately 1% of skin cancers, it causes the greatest majority of skin cancer deaths. We present the case of a 27-year-old active duty Marine who presented with a hyperpigmented macule at his lateral neck that was a malignant melanoma in situ. This article reviews risk factors for the development of melanoma, offers guidelines for primary care providers, reviews resources for providers in a deployed or austere environment, offers recommendations for prevention and early diagnosis, and discusses follow up. Reprint & Copyright © 2017 Association of Military Surgeons of the U.S.

  17. For a pedagogy of the duty

    Directory of Open Access Journals (Sweden)

    Carlos FERNANDES MAIA

    2016-03-01

    Full Text Available The author starts from some considerations about Froebel's and the New School pedagogy and tries to oppose to that orientation of rights the need of a perspective of the duty. This is the best ethical way to manifest the human fulfilment in the present and the possibility of future perfection. With the story "Natal" (Christmas, from Miguel Torga, the value of an ethic of initiative and sharing as the best way to achieve liberty is exemplified.

  18. 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)

  19. Aging assessment of Westinghouse PWR and General Electric BWR containment isolation functions

    Energy Technology Data Exchange (ETDEWEB)

    Lee, B.S.; Travis, R.; Grove, E.; DiBiasio, A.

    1996-03-01

    A study was performed to assess the effects of aging on the Containment Isolation (CI) functions of Westinghouse Pressurized Water Reactors and General Electric Boiling Water Reactors. This study is part of the Nuclear Plant Aging Research (NPAR) program, sponsored by the U.S. Nuclear Regulatory Commission. The objectives of this program are to provide an understanding of the aging process and how it affects plant safety so that it can be properly managed. This is one of a number of studies performed under the NPAR program which provide a technical basis for the identification and evaluation of degradation caused by age. Failure data from two national databases, Nuclear Plant Reliability Data System (NPRDS) and Licensee Event Reports (LERs), as well as plant specific data were reviewed and analyzed to understand the effects of aging on the CI functions. This study provided information on the effects of aging on component failure frequency, failure modes, and failure causes. Current inspection, surveillance, and monitoring practices were also reviewed.

  20. Clinical negligence and duty of candour.

    Science.gov (United States)

    Shekar, Vinita; Singh, Mark; Shekar, Kishore; Brennan, Peter

    2011-12-01

    The Department of Health is considering imposing a legal duty of candour on health care providers to ensure that an apology and explanation are given to patients when errors occur during medical treatment. This aims to improve quality of care and reduce adverse events during medical treatment. We present the current system of clinical negligence and the future of medical ethics. We discuss relevant cases with regard to duty of candour, and highlight the existence of serious imbalances in which patients' rights and corresponding ethical duties of professionals predominate over the responsibilities of patients themselves. It is known that most adverse events arise because of multiple factors for which no individual should be blamed. To improve healthcare services there is a need for a system in which lessons can be learnt from mistakes, and services can be improved in the interest of patient safety, and for transparency in the broad principles on which the decisions are based within which clinical performance is supervised and monitored. Copyright © 2010 The British Association of Oral and Maxillofacial Surgeons. Published by Elsevier Ltd. All rights reserved.

  1. Solution of a benchmark set problems for BWR and PWR reactors with UO{sub 2} and MOX fuels using CASMO-4; Solucion de un Conjunto de Problemas Benchmark para Reactores BWR y PWR con Combustible UO{sub 2} y MOX Usando CASMO-4

    Energy Technology Data Exchange (ETDEWEB)

    Martinez F, M.A.; Valle G, E. del; Alonso V, G. [IPN, ESFM, 07738 Mexico D.F. (Mexico)]. e-mail: mike_ipn_esfm@hotmail. com

    2007-07-01

    In this work some of the results for a group of benchmark problems of light water reactors that allow to study the physics of the fuels of these reactors are presented. These benchmark problems were proposed by Akio Yamamoto and collaborators in 2002 and they include two fuel types; uranium dioxide (UO{sub 2}) and mixed oxides (MOX). The range of problems that its cover embraces three different configurations: unitary cell for a fuel bar, fuel assemble of PWR and fuel assemble of BWR what allows to carry out an understanding analysis of the problems related with the fuel performance of new generation in light water reactors with high burnt. Also these benchmark problems help to understand the fuel administration in core of a BWR like of a PWR. The calculations were carried out with CMS (of their initials in English Core Management Software), particularly with CASMO-4 that is a code designed to carry out analysis of fuels burnt of fuel bars cells as well as fuel assemblies as much for PWR as for BWR and that it is part in turn of the CMS code. (Author)

  2. Implementation of a Newton-Krylov iterative method to address strong non-linear feedback effects in FORMOSA-B BWR core simulator

    Science.gov (United States)

    Kastanya, Doddy Febrian

    A Newton-BICGSTAB solver has been developed to reduce the CPU execution time of the FORMOSA-B boiling water reactor (BWR) core simulator. The new solver treats the strong non-linearities in the problem explicitly using the Newton's method, replacing the traditionally used nested iterative approach. Taking advantage of the higher convergence rate provided by the Newton's method, assuming that a good initial estimate of the unknowns is provided, and utilizing an efficient preconditioned BICGSTAB solver, we have developed a computationally efficient Newton-BICGSTAB solver to evaluate the three-dimensional, two-group neutron diffusion equations coupled with a two-phase flow model within a BWR core simulator. The robustness of the solver has been tested against numerous BWR core configurations and consistent results have been observed each time. The best exact Newton-BICGSTAB solver performance provides an overall speedup of 2.07 to the core simulator, with reference to the traditional approach, i.e. outer (fission-source)-inner (red/black line SOR). When solving the same problem using the traditional approach but with the BICGSTAB solver as the inner iteration solver [traditional (BICGSTAB)], we observed a speedup of 1.85. This means that the Newton-BICGSTAB solver provides an additional 12% increase in the overall speedup over the traditional (BICGSTAB) solver. However, one needs to note that, on average, the exact Newton-BICGSTAB solver provides an overall speedup of around 1.70; whereas, on average, the traditional (BICGSTAB) provides an overall speedup of around 1.60. An investigation on the feasibility of implementing an inexact Newton-BICGSTAB solver indicates that further reduction in the execution time can likely be obtained through this approach. This study shows that the inexact Newton-BICGSTAB solver can provide speedups of 1.73 to 2.10 with respect to the traditional solver.

  3. Evaluation of potential severe accidents during low power and shutdown operations at Grand Gulf, Unit 1: Evaluation of severe accident risks for plant operational state 5 during a refueling outage. Supporting MELCOR calculations, Volume 6, Part 2

    Energy Technology Data Exchange (ETDEWEB)

    Kmetyk, L.N.; Brown, T.D. [Sandia National Labs., Albuquerque, NM (United States)

    1995-03-01

    To gain a better understanding of the risk significance of low power and shutdown modes of operation, the Office of Nuclear Regulatory Research at the NRC established programs to investigate the likelihood and severity of postulated accidents that could occur during low power and shutdown (LP&S) modes of operation at commercial nuclear power plants. To investigate the likelihood of severe core damage accidents during off power conditions, probabilistic risk assessments (PRAs) were performed for two nuclear plants: Unit 1 of the Grand Gulf Nuclear Station, which is a BWR-6 Mark III boiling water reactor (BWR), and Unit 1 of the Surry Power Station, which is a three-loop, subatmospheric, pressurized water reactor (PWR). The analysis of the BWR was conducted at Sandia National Laboratories while the analysis of the PWR was performed at Brookhaven National Laboratory. This multi-volume report presents and discusses the results of the BWR analysis. The subject of this part presents the deterministic code calculations, performed with the MELCOR code, that were used to support the development and quantification of the PRA models. The background for the work documented in this report is summarized, including how deterministic codes are used in PRAS, why the MELCOR code is used, what the capabilities and features of MELCOR are, and how the code has been used by others in the past. Brief descriptions of the Grand Gulf plant and its configuration during LP&S operation and of the MELCOR input model developed for the Grand Gulf plant in its LP&S configuration are given.

  4. Aging management guideline for commercial nuclear power plants-pumps

    Energy Technology Data Exchange (ETDEWEB)

    Booker, S.; Katz, D.; Daavettila, N.; Lehnert, D. [MDC-Ogden Environmental and Energy Services, Southfield, MI (United States)

    1994-03-01

    This Aging Management Guideline (AMG) describes recommended methods for effective detection and mitigation of age-related degradation mechanisms in BWR and PWR commercial nuclear power plant pumps important to license renewal. The intent of this AMG is to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR Part 54. This AMG is presented in a manner that allows personnel responsible for performance analysis and maintenance to compare their plant-specific aging mechanisms (expected or already experienced) and aging management program activities to the more generic results and recommendations presented herein.

  5. Nuclear power plant simulation on the AD10

    Energy Technology Data Exchange (ETDEWEB)

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

    1985-01-01

    A combination of advanced modeling techniques and the modern, special-purpose peripheral minicomputer AD10 is presented which affords realistic predictions of plant transient and severe off-normal events in LWR power plants through on-line simulations at a speed ten times greater than actual process speeds. Results are shown for a BWR plant simulation. The mathematical models account for nonequilibrium, nonhomogeneous two-phase flow effects in the coolant, for acoustical effects in the steam line and for the dynamics of the recirculation loop and feedwater train. Point kinetics incorporate reactivity feedback for void fraction, for fuel temperature, for coolant temperature, and for boron concentration. Control systems and trip logic are simulated for the nuclear steam supply system. 4 refs., 3 figs.

  6. Medium- and Heavy-Duty Vehicle Duty Cycles for Electric Powertrains

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, Kenneth; Bennion, Kevin; Miller, Eric; Prohaska, Bob

    2016-03-02

    NREL's Fleet Test and Evaluation group has extensive in-use vehicle data demonstrating the importance of understanding the vocational duty cycle for appropriate sizing of electric vehicle (EV) and power electronics components for medium- and heavy-duty EV applications. This presentation includes an overview of recent EV fleet evaluation projects that have valuable in-use data that can be leveraged for sub-system research, analysis, and validation. Peak power and power distribution data from in-field EVs are presented for four different vocations, including class 3 delivery vans, class 6 delivery trucks, class 8 transit buses, and class 8 port drayage trucks, demonstrating the impacts of duty cycle on performance requirements.

  7. A call of duty in hard times: Duty to vote and the Spanish Economic Crisis

    Directory of Open Access Journals (Sweden)

    Carol Galais

    2014-06-01

    Full Text Available Although scarce, the literature addressing the effects of the economy on voter turnout and political attitudes has yielded mixed results. By using individual, longitudinal data from Spain—a country devastated by the Great Recession—our study illuminates how the latest economic crisis has impacted citizens’ perceptions of voting. We analyze how economic conditions and perceptions of the economy have transformed the belief that voting is a civic duty, which is one of the strongest attitudinal predictors of turnout. Our results suggest that hard times slightly weaken citizens’ sense of civic duty, particularly among the youngest. However, the adverse effects of the economic crisis are compensated by the positive effects of the electoral context, and as a consequence there is no aggregate decline in civic duty during the period examined (2010–2012.

  8. Operational, safety and transmutation behavior of BWR with thorium-based fuel; Betriebs-, Sicherheits- und Transmutationsverhalten von SWR mit thoriumbasiertem Brennstoff

    Energy Technology Data Exchange (ETDEWEB)

    Winter, D.; Nabbi, R.; Thomauske, B. [RWTH Aachen (Germany). Inst. fuer Nuklearen Brennstoffkreislauf

    2012-11-01

    The contribution on operational, safety and transmutation behavior of BWR with thorium-based fuel is based on high-resolution simulation models for analysis of the neutron physics in case of heterogeneous flow profiles and neutron spectra in the reactor core using thorium-based fuel. It was shown that thorium-based fuel produces less TRU (transuranium elements) than UO2 fuel. Due to the beneficial neutron physical spectral properties it can be expected that this SWR fuel allows a more effective TRU transmutation. In addition more advantageous safety properties are expected.

  9. Stability analysis of a recycling circuit of a BWR type reactor. Theoretical study; Analisis de estabilidad de un circuito de recirculacion de un reactor del tipo BWR. Estudio teorico

    Energy Technology Data Exchange (ETDEWEB)

    Salinas H, J.G.; Espinosa P, G. [Universidad Autonoma Metropolitana-Iztapalapa, 09000 Mexico D.F. (Mexico); Gonzalez M, V.M. [Comision Nacional de Seguridad Nuclear y Salvaguardias, 04000 Mexico D.F. (Mexico)

    2000-07-01

    The Technology, Regulation and Services Management of the National Commission of Nuclear Safety and Safeguards financed and in coordinate form with the I.P.H. Department of the Metropolitan Autonomous-Iztapalapa University developed the present project with the purpose of studying the effect of the recycling system on the linear stability of a BWR reactor whose reference central is the Laguna Verde power station. The present project forms part of a work series focused to the linear stability of the nuclear reactor of the Unit 1 at Laguna Verde power station. The components of the recycling system considered for the study of stability are the recycling external circuit (recycling pumps, valves) and the internal circuit (downcomer, jet pumps, lower full, driers, separators). The mathematical model is obtained applying mass balances and movement quantity in each one of the mentioned circuits. With respect to the nucleus model two regions are considered, the first one is made of a flow in one phase and the second one of a flow in two phases. For modelling the biphasic region it is considered homogenous flow. Generally it is studied the system behavior in the frequency domain starting from the transfer function applied to four operational states which correspond to the lower stability zone in the map power-flow of the Unit 1 of Laguna Verde power station. The Nyquist diagrams corresponding to each state as well as their characteristic frequency were determined. The results show that exists a very clear dependence of the power-flow relation on the stability of the system. It was found that the boiling length is an important parameter for the linear stability of the system. The obtained results show that the characteristic frequencies in unstability zones are similar to the reported data of the Unit 1 of the Laguna Verde power station in the event of power oscillations carried out in January 1995. (Author)

  10. Hydraulic modeling and simulation of a System Division of Essential Service Water in a BWR plant with Flow master; Modelo hidraulico y simulacion de una division del Sistema de Agua de Servicio Esencial de una central BWR con Flowmaster

    Energy Technology Data Exchange (ETDEWEB)

    Vegazo Juzgado, L.; Rodriguez Garcia, G. M.; Mota Coloma, M.

    2012-07-01

    At the conclusion of the project can say that Flow master is a simulation tool that allows you to create your model from a library of components and obtain useful results from the point of view of the operation, engineering and maintenance. Compared to previous software from the point of view of use, can comment that Flow master is a tool which has an intuitive and user-friendly interaction between the user and the program thus facilitating the modeling of the system and definition of the components of same.

  11. Validation of the CASMO-4 code against SIMS-measured spatial gadolinium distributions inside a BWR pin

    Energy Technology Data Exchange (ETDEWEB)

    Holzgrewe, F.; Gavillet, D.; Restani, R.; Zimmermann, M.A

    2000-07-01

    The purpose of the present study was to establish a database, useful for the assessment of the predictive capabilities of assembly burnup codes with respect to the depletion of the burnable absorber gadolinium (Gd). An SVEA-96 fuel assembly containing one unique Gd rod, with an initial Gd{sub 2}O{sub 3}-content of 9 wt%, was irradiated for one cycle in a Swiss Boiling Water Reactor (BWR), and then transported to the PSI hotcells for post-irradiation examination. Relative radial and azimuthal Gd distributions were obtained from Secondary Ion Mass Spectrometry (SIMS) at three axial positions. Two perpendicular line scans were performed at each position in order to capture the expected asymmetry in the Gd depletion. Since such high-spatial-resolution experimental data for individual fuel pins are quite rare, they form a valuable basis for the further validation of the calculational methods in reactor physics codes. The goal of this study was to contribute to the validation of the micro-region depletion model of CASMO-4 with respect to its standard application of generating two-group cross sections for the 3-D core simulator SIMULATE-3. The only notable difference to the standard application is a more detailed noding scheme for the Gd pin, required to obtain an improved resolution of the calculated distributions. The comparison of measurements with calculational results was found to be quite insensitive to the axial position, and the agreement was found to be very good for all isotopes investigated. The two important neutron-absorbing isotopes {sup 155} Gd and {sup 157} Gd, in particular, show excellent agreement. In conclusion, the CASMO-4 micro-region depletion model has been demonstrated to accurately predict the evolution of the radial distribution of the burnable absorber gadolinium. (authors)

  12. Experimental data report for test TS-2; Reactivity initiated accident test in NSRR with pre-irradiated BWR fuel rod

    OpenAIRE

    1993-01-01

    本報告書は、1990年2月に実施した照射済BWR燃料を用いた2回目の反応度事故模擬実験であるTS-2について実験データをまとめたものである。TS-2実験に使用した試験燃料は初期濃縮度2.79%であり、敦賀1号炉で照射されたBWR7times7型燃料棒を短尺化したものである。短尺化に供した実用燃料のバンドル平均燃焼度は21.3Gwd/tであった。NSRRにおける照射実験は、大気圧、室温の静止水冷却条件下で行い、発熱量は72pm5cal/g・fuel(ピークエンタルピ66pm5cal/g・fuel)を与えた。その結果燃料破損は生じなかった。実験条件、実験方法、パルス照射時の燃料の過渡挙動及び照射後検査の結果をまとめて示した。...

  13. Experimental data report for test TS-1; Reactivity initiated accident test in NSRR with pre-irradiated BWR fuel rod

    OpenAIRE

    1992-01-01

    本報告書は、1989年10月に実施した照射済BWR燃料を用いた最初の反応度事故模擬実験であるTS-1について、実験データをまとめたものである。TS-1実験に使用した試験燃料は、初期濃縮度2.79%であり、敦賀1号炉で照射されたBWR7times7型燃料棒を短尺化したものである。短尺化に供した実用燃料のバンドル平均燃焼度は21.3GWd/tであった。NSRRにおける照射実験は、新たに開発した専用の2重カプセルを用い、大気圧・室温の静止水冷却条件下で行い、発熱量61cal/g・fuel(ピークエンタルピ55cal/g・fuel)を与えた。その結果、燃料破損は生じなかった。実験条件、実験方法、燃料燃焼度の測定結果、パルス照射時の燃料の過渡挙動及び照射後検査の結果をまとめて示した。...

  14. Inspection relief for BWR internal components with noble metal chemical application (NMCA)

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, J.A. [Exelom Corp., Clinton Power Station, Clinton, IL (United States); Pathania, R.S. [EPRI, Palo Alto, CA (United States)

    2001-07-01

    The BWRVIP has developed methods to verify effectiveness of NMCA for mitigation of intergranular stress corrosion cracking (IGSCC). One of the major problems in demonstrating the effectiveness of NMCA inside the reactor vessel is the difficulty of measuring the electrochemical driving force of IGSCC of the various reactor internal components. Many plants do not have direct ECP measurements available at pertinent locations such as the lower plenum. Even those plants that do have direct measurements available recognize that such local measurements may not be representative of all potentially susceptible component surfaces. Therefore, it is desirable to develop valid supplementary techniques that do not depend exclusively on direct measurements of the ECP at specific locations to reliably demonstrate NMCA effectiveness. This paper describes the verification methods developed by the BWRVIP and reports on an industry survey which shows how extensively these recommendations are being implemented. Furthermore, the effectiveness of NMCA is dependent upon the uniform application and the durability of the catalytic surface. Results are provided from the durability monitor at the Duane Arnold Energy Center (DAEC) NMCA demonstration project. Based on the crack growth modeling and radiolysis results, a vessel internals inspection program can be developed based on Factors of Improvement (FOIs) for plants that have implemented NMCA. The FOI calculated for each internal component based on crack growth modeling results can be applied to revise the internals inspection interval established in the various BWRVIP inspection and evaluation (I and E) documents. For an example, this paper discusses revisions to the inspection intervals for recirculation piping welds. (author)

  15. Physical protection of nuclear facilities. Quarterly progress report, July--September 1978. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Chapman, L.D. (ed.)

    1979-01-01

    Major activities during the fourth quarter of FY78 included (1) the vital area analysis of operational reactors and characterization of the Standardized Nuclear Unit Power Plant System (SNUPPS), (2) the algorithm development of a new pathfinding computer code, (3) the completion of contractor-supported work for the component generic data base, (4) the refinement of tests related to human parameters modeling, and (5) the addition of improvements to and demonstration of the Safeguards Automated Facility Evaluation (SAFE), Safeguards Network Analysis Procedure (SNAP), and Fixed-Site Neutralization Model (FSNM) methodologies.

  16. Definition of loss-of-coolant accident radiation source: summary and conclusions. [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Bonzon, L.L.; Lurie, N.A.; Houston, D.H.; Naber, J.A.

    1978-05-01

    The radiation energy release rates and spectra corresponding to those sources specified in USNRC Regulatory Guide 1.89 for the radiation qualification of Class 1E equipment were calculated. The effects of several parameters (some not specific in the Guide), such as reactor fuel composition, operating duration and power level, and treatment of progeny, are evaluated. The results are presented as time-dependent beta and gamma-ray energy release rates and spectra which are fundamental quantities that are not specific to a plant design but are generally applicable to any nuclear power station.

  17. Dose reduction and optimization studies (ALARA) at nuclear power facilities. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Baum, J.W.; Meinhold, C.B.

    1983-01-01

    Brookhaven National Laboratory (BNL) has been commissioned by the Nuclear Regulatory Commission (NRC) to study dose-reduction techniques and effectiveness of as low as reasonably achievable (ALARA) planning at LWR plants. These studies have the following objectives: identify high-dose maintenance tasks; identify dose-reduction techniques; examine incentives for dose reduction; evaluate cost-effectiveness and optimization of dose-reduction techniques; and compile an ALARA handbook on data, engineering modifications, cost-effectiveness calculations, and other information of interest to ALARA practioners.

  18. Fitness for duty in the nuclear industry: Update of the technical issues 1996

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, N.; Grant, T. [eds.] [Battelle Seattle Research Center, WA (United States)

    1996-05-01

    The purpose of this report is to provide an update of information on the technical issues surrounding the creation, implementation, and maintenance of fitness-for-duty (FFD) policies and programs. It has been prepared as a resource for Nuclear Regulatory Commission (NRC) and nuclear power plant personnel who deal with FFD programs. It contains a general overview and update on the technical issues that the NRC considered prior to the publication of its original FFD rule and the revisions to that rule (presented in earlier NUREG/CRs). It also includes chapters that address issues about which there is growing concern and/or about which there have been substantial changes since NUREG/CR-5784 was published. Although this report is intended to support the NRC`s rule making on fitness for duty, the conclusions of the authors of this report are their own and do not necessarily represent the opinions of the NRC.

  19. Effect of pulse duty cycle on Inconel 718 laser welds

    Science.gov (United States)

    McCay, M. H.; McCay, T. D.; Dahotre, N. B.; Sharp, C. M.; Sedghinasab, A.; Gopinathan, S.

    1989-01-01

    Crack sensitive Inconel 718 was laser pulse welded using a 3.0 kW CO2 laser. Weld shape, structure, and porosity were recorded as a function of the pulse duty cycle. Within the matrix studied, the welds were found to be optimized at a high (17 ms on, 7 ms off) duty cycle. These welds were superior in appearance and lack of porosity to both low duty cycle and CW welds.

  20. Antidumping duties, undertakings, and foreign direct investment in the EU.

    OpenAIRE

    Belderbos, René

    2004-01-01

    We study the effects of EU antidumping policy when foreign firms can ‘jump’ antidumping duties through foreign direct investment (FDI) in the EU. We show that duty jumping or duty pre-empting FDI occurs if the EU administration has broader objectives than protecting EU industry's profitability and if cost advantages of foreign firms are transferable abroad. The (expectation of) price undertakings reduces the incentives to engage in FDI and may even discourage FDI as long as products are not t...

  1. Legal consequences of the moral duty to report errors.

    Science.gov (United States)

    Hall, Jacqulyn Kay

    2003-09-01

    Increasingly, clinicians are under a moral duty to report errors to the patients who are injured by such errors. The sources of this duty are identified, and its probable impact on malpractice litigation and criminal law is discussed. The potential consequences of enforcing this new moral duty as a minimum in law are noted. One predicted consequence is that the trend will be accelerated toward government payment of compensation for errors. The effect of truth-telling on individuals is discussed.

  2. Social Duty and Her Function in Communication Strategy of Firm

    OpenAIRE

    TANASOIU Georgiana Lavinia; Enea, Constanta

    2008-01-01

    Social responsibility is not charity, it’s a duty. Today we see all major companies following social responsibility. Social duty is not only attention allotted consumers, customers and contractors, communions and environment, as well employees and implicit their family. In concept triple bottom line, social duty presume achievements of social level, financial plane and environment level and follow a positive impact on society and, in same time, financial achievements. Education is an area...

  3. How cognitive enhancement can change our duties

    Directory of Open Access Journals (Sweden)

    Filippo eSantoni de Sio

    2014-07-01

    Full Text Available This theoretical paper draws the scientific community’s attention to how pharmacological cognitive enhancement may impact on society and law. Namely, if safe, reliable, and effective techniques to enhance mental performance are eventually developed, then this may under some circumstances impose new duties onto people in high-responsibility professions – e.g. surgeons or pilots – to use such substances to minimize risks of adverse outcomes or to increase the likelihood of good outcomes. By discussing this topic, we also hope to encourage scientists to bring their expertise to bear on this current public debate.

  4. The duty to consult and legal obligations

    Energy Technology Data Exchange (ETDEWEB)

    Carpenter, A.W. [Lawson Lundell Lawson and MacIntosh, Calgary, AB (Canada)

    2002-07-01

    Aboriginal law in Canada has been evolving and industry is beginning to engage in the change. This presentation describes the legal aspects regarding Aboriginal rights and the duty to consult First Nations regarding treaty rights. The implications for First Nations and industry are described. Aboriginal peoples of Canada include the Indian, Inuit and Metis populations. Aboriginal titles exist, therefore they are constitutionally protected. The paper describes recent decisions regarding the Mikisew Cree First Nation versus Canada, the Taku River Tlinget versus Ringstad, and the Haida Nation versus British Columbia and Weyerhaeuser.

  5. The Disquietude of Duty Assuming Kant

    Directory of Open Access Journals (Sweden)

    Max Maureira Pacheco

    2014-11-01

    Full Text Available For Kant, the moral duty is determined universally, that is, on account of its form, in the moral norm. However the moral norm is opposed to particularity, determined by what is not the norm itself, hence being the origin of singularity. The singularized norm is opposed, from experience, by its negation in individual cases. To assume Kant demands the reconciliation of the singular, manifested incases, with the universal. This article deals with this question, demonstrating, above all, the practical difficulties linked to the moral experience in its totality.

  6. Critical component wear in heavy duty engines

    CERN Document Server

    Lakshminarayanan, P A

    2011-01-01

    The critical parts of a heavy duty engine are theoretically designed for infinite life without mechanical fatigue failure. Yet the life of an engine is in reality determined by wear of the critical parts. Even if an engine is designed and built to have normal wear life, abnormal wear takes place either due to special working conditions or increased loading.  Understanding abnormal and normal wear enables the engineer to control the external conditions leading to premature wear, or to design the critical parts that have longer wear life and hence lower costs. The literature on wear phenomenon r

  7. Epistemic duties and failure to understand one’s evidence

    Directory of Open Access Journals (Sweden)

    Scott Stapleford

    2012-04-01

    Full Text Available The paper defends the thesis that our epistemic duty is the duty to proportion our beliefs to the evidence we possess. An inclusive view of evidenced possessed is put forward on the grounds that it makes sense of our intuitions about when it is right to say that a person ought to believe some proposition P. A second thesis is that we have no epistemic duty to adopt any particular doxastic attitudes. The apparent tension between the two theses is resolved by applying the concept of duty to belief indirectly.

  8. Comparison of test and earthquake response modeling of a nuclear power plant containment building

    Energy Technology Data Exchange (ETDEWEB)

    Srinivasan, M.G.; Kot, C.A.; Hsieh, B.J.

    1985-01-01

    The reactor building of a BWR plant was subjected to dynamic testing, a minor earthquake, and a strong earthquake at different times. Analytical models simulating each of these events were devised by previous investigators. A comparison of the characteristics of these models is made in this paper. The different modeling assumptions involved in the different simulation analyses restrict the validity of the models for general use and also narrow the comparison down to only a few modes. The dynamic tests successfully identified the first mode of the soil-structure system.

  9. RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - user`s manual

    Energy Technology Data Exchange (ETDEWEB)

    Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.; Neymotin, L.Y.

    1998-03-01

    This document is the User`s Manual for the Boiling Water Reactor (BWR), and Simplified Boiling Water Reactor (SBWR) systems transient code RAMONA-4B. The code uses a three-dimensional neutron-kinetics model coupled with a multichannel, nonequilibrium, drift-flux, phase-flow model of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients. Chapter 1 gives an overview of the code`s capabilities and limitations; Chapter 2 describes the code`s structure, lists major subroutines, and discusses the computer requirements. Chapter 3 is on code, auxillary codes, and instructions for running RAMONA-4B on Sun SPARC and IBM Workstations. Chapter 4 contains component descriptions and detailed card-by-card input instructions. Chapter 5 provides samples of the tabulated output for the steady-state and transient calculations and discusses the plotting procedures for the steady-state and transient calculations. Three appendices contain important user and programmer information: lists of plot variables (Appendix A) listings of input deck for sample problem (Appendix B), and a description of the plotting program PAD (Appendix C). 24 refs., 18 figs., 11 tabs.

  10. Performance of iron-chromium-aluminum alloy surface coatings on Zircaloy 2 under high-temperature steam and normal BWR operating conditions

    Science.gov (United States)

    Zhong, Weicheng; Mouche, Peter A.; Han, Xiaochun; Heuser, Brent J.; Mandapaka, Kiran K.; Was, Gary S.

    2016-03-01

    Iron-chromium-aluminum (FeCrAl) coatings deposited on Zircaloy 2 (Zy2) and yttria-stabilized zirconia (YSZ) by magnetron sputtering have been tested with respect to oxidation weight gain in high-temperature steam. In addition, autoclave testing of FeCrAl-coated Zy2 coupons under pressure-temperature-dissolved oxygen coolant conditions representative of a boiling water reactor (BWR) environment has been performed. Four different FeCrAl compositions have been tested in 700 °C steam; compositions that promote alumina formation inhibited oxidation of the underlying Zy2. Parabolic growth kinetics of alumina on FeCrAl-coated Zy2 is quantified via elemental depth profiling. Autoclave testing under normal BWR operating conditions (288 °C, 9.5 MPa with normal water chemistry) up to 20 days demonstrates observable weight gain over uncoated Zy2 simultaneously exposed to the same environment. However, no FeCrAl film degradation was observed. The 900 °C eutectic in binary Fe-Zr is addressed with the FeCrAl-YSZ system.

  11. Strain-induced corrosion cracking in ferritic components of BWR primary circuits; Risskorrosion in druckfuehrenden ferritischen Komponenten des Primaerkreislaufes von Siedewasserreaktoren

    Energy Technology Data Exchange (ETDEWEB)

    Seifert, H.-P.; Ritter, S.; Ineichen, U.; Tschanz, U.; Gerodetti, B

    2003-04-01

    The present final report of the RIKORR project is a summary of a literature survey and of the experimental work performed by PSI on the environmentally-assisted cracking (EAC) and dynamic strain ageing (DSA) susceptibility of low-alloy steels (LAS) in high-temperature (HT) water. Within this project, the EAC crack growth behaviour of different low-alloy RPV steels, weld filler and weld heat-affected zone materials has been investigated under simulated transient and steady-state BWR/NWC power operation conditions. The strain-induced corrosion cracking (SICC) / low-frequency corrosion fatigue (CF) and stress corrosion cracking (SCC) crack growth behaviour of different low-alloy RPV steels under simulated transient and stationary BWR/NWC conditions was characterized by slow rising load / low-frequency corrosion fatigue and constant load / periodical partial unloading / ripple load tests with pre-cracked fracture mechanics specimens in oxygenated HT water at temperatures of either 288, 250, 200 or 150 {sup o}C. Modern high-temperature water loops, on-line crack growth monitoring and fractographic analysis by scanning electron microscopy (SEM) were used to quantify the cracking response. (author)

  12. 40 CFR 86.1816-08 - Emission standards for complete heavy-duty vehicles.

    Science.gov (United States)

    2010-07-01

    ... U.S. heavy-duty vehicle sales of complete heavy-duty Otto-cycle motor vehicles for model year 2008... complete heavy-duty Otto-cycle motor vehicles for model year 2008. (2)(i) Manufacturers certifying vehicles... Vehicles, Light-Duty Trucks, and Complete Otto-Cycle Heavy-Duty Vehicles § 86.1816-08 Emission...

  13. Source terms: an investigation of uncertainties, magnitudes, and recommendations for research. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Levine, S.; Kaiser, G. D.; Arcieri, W. C.; Firstenberg, H.; Fulford, P. J.; Lam, P. S.; Ritzman, R. L.; Schmidt, E. R.

    1982-03-01

    The purpose of this document is to assess the state of knowledge and expert opinions that exist about fission product source terms from potential nuclear power plant accidents. This is so that recommendations can be made for research and analyses which have the potential to reduce the uncertainties in these estimated source terms and to derive improved methods for predicting their magnitudes. The main reasons for writing this report are to indicate the major uncertainties involved in defining realistic source terms that could arise from severe reactor accidents, to determine which factors would have the most significant impact on public risks and emergency planning, and to suggest research and analyses that could result in the reduction of these uncertainties. Source terms used in the conventional consequence calculations in the licensing process are not explicitly addressed.

  14. Source terms: an investigation of uncertainties, magnitudes, and recommendations for research. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Levine, S.; Kaiser, G. D.; Arcieri, W. C.; Firstenberg, H.; Fulford, P. J.; Lam, P. S.; Ritzman, R. L.; Schmidt, E. R.

    1982-03-01

    The purpose of this document is to assess the state of knowledge and expert opinions that exist about fission product source terms from potential nuclear power plant accidents. This is so that recommendations can be made for research and analyses which have the potential to reduce the uncertainties in these estimated source terms and to derive improved methods for predicting their magnitudes. The main reasons for writing this report are to indicate the major uncertainties involved in defining realistic source terms that could arise from severe reactor accidents, to determine which factors would have the most significant impact on public risks and emergency planning, and to suggest research and analyses that could result in the reduction of these uncertainties. Source terms used in the conventional consequence calculations in the licensing process are not explicitly addressed.

  15. "A Kantian care ethics suicide duty".

    Science.gov (United States)

    Cooley, Dennis R

    2013-01-01

    Standard arguments for a duty to die or to commit suicide generally rely upon contractarian or other form of justice or the Principle of Beneficence. Even though some of these arguments might appear deontological, there is an explicit or implicit consequentialist common thread in all of them in which utility of some sort is maximized only through the taking of one's own life. Hence, most arguments for a suicide duty are consequentialist in nature. There are a number of relatively unexplored deontological arguments that make plausible cases for the mandatory taking of one's own life. For example, although Kant is widely thought to prohibit all suicides, a careful reading of his work can show a plausible case based on the Categorical Imperative. If it is necessary to preserve the individual's moral life, then everyone could will the generalized maxim governing the situation as a law of nature. Unfortunately, Kant's argument is weakened by his poor understanding of moral psychology. To strengthen Kant's case, care-relationship ethics can be combined with the argument to produce a plausible case that people are obligated to kill themselves if a number of criteria are satisfied.

  16. Canonical duties, liabilities of trustees and administrators.

    Science.gov (United States)

    Morrisey, F G

    1985-06-01

    The new Code of Canon Law outlines a number of duties of those who have responsibility for administering the Church's temporal goods. Before assuming office, administrators must pledge to be efficient and faithful, and they must prepare an inventory of goods belonging to the juridic person they serve. Among their duties, administrators must: Ensure that adequate insurance is provided; Use civilly valid methods to protect canonical ownership of the goods; Observe civil and canon law prescriptions as well as donors' intentions; Collect and safeguard revenues, repay debts, and invest funds securely; Maintain accurate records, keep documents secure, and prepare an annual budget; Prepare an annual report and present it to the Ordinary where prescribed; Observe civil law concerning labor and social policy, and pay employees a just and decent wage. Administrators who carry out acts that are invalid canonically are liable for such acts. The juridic person is not liable, unless it derived benefit from the transaction. Liability is especially high when the sale of property is involved or when a contract is entered into without proper cannonical consent. Although Church law is relatively powerless to punish those who have been negligent, stewards, administrators, and trustees must do all they can to be truthful to the responsibility with which they have been entrusted.

  17. Constraining duty cycles through a Bayesian technique

    CERN Document Server

    Romano, P; Segreto, A; Ducci, L; Vercellone, S

    2014-01-01

    The duty cycle (DC) of astrophysical sources is generally defined as the fraction of time during which the sources are active. However, DCs are generally not provided with statistical uncertainties, since the standard approach is to perform Monte Carlo bootstrap simulations to evaluate them, which can be quite time consuming for a large sample of sources. As an alternative, considerably less time-consuming approach, we derived the theoretical expectation value for the DC and its error for sources whose state is one of two possible, mutually exclusive states, inactive (off) or flaring (on), as based on a finite set of independent observational data points. Following a Bayesian approach, we derived the analytical expression for the posterior, the conjugated distribution adopted as prior, and the expectation value and variance. We applied our method to the specific case of the inactivity duty cycle (IDC) for supergiant fast X-ray transients. We also studied IDC as a function of the number of observations in the ...

  18. Rights and duties in parenting practices

    Directory of Open Access Journals (Sweden)

    Ana Paula Solans

    2015-10-01

    Full Text Available The aim of this presentation is to present the results of three qualitative research on the exercise of rights and duties on Parenting Practices (PP, held in Buenos Aires, Argentina. They included interviews with mothers of children with Unsatisfied Basic Needs concretized between 2009 and 2013. Their analysis revealed that in this set of households were carried out three types of PP: imposition, guide and free will, the latter was the most used. As part of this practice, children managed their hours of sleep, wakefulness and leisure, without the intervention of their parents. It was noted, for example, that children over 10 years decided on matters concerning their schooling, absenting progressively to school, to abandonment. These practices were respected by their parents. By default, the postponement of pleasure (tolerance to frustration will not be exercised: they let children do at will. A trend of teenage pregnancy and the formation of pairs of children between 14-16 years with parental consent was also noted. In this sense, even when children lived in a house in contact with their parents, with a supply of food and available school, the indiscriminate exercise of free will put children's health at risk and full development, curtailing their rights. We recommend further studies such timely interventions to promote programs and projects designed to guide parents on issues related to the development of children as subjects of Rights and Duties.

  19. Calibration of burnup monitor in the Rokkasho reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    Oheda, K.; Naito, H.; Hirota, M. [Japan Nuclear Fuel Ltd., Aomori (Japan); Natsume, K. [Toshiba Corp., Yokohama, Kawasaki, Kanagawa (Japan); Kumanomido, H. [Toshiba Corp., Kawasaki, Kanagawa (Japan)

    1998-07-01

    The Rokkasho Reprocessing Plant has adopted a credit for burnup in criticality control in the Spent Fuel Storage Facility (SFSF) and the Dissolution Facility. The burnup monitor system, prepared for BWR and PWR type fuel assemblies, nondestructively measures the burnup value and determines the residual U-235 enrichment in a spent fuel assembly, and criticality is controlled by the value of residual U-235 enrichment in SFSF and by the value of top 50 cm average burnup in the Dissolution Facility. The burnup monitor consists of three measurement systems; a Boss gamma-ray profile measurement system, a high resolution gamma-ray spectrometry system, and a passive neutron measurement system. The monitor sensitivity is calibrated against operator-declared burnup values through repetitive measurements of 100 spent fuel assemblies: BWR 8 X 8, PWR 14 X 14. and 17 X 17. The outline of the measurement methods, objectives of the calibration, actual calibration method, and an example of calibration performed in a demonstration experiment are presented. (author)

  20. Decay profiles of {beta} and {gamma} for a radionuclide inventory in equilibrium cycle of a BWR type reactor; Perfiles de decaimiento de radiacion {beta} y {gamma} para un inventario de radionuclidos en ciclo de equilibrio de un reactor tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Salaices, M.; Sandoval, S.; Ovando, R. [Instituto de Investigaciones Electricas. Gerencia de Energia Nuclear, Av. Reforma 113 Col. Palmira. 62490 Cuernavaca, Morelos (Mexico)]. e-mail: sal@iie.org.mx

    2007-07-01

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

  1. Fuel design with low peak of local power for BWR reactors with increased nominal power; Diseno de un combustible con bajo pico de potencia local para reactores BWR con potencia nominal aumentada

    Energy Technology Data Exchange (ETDEWEB)

    Perusquia C, R.; Montes, J.L.; Hernandez, J.L.; Ortiz, J.J.; Castillo, A. [ININ, 52750 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: mrpc@nuclear.inin.mx

    2006-07-01

    The Federal Commission of Electricity recently announcement the beginning of the works related with the increase of the power to 120% of the original nominal one in the Boiling Water Reactors (BWR) of the Laguna Verde Central (CLV): In the National Institute of Nuclear Research (ININ) are carried out studies of the impact on the design of the recharge of derived fuel of this increase. One of the main effects of the power increase type that it is promoting, is the increment of the flow of generated vapor, what takes, to a bigger fraction of vacuum in the core presenting increased values of the maximum fraction to the limit, so much of the ratio of lineal heat generation (XFLPD) as of the ratio of critic power (MFLCPR). In the made studies, it is found that these fractions rise lineally with the increase of the nominal power. Considering that the reactors of the CLV at the moment operate to 105% of the original nominal power, it would imply an increment of the order of 13.35% in the XFLPD and in the MFLCPR operating to a nominal power of 120% of the original one. This would propitiate bigger problems to design appropriately the fuel cycle and the necessity, almost unavoidable, of to resort to a fuel assembly type more advanced for the recharges of the cores. As option, in the ININ the feasibility of continuing using the same type of it fuel assembles that one has come using recently in the CLV, the type GE12 is analyzed. To achieve it was outlined to diminish the peak factor of local power (LPPF) of the power cells that compose the fuel recharge in 13.35%. It was started of a fuel design previously used in the recharge of the unit 1 cycle 12 and it was re-design to use it in the recharge design of the cycle 13 of the unit 1, considering an increase to 120% of the original power and the same requirements of cycle extension. For the re-design of the fuel assembly cell it was used the PreDiCeldas computer program developed in the ININ. It was able to diminish the LPPF

  2. 76 FR 66089 - Access Authorization Program for Nuclear Power Plants

    Science.gov (United States)

    2011-10-25

    ... COMMISSION Access Authorization Program for Nuclear Power Plants AGENCY: Nuclear Regulatory Commission... revision to Regulatory Guide 5.66, ``Access Authorization Program for Nuclear Power Plants.'' This guide... Authorization Requirements for Nuclear Power Plants,'' and 10 CFR part 26, ``Fitness for Duty Programs.'' The......

  3. Generating human reliability estimates using expert judgment. Volume 2. Appendices. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Comer, M.K.; Seaver, D.A.; Stillwell, W.G.; Gaddy, C.D.

    1984-11-01

    The US Nuclear Regulatory Commission is conducting a research program to determine the practicality, acceptability, and usefulness of several different methods for obtaining human reliability data and estimates that can be used in nuclear power plant probabilistic risk assessments (PRA). One method, investigated as part of this overall research program, uses expert judgment to generate human error probability (HEP) estimates and associated uncertainty bounds. The project described in this document evaluated two techniques for using expert judgment: paired comparisons and direct numerical estimation. Volume 2 provides detailed procedures for using the techniques, detailed descriptions of the analyses performed to evaluate the techniques, and HEP estimates generated as part of this project. The results of the evaluation indicate that techniques using expert judgment should be given strong consideration for use in developing HEP estimates. Judgments were shown to be consistent and to provide HEP estimates with a good degree of convergent validity. Of the two techniques tested, direct numerical estimation appears to be preferable in terms of ease of application and quality of results.

  4. Analysis of Dynamic Insertion of Control Rod of BWR under Seismic Excitation

    Science.gov (United States)

    Koide, Yuichi; Nakagawa, Masaki; Fukushi, Naoki; Ishigaki, Hirokuni; Okumura, Kazue

    The dynamic characteristics of control rod for boiling water reactor being inserted under seismic excitation were investigated using non-linear analytical models. The capability of managing the insertion of control rod is one of the most important factors affecting the safety of nuclear power plant undergoing seismic events. Predicting the behavior of control rod being inserted during earthquakes is important when designing how rod should be controlled during seismic events. We developed analytical models using the finite element method (FEM). The effect of the interaction force between the control rod and the fuel assemblies is considered in non-linear analysis. This interaction force causes resistance force to be applied to the control rod when they are being inserted. The validity of the analytical models was confirmed by comparing the analytical results with the experimental ones. The effects of input seismic motion and structural parameters on the insertion time ware investigated using the analytical models. These analytical methods can be used to predict the time to insert the control rod into the core region of reactor, and are useful for designing control rod system that can survive seismic events.

  5. 33 CFR 5.27 - Assignment to specific duties.

    Science.gov (United States)

    2010-07-01

    ... 33 Navigation and Navigable Waters 1 2010-07-01 2010-07-01 false Assignment to specific duties. 5.27 Section 5.27 Navigation and Navigable Waters COAST GUARD, DEPARTMENT OF HOMELAND SECURITY GENERAL COAST GUARD AUXILIARY § 5.27 Assignment to specific duties. Members of the Auxiliary shall not be...

  6. Classification of Gait Types Based on the Duty-factor

    DEFF Research Database (Denmark)

    Fihl, Preben; Moeslund, Thomas B.

    2007-01-01

    This paper deals with classification of human gait types based on the notion that different gait types are in fact different types of locomotion, i.e., running is not simply walking done faster. We present the duty-factor, which is a descriptor based on this notion. The duty-factor is independent...

  7. 7 CFR 7.25 - County executive director duties.

    Science.gov (United States)

    2010-01-01

    ... 7 Agriculture 1 2010-01-01 2010-01-01 false County executive director duties. 7.25 Section 7.25... CONSERVATION STATE, COUNTY AND COMMUNITY COMMITTEES § 7.25 County executive director duties. (a) The county executive director shall execute the policies established by the county committee and be responsible for...

  8. 77 FR 73911 - Flightcrew Member Duty and Rest Requirements

    Science.gov (United States)

    2012-12-12

    ... and Rest Requirements AGENCY: Federal Aviation Administration (FAA), DOT. ACTION: Availability of... Regulatory Impact Analysis of its final rule amending its existing flight, duty and rest regulations... Register as Flight Crew Member Duty and Rest Requirements on January 4, 2012. 77 FR 330. The regulations...

  9. 75 FR 63424 - Flightcrew Member Duty and Rest Requirements

    Science.gov (United States)

    2010-10-15

    ... Federal Aviation Administration 14 CFR Parts 117 and 121 RIN 2120-AJ58 Flightcrew Member Duty and Rest..., to amend its existing flight, duty and rest regulations applicable to certificate holders and their... and Rest Requirements.'' The proposed regulation recognizes the growing similarities between the types...

  10. 75 FR 62486 - Flightcrew Member Duty and Rest Requirements

    Science.gov (United States)

    2010-10-12

    ... Federal Aviation Administration 14 CFR Parts 117 and 121 RIN 2120-AJ58 Flightcrew Member Duty and Rest... flight, duty and rest regulations applicable to certificate holders and their flightcrew members. The FAA... and Rest Requirements'' (75 FR 55852). The proposed regulation recognizes the growing similarities...

  11. 38 CFR 3.301 - Line of duty and misconduct.

    Science.gov (United States)

    2010-07-01

    ... misconduct. 3.301 Section 3.301 Pensions, Bonuses, and Veterans' Relief DEPARTMENT OF VETERANS AFFAIRS... Entitlement Considerations § 3.301 Line of duty and misconduct. (a) Line of duty. Direct service connection..., and not the result of the veteran's own willful misconduct or, for claims filed after October 31,...

  12. 15 CFR 270.105 - Duties of a Team.

    Science.gov (United States)

    2010-01-01

    ... 15 Commerce and Foreign Trade 1 2010-01-01 2010-01-01 false Duties of a Team. 270.105 Section 270... OF STANDARDS AND TECHNOLOGY, DEPARTMENT OF COMMERCE NATIONAL CONSTRUCTION SAFETY TEAMS NATIONAL CONSTRUCTION SAFETY TEAMS Establishment and Deployment of Teams § 270.105 Duties of a Team. (a) A Team's...

  13. 5 CFR 630.210 - Uncommon tours of duty.

    Science.gov (United States)

    2010-01-01

    ... 5 Administrative Personnel 1 2010-01-01 2010-01-01 false Uncommon tours of duty. 630.210 Section 630.210 Administrative Personnel OFFICE OF PERSONNEL MANAGEMENT CIVIL SERVICE REGULATIONS ABSENCE AND LEAVE Definitions and General Provisions for Annual and Sick Leave § 630.210 Uncommon tours of duty. (a...

  14. 28 CFR 49.2 - Duties of custodian.

    Science.gov (United States)

    2010-07-01

    ... 28 Judicial Administration 2 2010-07-01 2010-07-01 false Duties of custodian. 49.2 Section 49.2 Judicial Administration DEPARTMENT OF JUSTICE (CONTINUED) ANTITRUST CIVIL PROCESS ACT § 49.2 Duties of custodian. (a) Upon taking physical possession of documentary material, answers to interrogatories,...

  15. 40 CFR 1065.512 - Duty cycle generation.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 32 2010-07-01 2010-07-01 false Duty cycle generation. 1065.512... cycle generation. (a) Generate duty cycles according to this section if the standard-setting part... sequence of paired values for speed and torque or for speed and power. (b) Transform normalized values...

  16. 19 CFR 10.625 - Refunds of excess customs duties.

    Science.gov (United States)

    2010-04-01

    ... OF THE TREASURY ARTICLES CONDITIONALLY FREE, SUBJECT TO A REDUCED RATE, ETC. Dominican Republic... last CAFTA-DR country will be liquidated or reliquidated at the applicable rate of duty for that good... excess customs duties paid with respect to such entry, with interest accrued from the date of entry...

  17. 7 CFR 7.22 - Community committee duties.

    Science.gov (United States)

    2010-01-01

    ... 7 Agriculture 1 2010-01-01 2010-01-01 false Community committee duties. 7.22 Section 7.22... CONSERVATION STATE, COUNTY AND COMMUNITY COMMITTEES § 7.22 Community committee duties. (a) The community... community committee shall: (1) Serve as an advisor and consultant to the county committee; (2)...

  18. The Department Head: A Survey of Duties and Responsibilities.

    Science.gov (United States)

    Papalia, Anthony

    This study surveys 107 foreign language departments in secondary schools in western New York and identifies duties and practices of those responsible for the departmental leadership. The report also determines the amount of released time granted to perform departmental duties. The educational preparation and work experience of supervisory staff…

  19. 31 CFR 401.4 - Duties of Bureau of Customs.

    Science.gov (United States)

    2010-07-01

    ... 31 Money and Finance: Treasury 2 2010-07-01 2010-07-01 false Duties of Bureau of Customs. 401.4... TRANSPORT COUNTERFEIT COINS, OBLIGATIONS, SECURITIES, AND PARAPHERNALIA § 401.4 Duties of Bureau of Customs... director of customs pursuant to the said act of August 9, 1939, and the regulations in this part,...

  20. 31 CFR 406.4 - Duties of customs officers.

    Science.gov (United States)

    2010-07-01

    ... 31 Money and Finance: Treasury 2 2010-07-01 2010-07-01 false Duties of customs officers. 406.4... 1934 AND GOLD REGULATIONS § 406.4 Duties of customs officers. The appropriate officials of the Bureau of Customs are hereby authorized and designated as the officers who shall perform such...

  1. Understanding the Code: exceptions to the duty of patient confidentiality.

    Science.gov (United States)

    Griffith, Richard

    2015-07-01

    Last month's article considered the scope of a district nurse's duty to maintain the confidentiality of patient information under the Nursing and Midwifery Council (NMC) Code, their contract of employment, and the law. This month, Richard Griffith considers the exceptions to these duties and sets out when a district nurse would be justified in disclosing patient information.

  2. 46 CFR 515.32 - Freight forwarder duties.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 9 2010-10-01 2010-10-01 false Freight forwarder duties. 515.32 Section 515.32 Shipping... Commission § 515.32 Freight forwarder duties. (a) Notice of shipper affiliation. When a licensed freight... licensed freight forwarder shall have the option of: (1) Identifying itself as such and/or, where...

  3. 28 CFR 549.13 - Programming, duty, and housing restrictions.

    Science.gov (United States)

    2010-07-01

    ... 28 Judicial Administration 2 2010-07-01 2010-07-01 false Programming, duty, and housing... INSTITUTIONAL MANAGEMENT MEDICAL SERVICES Infectious Disease Management § 549.13 Programming, duty, and housing restrictions. (a) The CD will assess any inmate with an infectious disease for appropriateness for...

  4. 42 CFR 440.80 - Private duty nursing services.

    Science.gov (United States)

    2010-10-01

    ... 42 Public Health 4 2010-10-01 2010-10-01 false Private duty nursing services. 440.80 Section 440... nursing services. Private duty nursing services means nursing services for recipients who require more individual and continuous care than is available from a visiting nurse or routinely provided by the...

  5. Duty to report: Legal implications of nurses stealing from patients.

    Science.gov (United States)

    Dimond, Bridgit

    This article explores the situation when a nurse receives a report from a patient that a colleague is stealing from the patient. It looks at the duty of the nurse and the issues which may arise when she tries to put her duty into action. It also considers the legal situation of the colleague who is the subject of the allegations.

  6. Primum nocere: medical brain drain and the duty to stay

    NARCIS (Netherlands)

    Ferracioli, L.; de Lora, P.

    2015-01-01

    In this essay, we focus on the moral justification of a highly controversial measure to redress medical brain drain: the duty to stay. We argue that the moral justification for this duty lies primarily in the fact that medical students impose high risks on their fellow citizens while receiving their

  7. 45 CFR 1618.5 - Duties of the Corporation.

    Science.gov (United States)

    2010-10-01

    ... 45 Public Welfare 4 2010-10-01 2010-10-01 false Duties of the Corporation. 1618.5 Section 1618.5 Public Welfare Regulations Relating to Public Welfare (Continued) LEGAL SERVICES CORPORATION ENFORCEMENT PROCEDURES § 1618.5 Duties of the Corporation. (a) Whenever there is reason to believe that a recipient or...

  8. 38 CFR 1.201 - Employee's duty to report.

    Science.gov (United States)

    2010-07-01

    ... 38 Pensions, Bonuses, and Veterans' Relief 1 2010-07-01 2010-07-01 false Employee's duty to report... PROVISIONS Referrals of Information Regarding Criminal Violations § 1.201 Employee's duty to report. All VA employees with knowledge or information about actual or possible violations of criminal law related to VA...

  9. Student Drivers and the Fiduciary Duty of School Boards

    Science.gov (United States)

    Donlevy, James Kent; Gereluk, Dianne; Brandon, Jim; Patterson, Peggy

    2017-01-01

    Following "E.D.G. v. Hammer", Canadian law has held that school boards, although they have a fiduciary duty to their students, do not guarantee the safety of their students from the acts of their employees. The scope of that fiduciary duty is narrow, restricted to a board acting with disloyalty, in bad faith, or in a conflict of interest…

  10. 25 CFR 11.1003 - Law enforcement officer's duties.

    Science.gov (United States)

    2010-04-01

    ... 25 Indians 1 2010-04-01 2010-04-01 false Law enforcement officer's duties. 11.1003 Section 11.1003 Indians BUREAU OF INDIAN AFFAIRS, DEPARTMENT OF THE INTERIOR LAW AND ORDER COURTS OF INDIAN OFFENSES AND LAW AND ORDER CODE Juvenile Offender Procedure § 11.1003 Law enforcement officer's duties. A law...

  11. 25 CFR 11.1103 - Law enforcement officer's duties.

    Science.gov (United States)

    2010-04-01

    ... 25 Indians 1 2010-04-01 2010-04-01 false Law enforcement officer's duties. 11.1103 Section 11.1103 Indians BUREAU OF INDIAN AFFAIRS, DEPARTMENT OF THE INTERIOR LAW AND ORDER COURTS OF INDIAN OFFENSES AND LAW AND ORDER CODE Minor-in-Need-of-Care Procedure § 11.1103 Law enforcement officer's duties. Upon...

  12. 75 FR 12734 - Honey from Argentina: Rescission of Countervailing Duty Administrative Review

    Science.gov (United States)

    2010-03-17

    ... International Trade Administration Honey from Argentina: Rescission of Countervailing Duty Administrative Review... the countervailing duty order on honey from Argentina. See Antidumping or Countervailing Duty Order..., 2009). On December 31, 2009, the American Honey Producers Association and the Sioux Honey...

  13. 76 FR 1971 - Drill Pipe From the People's Republic of China: Final Affirmative Countervailing Duty...

    Science.gov (United States)

    2011-01-11

    ... bank loan benchmark. Subsequently, on September 14, 2010, the DP Master Group filed rebuttal comments... China: Preliminary Affirmative Countervailing Duty Determination, 75 FR 33245 (June 11, 2010...: Alignment of Final Countervailing Duty Determination with Final Antidumping Duty Determination, 75 FR...

  14. Carbon 14 distribution in irradiated BWR fuel cladding and released carbon 14 after aqueous immersion of 6.5 years

    Energy Technology Data Exchange (ETDEWEB)

    Sakuragi, T. [Radioactive Waste Management Funding and Research Center, Tsukishima 1-15-7, Chuo City, Tokyo, 104-0052 (Japan); Yamashita, Y.; Akagi, M.; Takahashi, R. [TOSHIBA Corporation, Ukishima Cho 4-1, Kawasaki Ward, Kawasaki, 210-0862 (Japan)

    2016-07-01

    Spent fuel cladding which is highly activated and strongly contaminated is expected to be disposed of in an underground repository. A typical activation product in the activated metal waste is carbon 14 ({sup 14}C), which is mainly generated by the {sup 14}N(n,p){sup 14}C reaction and produces a significant exposure dose due to the large inventory, long half-life (5730 years), rapid release rate, and the speciation and consequent migration parameters. In the preliminary Japanese safety case, the release of radionuclides from the metal matrix is regarded as the corrosion-related congruent release, and the cladding oxide layer is regarded as a source of instant release fraction (IRF). In the present work, specific activity of {sup 14}C was measured using an irradiated BWR fuel cladding (Zircaloy-2, average rod burnup of 41.6 GWd/tU) which has an external oxide film having a thickness of 25.3 μm. The {sup 14}C specific activity of the base metal was 1.49*10{sup 4} Bq/g, which in the corresponding burnup is comparable to values in the existing literature, which were obtained from various irradiated claddings. Although the specific activity in oxide was 2.8 times the base metal activity due to the additive generation by the {sup 17}O(n,α){sup 14}C reaction, the {sup 14}C abundance in oxide was less than 10% of total inventory. A static leaching test using the cladding tube was carried out in an air-tight vessel filled with a deoxygenated dilute NaOH solution (pH of 12.5) at room temperature. After 6.5 years, {sup 14}C was found in each leachate fraction of gas phase and dissolved organics and inorganics, the total of which was less than 0.01% of the {sup 14}C inventory of the immersed cladding tube. A simple calculation based on the congruent release with Zircaloy corrosion has suggested that the 96.7% of released {sup 14}C was from the external oxide layer and 3.3% was from the base Zircaloy metal. However, both the {sup 14}C abundance and the low leaching rate

  15. Experimental result of BWR post-CHF tests. Critical heat flux and post-CHF heat transfer coefficient. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Iguchi, Tadashi; Anoda, Yoshinari [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Iwaki, Chikako [Toshiba Corp., Tokyo (Japan)

    2002-02-01

    Authors performed post-CHF experiments under wider pressure ranges of 2 MPa - 18 MPa, wider mass flux ranges of 33 kg/m{sup 2}s - 1651 kg/m{sup 2}s and wider superheat of heaters up to 500 K in comparison to experimental ranges at previous post-CHF experiments. Data on boiling transition, critical heat flux and post-CHF heat transfer coefficient were obtained. Used test section was 4x4-rod bundle with heaters, which diameter and length were the same as those of BWR nuclear fuels. As the result of the experiments, it was found that the boiling transition occurred just below several grid spacers, and that the fronts of the boiling transition region proceeded lower with increase of heated power. Heat transfer was due to nucleate boiling above grid spacers, while it was due to film boiling below grid spacers. Consequently, critical heat flux is affected on the distance from the grid spacers. Critical heat flux above the grid spacers was about 15% higher than that below the grid spacers, by comparing them under the same local condition. Heat transfer by steam turbulent flow was dominant to post-CHF heat transfer, when superheat of heaters was sufficiently high. Then, post-CHF heat transfer coefficient was predicted with heat transfer correlations for single-phase flow. On the other hand, when superhead of heaters was not sufficiently high, post-CHF heat transfer coefficient was higher than the prediction with heat transfer correlations for single-phase flow. Mass flux effect on post-CHF heat transfer coefficient was described by standardization of post-CHF heat transfer coefficient with the prediction for single-phase flow. However, pressure effect, superheat effect and effect of position were not described. Authors clarified that those effects could be described with functions of heater temperature and position. Post-CHF heat transfer coefficient was lowest just blow the grid spacers, and it increased with the lower positions. It increased by about 30% in one span of

  16. Solution of the transport equation in stationary state, in one and two dimensions, for BWR assemblies using nodal methods; Solucion de la ecuacion de transporte en estado estacionario, en 1 y 2 dimensiones, para ensambles tipo BWR usando metodos nodales

    Energy Technology Data Exchange (ETDEWEB)

    Xolocostli M, J.V

    2002-07-01

    The main objective of this work is to solve the neutron transport equation in one and two dimensions (slab geometry and X Y geometry, respectively), with no time dependence, for BWR assemblies using nodal methods. In slab geometry, the nodal methods here used are the polynomial continuous (CMPk) and discontinuous (DMPk) families but only the Linear Continuous (also known as Diamond Difference), the Quadratic Continuous (QC), the Cubic Continuous (CC), the Step Discontinuous (also known as Backward Euler), the Linear Discontinuous (LD) and the Quadratic Discontinuous (QD) were considered. In all these schemes the unknown function, the angular neutron flux, is approximated as a sum of basis functions in terms of Legendre polynomials, associated to the values of the neutron flux in the edges (left, right, or both) and the Legendre moments in the cell, depending on the nodal scheme used. All these schemes were implemented in a computer program developed in previous thesis works and known with the name TNX. This program was modified for the purposes of this work. The program discreetizes the domain of concern in one dimension and determines numerically the angular neutron flux for each point of the discretization when the number of energy groups and regions are known starting from an initial approximation for the angular neutron flux being consistent with the boundary condition imposed for a given problem. Although only problems with two-energy groups were studied the computer program does not have limitations regarding the number of energy groups and the number of regions. The two problems analyzed with the program TNX have practically the same characteristics (fuel and water), with the difference that one of them has a control rod. In the part corresponding to two-dimensional problems, the implemented nodal methods were those designated as hybrids that consider not only the edge and cell Legendre moments, but also the values of the neutron flux in the corner points

  17. Solution of the transport equation in stationary state, in one and two dimensions, for BWR assemblies using nodal methods; Solucion de la ecuacion de transporte en estado estacionario, en 1 y 2 dimensiones, para ensambles tipo BWR usando metodos nodales

    Energy Technology Data Exchange (ETDEWEB)

    Xolocostli M, J.V

    2002-07-01

    The main objective of this work is to solve the neutron transport equation in one and two dimensions (slab geometry and X Y geometry, respectively), with no time dependence, for BWR assemblies using nodal methods. In slab geometry, the nodal methods here used are the polynomial continuous (CMPk) and discontinuous (DMPk) families but only the Linear Continuous (also known as Diamond Difference), the Quadratic Continuous (QC), the Cubic Continuous (CC), the Step Discontinuous (also known as Backward Euler), the Linear Discontinuous (LD) and the Quadratic Discontinuous (QD) were considered. In all these schemes the unknown function, the angular neutron flux, is approximated as a sum of basis functions in terms of Legendre polynomials, associated to the values of the neutron flux in the edges (left, right, or both) and the Legendre moments in the cell, depending on the nodal scheme used. All these schemes were implemented in a computer program developed in previous thesis works and known with the name TNX. This program was modified for the purposes of this work. The program discreetizes the domain of concern in one dimension and determines numerically the angular neutron flux for each point of the discretization when the number of energy groups and regions are known starting from an initial approximation for the angular neutron flux being consistent with the boundary condition imposed for a given problem. Although only problems with two-energy groups were studied the computer program does not have limitations regarding the number of energy groups and the number of regions. The two problems analyzed with the program TNX have practically the same characteristics (fuel and water), with the difference that one of them has a control rod. In the part corresponding to two-dimensional problems, the implemented nodal methods were those designated as hybrids that consider not only the edge and cell Legendre moments, but also the values of the neutron flux in the corner points

  18. Legal duties to respect abortion choices.

    Science.gov (United States)

    Dickens, Bernard M

    2003-01-01

    This paper addresses legal protection of individual choices to obtain abortion services, to decline to perform abortions on grounds of religious objection, and to participate in these procedures. It considers legal duties to respect women as decision-makers in their own lives, including when they decide to continue pregnancy. The choice to decline participation in abortions is an aspect of religious freedom available to physicians, nurses, and, for instance, pharmacists, but not artificial legal persons such as hospital and clinic corporations. Refusal does not extend to ancillary functions such as serving meals, routine pre-operative and post-operative care of abortion patients or typing abortion referral letters. Physicians practising in proximate care must be trained in appropriate medical management of incomplete and threatened abortion even when they would refuse to apply such techniques to induce abortion.

  19. [On the moral dutifulness of using vaccinations].

    Science.gov (United States)

    Refolo, P; González-Melado, F J; Di Pietro, M L

    2015-01-01

    People had contradictory opinions on using vaccinations over time: an initial opposition, later large favour and then doubts and perplexities. In recent times, some movements, blogs and associations stigmatize the use of vaccinations and they are increasingly asking to remove mandatory vaccinations in countries where they are active. The impact of the antivaccination campaigns should not be underestimated, considering that, for example, in Italy, due to these campaigns, adhesions to vaccinations are decreasing by 1% per year, and in reference to rubella and measles, adhesions decreased by 25% in some regions of the country. Overcoming the choice between mandatory and recommended vaccinations, the paper deals with the topic of using preventive immunization starting from the concept of "moral dutifulness".

  20. Lightweight Composite Materials for Heavy Duty Vehicles

    Energy Technology Data Exchange (ETDEWEB)

    Pruez, Jacky; Shoukry, Samir; Williams, Gergis; Shoukry, Mark

    2013-08-31

    The main objective of this project is to develop, analyze and validate data, methodologies and tools that support widespread applications of automotive lightweighting technologies. Two underlying principles are guiding the research efforts towards this objective: • Seamless integration between the lightweight materials selected for certain vehicle systems, cost-effective methods for their design and manufacturing, and practical means to enhance their durability while reducing their Life-Cycle-Costs (LCC). • Smooth migration of the experience and findings accumulated so far at WVU in the areas of designing with lightweight materials, innovative joining concepts and durability predictions, from applications to the area of weight savings for heavy vehicle systems and hydrogen storage tanks, to lightweighting applications of selected systems or assemblies in light–duty vehicles.

  1. Genetic ignorance, moral obligations and social duties.

    Science.gov (United States)

    Takala, T; Häyry, M

    2000-02-01

    In a contribution to The Journal of Medicine and Philosophy, Professor Rosamond Rhodes argues that individuals sometimes have an obligation to know about their genetic disorders, because this is required by their status as autonomous persons. Her analysis, which is based on Kant's concept of autonomy and Aristotle's notion of friendship, is extended here to consequentialist concerns. These are of paramount importance if, as we believe and Professor Rhodes herself implies, the Kantian and Aristotelian doctrines can be helpful only in the sphere of private morality, not in the public realm. Better tools for assessing the right to genetic ignorance as an issue of public policy can, we contend, be found in Mill's ideas concerning liberty and the prevention of harm. Our own conclusion, based on the Millian way of thinking, is that individuals probably do have the right to remain in ignorance in the cases Professor Rhodes presents as examples of a duty to know.

  2. Medium Duty Electric Vehicle Demonstration Project

    Energy Technology Data Exchange (ETDEWEB)

    Mackie, Robin J. D. [Smith Electric Vehicles Corporation, Kansas City, MO (United States)

    2015-05-31

    The Smith Electric Vehicle Demonstration Project (SDP) was integral to the Smith business plan to establish a manufacturing base in the United States (US) and produce a portfolio of All Electric Vehicles (AEV’s) for the medium duty commercial truck market. Smith focused on the commercial depot based logistics market, as it represented the market that was most ready for the early adoption of AEV technology. The SDP enabled Smith to accelerate its introduction of vehicles and increase the size of its US supply chain to support early market adoption of AEV’s that were cost competitive, fully met the needs of a diverse set of end users and were compliant with Federal safety and emissions requirements. The SDP accelerated the development and production of various electric drive vehicle systems to substantially reduce petroleum consumption, reduce vehicular emissions of greenhouse gases (GHG), and increase US jobs.

  3. Good faith in corporate law – an independent fiduciary duty or an element of the duty of loyalty?

    Directory of Open Access Journals (Sweden)

    Adina Ponta

    2016-12-01

    Full Text Available Taking the duty of loyalty as a starting point, which we consider to be the director’s core fiduciary duty, this paper aims at identifying the contours of good faith in corporate law and the interpretations of this institution in corporate governance. The objective of the paper is to demonstrate the autonomy of good faith, along with the duty of care and the duty of loyalty. The paper displays the traditional legal approaches of this institution, both in continental civil law and in common law literature and jurisprudence and exhaustively describes the obligations that compose or even define this concept. Due to its amplitude, the duty of good faith enabled courts to articulate subsidiary fiduciary duties that meet social changes and transformation within business law. By means of cited case law, the conclusion will show that due to the nature, content and effects of situations where specific obligations are met, these may not be incorporated as elements of the traditional duty of care or duty of loyalty.

  4. Study of instabilities in phase by using the tool {sup D}ynamics{sup :} analysis of the evolution space temporary of the waves of density in channels of reactors BWR; Estudio de las Inestabilidades en Fase Mediante la Herramienta Dinamics: analisis de la Evolucion Espacio Temporal de las Ondas de Densidad en Canales de Reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Munoz-Cobo, J. L.; Escriva, R.; Merino, R.; Melara, J.

    2013-07-01

    This paper presents the basics of Dynamics V2 to code It allows calculations of stability for oscillations in phase in BWR reactors in the time domain. The equations of the model are exposed and is the integration of the equations. The model can be used in a large number of nodes thrust for the calculations to an acceptable computational cost, it has simplified dynamics of recirculation loop and the code has been incorporated the Oscillation in phase boundary conditions. The code incorporates the equations of boiling sub-cooled which allows to make more realistic calculations as well as subroutines to calculate the subroutines-based properties of the MATPRO and ASME.

  5. The material concept in German light water reactors. Contribution to plant safety economic efficiency and failure provision; Das Werkstoffkonzept in deutschen Leichtwasserreaktoren. Beitrag zur Anlagensicherheit, Wirtschaftlichkeit und Schadensvorsorge

    Energy Technology Data Exchange (ETDEWEB)

    Ilg, Ulf [EnBW Kernkraft GmbH (Germany). Kernkraftwerk Philippsburg; Koenig, Guenter [EnBW Kernkraft GmbH (Germany). Kernkraftwerk Neckarwestheim; Erve, Manfred [AREVA NP GmbH, Erlangen (Germany)

    2008-07-01

    In the design and construction stage of nuclear power plants relevant decisions may affect the service life of a component, and thus influence safety and availability of the plant. The German ''basic safety concept'' has an important effect on the quality of the BOL (begin of life) status. Materials selection and qualification are of significant importance for the component lifetime and the profitability of the plant. Examples for the implementation of this concept are demonstrated for the steam generator tubing material Incoloy 800, the inside-plated ferritic compound tubes as control rod drive mechanism nozzle through the RPV head of BWR plants that are not susceptible for corrosion enhanced cracking that was observed for Inconel 600 tubing. A fundamental failure analysis of crack formation in Ti stabilized austenitic pipes of BWR plants found since 1993 were definitely identified as intergranular stress corrosion caused by a local sensitization of the welding process induced overheated structured in the heat affected zone. This allowed target-oriented mitigation measures. The safety culture implemented in German nuclear plants in connection with the break preclusion or integrity concept, respectively, including a continuous actualization with respect to the state-of-the art are the technical prerequisites for damage precaution and possible life time extension.

  6. Calculations for a BWR Lattice with Adjacent Gadolinium Pins Using the Monte Carlo Cell Code Serpent v.1.1.7

    Directory of Open Access Journals (Sweden)

    Diego Ferraro

    2011-01-01

    Full Text Available Monte Carlo neutron transport codes are usually used to perform criticality calculations and to solve shielding problems due to their capability to model complex systems without major approximations. However, these codes demand high computational resources. The improvement in computer capabilities leads to several new applications of Monte Carlo neutron transport codes. An interesting one is to use this method to perform cell-level fuel assembly calculations in order to obtain few group constants to be used on core calculations. In the present work the VTT recently developed Serpent v.1.1.7 cell-oriented neutronic calculation code is used to perform cell calculations of a theoretical BWR lattice benchmark with burnable poisons, and the main results are compared to reported ones and with calculations performed with Condor v.2.61, the INVAP's neutronic collision probability cell code.

  7. Preliminary Assessment of the Possible BWR Core/Vessel Damage States for Fukushima Daiichi Station Blackout Scenarios Using RELAP/SCDAPSIM

    Directory of Open Access Journals (Sweden)

    C. M. Allison

    2012-01-01

    Full Text Available Immediately after the accident at Fukushima Daiichi, Innovative Systems Software and other members of the international SCDAP Development and Training Program started an assessment of the possible core/vessel damage states of the Fukushima Daiichi Units 1–3. The assessment included a brief review of relevant severe accident experiments and a series of detailed calculations using RELAP/SCDAPSIM. The calculations used a detailed RELAP/SCDAPSIM model of the Laguna Verde BWR vessel and related reactor cooling systems. The Laguna Verde models were provided by the Comision Nacional de Seguridad Nuclear y Salvaguardias, the Mexican nuclear regulatory authority. The initial assessment was originally presented to the International Atomic Energy Agency on March 21 to support their emergency response team and later to our Japanese members to support their Fukushima Daiichi specific analysis and model development.

  8. Effects of hydrogen water chemistry on corrosion fatigue behavior of cold-worked 304L stainless steel in simulated BWR coolant environments

    Energy Technology Data Exchange (ETDEWEB)

    Chiang, M.F., E-mail: mfchiang@iner.gov.tw [Institute of Nuclear Energy Research, Division of Nuclear Fuels and Materials, Lungtan, Taoyuan 325, Taiwan (China); Young, M.C.; Huang, J.Y. [Institute of Nuclear Energy Research, Division of Nuclear Fuels and Materials, Lungtan, Taoyuan 325, Taiwan (China)

    2011-04-15

    Corrosion fatigue behavior of stainless steel 304L (SS304L) in a simulated BWR coolant with hydrogen injection was investigated. Hydrogen water chemistry slightly mitigated the corrosion fatigue degradation of the as-received SS304L specimens, but, on the contrary, it slightly increased the corrosion fatigue crack growth rates (CFCGRs) of the cold-worked specimens. All the CFCGR-tested specimens showed similar fracture features, except for the amounts of deposited corrosion debris. The results indicated that decreasing the oxygen concentration of water environment is not an effective measure to suppress the fatigue crack growth rate of cold-worked SS304L. The CFCGRs of the SS304L were determined by an interaction between corrosion, oxide-induced crack closure and cold work in corrosive environments. At a specific level of reduction, cold work could enhance the corrosion fatigue resistance of SS304 both in the air-saturated and HWC coolant environments.

  9. Effects of hydrogen water chemistry on corrosion fatigue behavior of cold-worked 304L stainless steel in simulated BWR coolant environments

    Science.gov (United States)

    Chiang, M. F.; Young, M. C.; Huang, J. Y.

    2011-04-01

    Corrosion fatigue behavior of stainless steel 304L (SS304L) in a simulated BWR coolant with hydrogen injection was investigated. Hydrogen water chemistry slightly mitigated the corrosion fatigue degradation of the as-received SS304L specimens, but, on the contrary, it slightly increased the corrosion fatigue crack growth rates (CFCGRs) of the cold-worked specimens. All the CFCGR-tested specimens showed similar fracture features, except for the amounts of deposited corrosion debris. The results indicated that decreasing the oxygen concentration of water environment is not an effective measure to suppress the fatigue crack growth rate of cold-worked SS304L. The CFCGRs of the SS304L were determined by an interaction between corrosion, oxide-induced crack closure and cold work in corrosive environments. At a specific level of reduction, cold work could enhance the corrosion fatigue resistance of SS304 both in the air-saturated and HWC coolant environments.

  10. Role of Customs Duties in the Formation of Budget Revenues

    Directory of Open Access Journals (Sweden)

    Mirela - Anca Postole

    2013-02-01

    Full Text Available Accession to the European Union, starting price and trade exchange liberalization, alongside a strong exchange rate reform, required a depth rethinking of the customs duty system and also influenced the role of this category of tax in establishing budgetary resources.This study reviews the impact of customs duties on changing levels of revenues collected at the state budget. The analysis used is the econometric modeling based on a single- factor regression model.But in Romania, customs duties do not have any major impact on budget revenues and the effects of their collection on the state budget revenues are felt within two months of collection.

  11. IMPACT OF CPO EXPORT DUTIES ON MALAYSIAN PALM OIL INDUSTRY

    OpenAIRE

    Ibragimov Abdulla; Fatimah Mohamed Arshad; B. K. Bala; Kusairi Mohd Noh; Muhammad Tasrif

    2014-01-01

    In January 2013, Malaysia reduced the export duty structure to be in line with the Indonesia’s duty structure. Both countries export crude and processed palm oil. Since Malaysia and Indonesia are close competitors and they compete in the same market, a change in export duty rate in one country will affect the other. Indonesia, as the world’s biggest palm oil producer, has drastically widened the values between the crude palm oil and refined palm oil export taxes since October 2011...

  12. The Perceived Importance of HR duties to Danish line managers

    DEFF Research Database (Denmark)

    Brandl, Julia; Madsen, Mona Toft; Madsen, Henning

    2009-01-01

    of particular HR duties and how the importance assigned to HR duties varies across managers. Based on a survey of 1,500 Danish managers, we find that 'motivating others' is considered the most important HR duty whereas 'team building', 'handling conflicts' and 'coaching' are considered the least important HR......Today, HR scholars widely acknowledge that realising HRM requires the involvement of all managers and that the personal motivation of line managers plays an important role in their successful involvement. Yet, previous research has neglected to study how line managers rate the importance...

  13. Boiling water reactors with uranium-plutonium mixed oxide fuel. Report 5: Analysis of the reactivity coefficients and the stability of a BWR loaded with MOx fuel

    Energy Technology Data Exchange (ETDEWEB)

    Demaziere, C. [CEA Centre d' Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Reacteurs Nucleaires

    2000-01-01

    This report is a part of the project titled 'Boiling Water Reactors With Uranium-Plutonium Mixed Oxide (MOx) Fuel'. The aim of this study is to model the impact of a core loading pattern containing MOx bundles upon the main characteristics of a BWR (reactivity coefficients, stability, etc.). For this purpose, the Core Management System (CMS) codes of Studsvik Scandpower are used. This package is constituted by CASMO-4/TABLES-3/SIMULATE-3. It has been shown in previous reports that these codes are able to accurately represent and model MOx bundles. This report is thus devoted to the study of BWR cores loaded (partially or totally) with MOx bundles. The plutonium quality used is the Pu type 2016 (mostly Pu-239, 56 %, and Pu-240, 26 %), but a variation of the plutonium isotopic vector was also investigated, in case of a partial MOx loading. One notices that the reactivity coefficients do not present significant changes in comparison with a full UOx loading. Nevertheless, two main problems arise: the shutdown margin at BOC is lower than 1 % and the stability to in-phase oscillations is slightly decreased. (The SIMULATE-3 version used for this study does not contain the latest MOx enhancements described in literature, since these code developments have not been provided to the department. Nevertheless, as the nominal average enrichment of the MOx bundles is 5.41 % (total amount of plutonium), which can still be considered as a relatively low enrichment, the accuracy of the CMS codes is acceptable without the use of the MOx improvements for this level of Pu enrichment.

  14. 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)

  15. Development and validation of advanced CFD models for detailed predictions of void distribution in a BWR bundle

    Science.gov (United States)

    Neykov, Boyan

    In recent years, a commonly adopted approach is to use Computational Fluid Dynamics (CFD) codes as computational tools for simulation of different aspects of the nuclear reactor thermal-hydraulic performance where high-resolution and high-fidelity modeling is needed. Within the framework of this PhD work, the CFD code STAR-CD [1] is used for investigations of two phase flow in air-water systems as well as boiling phenomena in simple pipe geometry and in a Boiling Water Reactor (BWR) fuel assembly. Based on the two-fluid Eulerian solver, improvements of the STAR-CD code in the treatment of the drag, lift and wall lubrication forces in a dispersed two phase flow at high vapor (gas) phase fractions are investigated and introduced. These improvements constitute a new two phase modeling framework for STAR-CD, which has been shown to be superior as compared to the default models in STAR-CD. The conservation equations are discretized using the finite-volume method and solved using a solution procedure is based on Pressure Implicit with Splitting of Operators (PISO) algorithm, adapted to the solution of the two-fluid model. The improvements in the drag force modeling include investigation and integration of models with dependence on both void fraction and bubble diameter. The set of the models incorporated into STAR-CD is selected based on an extensive literature review focused on two phase systems with high vapor fractions. The research related to the modeling of wall lubrication force is focused on the validation of the already existing model in STAR-CD. The major contribution of this research is the development and implementation of an improved correlation for the lift coefficient used in the lift force formula. While a variety of correlations for the lift coefficient can be found in the open literature, most of those were derived from experiments conducted at low vapor (gas) phase fractions and are not applicable to the flow conditions existing in the BWRs. Therefore

  16. The Perceived Importance of HR duties to Danish line managers

    DEFF Research Database (Denmark)

    Brandl, Julia; Madsen, Mona Toft; Madsen, Henning

    2009-01-01

    Today, HR scholars widely acknowledge that realising HRM requires the involvement of all managers and that the personal motivation of line managers plays an important role in their successful involvement. Yet, previous research has neglected to study how line managers rate the importance...... of particular HR duties and how the importance assigned to HR duties varies across managers. Based on a survey of 1,500 Danish managers, we find that 'motivating others' is considered the most important HR duty whereas 'team building', 'handling conflicts' and 'coaching' are considered the least important HR...... duties. Female top managers in the public sector exhibit the greatest interest in HR whereas men at lower managerial levels in the private sector give lowest priority to HR work. We conclude with possible explanations for the observed differences in a Danish context and beyond and provide suggestions...

  17. Evolution of high duty cycle echolocation in bats

    DEFF Research Database (Denmark)

    Fenton, M. B.; Faure, P. A.; Ratcliffe, J. M.

    2012-01-01

    Duty cycle describes the relative 'on time' of a periodic signal. In bats, we argue that high duty cycle (HDC) echolocation was selected for and evolved from low duty cycle (LDC) echolocation because increasing call duty cycle enhanced the ability of echolocating bats to detect, lock onto and track...... fluttering insects. Most echolocators (most bats and all birds and odontocete cetaceans) use LDC echolocation, separating pulse and echo in time to avoid forward masking. They emit short duration, broadband, downward frequency modulated (FM) signals separated by relatively long periods of silence....... In contrast, bats using HDC echolocation emit long duration, narrowband calls dominated by a single constant frequency (CF) separated by relatively short periods of silence. HDC bats separate pulse and echo in frequency by exploiting information contained in Doppler-shifted echoes arising from their movements...

  18. All Digital Wide Range Msar Controlled Duty-Cycle Corrector

    Directory of Open Access Journals (Sweden)

    K.Sindhuja

    2014-07-01

    Full Text Available A clock with 50% duty cycle is very significant in many applications such as DDR-SDRAMs and double sampling analog-to-digital converters. This crisp presents a Modified Successive Approximation Register (MSAR controlled duty cycle corrector (DCC, to attain 50% duty cycle correction. Here MSAR adopts a binary search method to compress lock time while maintaining tight synchronization between effort and production clocks. The MSAR-DCC circuit has been implemented in a 0.18- µm CMOS process which corrects the duty rate within 5 cycles which has a closed loop characteristics. The measured power dissipation and area occupation are 5581nW and 0.033mm2 respectively.

  19. 7 CFR 46.26 - Duties of licensees.

    Science.gov (United States)

    2010-01-01

    ... Regulations of the Department of Agriculture AGRICULTURAL MARKETING SERVICE (Standards, Inspections, Marketing Practices), DEPARTMENT OF AGRICULTURE MARKETING OF PERISHABLE AGRICULTURAL COMMODITIES REGULATIONS (OTHER THAN RULES OF PRACTICE) UNDER THE PERISHABLE AGRICULTURAL COMMODITIES ACT, 1930 Duties of...

  20. The duty to do the best for one's patient.

    Science.gov (United States)

    Crisp, Roger

    2015-03-01

    This paper is a discussion of the duty of doctors to do what is best for their patients. What is required by this duty is shown to depend on the circumstances, including any financial constraints on the doctor. The duty to do the best is a duty of benevolence, and this virtue itself has to be understood as bounded by other virtues, including justice and professional responsibility. An Aristotelian account of medical benevolence is developed, and the issues of supererogation and individual judgement are discussed within this framework. The paper ends with the claim that the patient-centred conception of benevolence defended in the paper is in line with consequentialist and deontological ethical traditions. Published by the BMJ Publishing Group Limited. For permission to use (where not already granted under a licence) please go to http://group.bmj.com/group/rights-licensing/permissions.

  1. 7 CFR 917.34 - Duties of Control Committee.

    Science.gov (United States)

    2010-01-01

    ... CALIFORNIA Order Regulating Handling Administrative Bodies § 917.34 Duties of Control Committee. The Control... develop and provide the commodity committees data on shared expenses to facilitate equitable...

  2. Analysis of the noise of the jet pumps of the Unit 2 of the Laguna Verde nuclear power plant; Analisis de ruido de las bombas de chorro de la Unidad 2 de la Central Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Castillo D, R.; Ortiz V, J.; Ruiz E, J.A. [ININ, A.P. 18-1027, 11801 Mexico D.F. (Mexico); Calleros M, G. [CFE, Central Nucleoelectrica de Laguna Verde, Alto Lucero, Veracruz (Mexico)]. E-mail: rcd@nuclear.inin-mx

    2004-07-01

    The use of the analysis of noise for the detection of badly functioning of the components of a BWR it is a powerful tool in the determination of abnormal conditions of operation, during the life of a nuclear plant of power. From the eighties, some nuclear reactors have presented problems related with the jet pumps and the knots of the recirculation. The Regulatory Commission of the United States, in the I E bulletin 80-07, recommended to carry out a periodic supervision of the pressure drop of the jet pumps, to prevent structural failures. In this work, methods of analysis of noise are used for the detection of abnormal conditions of operation of the jet pumps of a BWR. Signals are analysed to low and high frequency of pressure drop with the NOISE software that is in development. The obtained results show the behavior of the jet pumps of jet 6 and 11 before and after a partial blockade in their throats where the pump 6 return to their condition of previous operation and the pump 11 present a new fall of pressure, inside the limit them permissible of operation. The methodology of the analysis of noise demonstrated to be an useful tool for the badly functioning detection, and you could apply to create a database to supervise the dynamic behavior of the jet pumps of an BWR. (Author)

  3. 12 CFR 917.2 - General authorities and duties of Bank boards of directors.

    Science.gov (United States)

    2010-01-01

    ... MANAGEMENT § 917.2 General authorities and duties of Bank boards of directors. (a) Management of a Bank. The...'s board of directors for that Bank's management is non-delegable. (b) Duties of Bank directors. Each Bank director shall have the duty to: (1) Carry out his or her duties as director in good faith, in...

  4. 19 CFR 159.58 - Dumping and countervailing duties; action by port director.

    Science.gov (United States)

    2010-04-01

    ... 19 Customs Duties 2 2010-04-01 2010-04-01 false Dumping and countervailing duties; action by port director. 159.58 Section 159.58 Customs Duties U.S. CUSTOMS AND BORDER PROTECTION, DEPARTMENT OF HOMELAND... Dumping and countervailing duties; action by port director. (a) Antidumping matters. Upon receipt...

  5. 19 CFR 152.2 - Notification to importer of increased duties.

    Science.gov (United States)

    2010-04-01

    ... § 152.2 Notification to importer of increased duties. If the port director believes that the entered... 19 Customs Duties 2 2010-04-01 2010-04-01 false Notification to importer of increased duties. 152.2 Section 152.2 Customs Duties U.S. CUSTOMS AND BORDER PROTECTION, DEPARTMENT OF HOMELAND...

  6. 19 CFR 144.12 - Contents of entry summary; estimated duties.

    Science.gov (United States)

    2010-04-01

    ..., Customs Form 7501, shall show the value, classification, and rate of duty as approved by the port director... 19 Customs Duties 2 2010-04-01 2010-04-01 false Contents of entry summary; estimated duties. 144.12 Section 144.12 Customs Duties U.S. CUSTOMS AND BORDER PROTECTION, DEPARTMENT OF HOMELAND...

  7. 19 CFR 19.36 - Requirements for duty-free store operations.

    Science.gov (United States)

    2010-04-01

    ... reasonable assurance to the port director that conditionally duty-free merchandise purchased therein will be... 19 Customs Duties 1 2010-04-01 2010-04-01 false Requirements for duty-free store operations. 19.36 Section 19.36 Customs Duties U.S. CUSTOMS AND BORDER PROTECTION, DEPARTMENT OF HOMELAND...

  8. 77 FR 4763 - Honey From Argentina: Notice of Initiation of Antidumping Duty New Shipper Review

    Science.gov (United States)

    2012-01-31

    ... International Trade Administration Honey From Argentina: Notice of Initiation of Antidumping Duty New Shipper... ] antidumping duty order on honey from Argentina. See Notice of Antidumping Duty Order: Honey From Argentina, 66...: Background On December 10, 2001, the Department published the antidumping duty order on honey from...

  9. 77 FR 21968 - Honey From Argentina: Rescission of Countervailing Duty Administrative Review

    Science.gov (United States)

    2012-04-12

    ... International Trade Administration Honey From Argentina: Rescission of Countervailing Duty Administrative Review... countervailing duty order on honey ] from Argentina. See Notice of Countervailing Duty Order: Honey From... opportunity to request an administrative review of the countervailing duty order on honey from Argentina...

  10. 76 FR 5332 - Honey From Argentina: Notice of Initiation of Antidumping Duty New Shipper Review

    Science.gov (United States)

    2011-01-31

    ... International Trade Administration Honey From Argentina: Notice of Initiation of Antidumping Duty New Shipper... antidumping duty order on honey from Argentina. See Notice of Antidumping Duty Order: Honey From Argentina, 66...: Background On December 10, 2001, the Department published the antidumping duty order on honey from...

  11. 77 FR 9622 - Proposed Information Collection; Comment Request; Applications for Watch Duty-Exemption and 7113...

    Science.gov (United States)

    2012-02-17

    ...-Exemption and 7113 Jewelry Duty-Refund Program AGENCY: International Trade Administration, Commerce. ACTION... watch duty- exemptions and watch and jewelry duty-refunds to program producers in the U.S. insular... the duty-refund program for the watch and jewelry producers. Form ITA-360P requires no...

  12. 40 CFR Appendix II to Part 1039 - Steady-State Duty Cycles

    Science.gov (United States)

    2010-07-01

    ... following duty cycles apply for variable-speed engines with maximum engine power below 19 kW: (1) The... variable-speed engines with maximum engine power at or above 19 kW: (1) The following duty cycle applies... Appendix II to Part 1039—Steady-State Duty Cycles (a) The following duty cycles apply for constant-speed...

  13. 77 FR 5767 - Certain Tin Mill Products From Japan: Rescission of Antidumping Duty Administrative Review

    Science.gov (United States)

    2012-02-06

    ... International Trade Administration Certain Tin Mill Products From Japan: Rescission of Antidumping Duty...) initiated an administrative review of the antidumping duty order covering certain tin mill products from... antidumping duty order on certain tin mill products from Japan. See Antidumping or Countervailing Duty...

  14. An Ethical Climate is a Duty of Care

    OpenAIRE

    Anona Armstrong; Ronald D. Francis

    2014-01-01

    The current emergence, once again, of corporate collapses due in no small way to unethical behaviour raises questions about the duties and responsibilities of boards of major organisations for building an ethical organisation. This paper argues that the legal duty of care to employees extends to creating an ethical work environment. It describes different types of ethical climates, how they are recognised and the consequences of their impact on the behaviours of their members. It illustrates ...

  15. Aging Management Guideline for commercial nuclear power plants: Electrical switchgear. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Toman, G.; Gazdzinski, R.; Schuler, K. [Ogden Environmental and Energy Services Co., Inc., Blue Bell, PA (United States)

    1993-07-01

    This Aging Management Guideline (AMG) provides recommended methods for effective detection and mitigation of age-related degradation mechanisms in BWR and PWR commercial nuclear power plant electrical switchgear important to license renewal. The latent of this AMG to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR Part 54. This AMG is presented in a manner which allows personnel responsible for performance analysis and maintenance, to compare their plant-specific aging mechanisms (expected or already experienced) and aging management program activities to the more generic results and recommendations presented herein.

  16. Aging management guideline for commercial nuclear power plants-stationary batteries. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Berg, R.; Shao, J.; Krencicki, G.; Giachetti, R. [Multiple Dynamics Corp., Southfield, MI (United States)

    1994-03-01

    The Aging Management Guideline (AMG) describes recommended methods for effective detection and mitigation of age-related degradation mechanisms in BWR and PWR commercial nuclear power plant stationary batteries important to license renewal. The intent of this AMG is to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR Part 54. This AMG is presented in a manner that allows personnel responsible for performance analysis and maintenance to compare their plant-specific aging mechanisms (expected or already experienced) and aging management program activities to the more generic results and recommendations presented herein.

  17. Aging Management Guideline for commercial nuclear power plants: Motor control centers; Final report

    Energy Technology Data Exchange (ETDEWEB)

    Toman, G.; Gazdzinski, R.; O`Hearn, E. [Ogden Environmental and Energy Services Co., Inc., Blue Bell, PA (United States)

    1994-02-01

    This Aging Management Guideline (AMG) provides recommended methods for effective detection and mitigation of age-related degradation mechanisms in Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) commercial nuclear power plant motor control centers important to license renewal. The intent of this AMG is to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR Part 54. This AMG is presented in a manner that allows personnel responsible for performance analysis and maintenance to compare their plant-specific aging mechanisms (expected or already experienced) and aging management program activities to the more generic results and recommendations presented herein.

  18. Wi-Fi Coexistence with Duty Cycled LTE-U

    Directory of Open Access Journals (Sweden)

    Yimin Pang

    2017-01-01

    Full Text Available Coexistence of Wi-Fi and LTE-Unlicensed (LTE-U technologies has drawn significant concern in industry. In this paper, we investigate the Wi-Fi performance in the presence of duty cycle based LTE-U transmission on the same channel. More specifically, one LTE-U cell and one Wi-Fi basic service set (BSS coexist by allowing LTE-U devices to transmit their signals only in predetermined duty cycles. Wi-Fi stations, on the other hand, simply contend the shared channel using the distributed coordination function (DCF protocol without cooperation with the LTE-U system or prior knowledge about the duty cycle period or duty cycle of LTE-U transmission. We define the fairness of the above scheme as the difference between Wi-Fi performance loss ratio (considering a defined reference performance and the LTE-U duty cycle (or function of LTE-U duty cycle. Depending on the interference to noise ratio (INR being above or below −62 dbm, we classify the LTE-U interference as strong or weak and establish mathematical models accordingly. The average throughput and average service time of Wi-Fi are both formulated as functions of Wi-Fi and LTE-U system parameters using probability theory. Lastly, we use the Monte Carlo analysis to demonstrate the fairness of Wi-Fi and LTE-U air time sharing.

  19. Light duty utility arm walkdown report

    Energy Technology Data Exchange (ETDEWEB)

    Smalley, J.L.

    1998-09-25

    This document is a report of the Light Duty Utility Arm (LDUA) drawing walkdown. The purpose of this walkdown was to validate the essential configuration of the LDUA in preparation of deploying the equipment in a Hanford waste tank. The LDUA system has, over the course of its development, caused the generation of a considerable number of design drawings. The number of drawings is estimated to be well over 1,000. A large number consist of vendor type drawings, furnished by both Pacific Northwest National Laboratory (PNNL) and SPAR Aerospace Limited (SPAR). A smaller number, approximately 200, are H-6 type drawing sheets in the Project Hanford Management Contract (PHMC) document control system. A preliminary inspection of the drawings showed that the physical configuration of the LDUA did not match the documented configuration. As a result of these findings, a scoping walkdown of 20 critical drawing sheets was performed to determine if a problem existed in configuration management of the LDUA system. The results of this activity showed that 18 of the 20 drawing sheets were found to contain errors or omissions of varying concern. Given this, Characterization Engineering determined that a walkdown of the drawings necessary and sufficient to enable safe operation and maintenance of the LDUA should be performed. A review team was assembled to perform a review of all of the drawings and determine the set which would need to be verified through an engineering walkdown. The team determined that approximately 150 H-6 type drawing sheets would need to be verified, 12 SPAR/PNNL drawing sheets would need to be verified and converted to H-6 drawings, and three to six new drawings would be created (see Appendix A). This report documents the results of that walkdown.

  20. Standard technical specifications combustion engineering plants: Bases (Sections 2.0--3.3). Volume 2, Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-04-01

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

  1. Effect of a Chloride Transient on the EAC Crack Growth Behaviour of Low-Alloy RPV Steels under Simulated BWR Operating Conditions (CASTOC WP 3, PSI Test 2)

    Energy Technology Data Exchange (ETDEWEB)

    Ritter, S.; Seifert, H.-P

    2002-11-01

    Within the CASTOC-project (5{sup t}h EU FW programme), the environmentally-assisted crack (EAC) growth behaviour of low-alloy reactor pressure vessel (RPV) steels is experimentally investigated under simulated transient and steady-state boiling water reactor (BWR) power operation conditions by six European laboratories. The present report is a summary of the second test of working package (WP) 3 with a NaCl transient, performed at Paul Scherrer Institut (PSI). In the first part of the experiment, an actively growing EAC crack with a crack growth rate (CGR) in the range of the 'low-sulphur SCC line' of the GE-model was generated by periodical partial unloading (PPU) in oxygenated high-temperature, high-purity water (T = 288 {sup o}C, DO = 8 ppm). Then a chloride transient of 49 ppb Cl{sup -} was applied for {approx}40 h. After this transient, the load level was reduced and the loading conditions were changed to pure cyclic loading. Thereupon a second transient with a chloride concentration of 49 ppb was applied. In both RPV steels, the first chloride transient of 49 ppb Cl{sup -} resulted in an acceleration of the EAC crack growth by more than one order of magnitude and in fast, stationary SCC crack growth during the constant load phase of the PPU cycles at K{sub I} values < 60 MPa.m{sup 1/2}. 3 h after adding chloride to the high-purity water, the EAC CGR started to increase in the high-sulphur RPV steel during the constant load phase of a PPU cycle and after 20 h a stationary EAC CGR value in the range of the 'high-sulphur SCC curve' of the GE-model was reached. After 5 h in high-purity water, the crack growth began to slow down after a partial unloading cycle and 15 h later it reached again a stationary CGR value in the range of the 'low-sulphur SCC curve' of the GE-model. The second chloride transient did not result in an acceleration of the crack growth in both investigated specimens. This was explained by crack closure effects

  2. Pulse width control loop as a duty cycle corrector

    Directory of Open Access Journals (Sweden)

    Jovanović Goran

    2004-01-01

    Full Text Available The clock distribution and generation circuitry forms a critical component of current synchronous digital systems. A digital system’s clocks must have not only low jitter, low skew, but also well-controlled duty cycle in order to facilitate versatile clocking techniques. In high-speed CMOS clock buffer design, the duty cycle of a clock is liable to be changed when the clock passes through a multistage buffer because the circuit is not pure digital [8]. In this paper, we propose a pulse width control loop referred as MPWCL (modified pulse width control loop that adopts the same architecture as the conventional PWCL, but with a new pulse generator and new charge pump circuit as a constituent of the duty cycle detector. Thanks to using new building blocks the proposed pulse width control loop can control the duty cycle in a wide range, and what is more important it becomes operative in saturation region too, what provides conditional for fast locking time. For 1.2 µm double-metal double-poly CMOS process with Vdd = 5 V and operating frequency of 133 MHz, results of SPICE simulation show that the duty cycle can be well controlled in the range from 20 % up to 80 % if the loop parameters are properly chosen.

  3. Evolution of high duty cycle echolocation in bats.

    Science.gov (United States)

    Fenton, M Brock; Faure, Paul A; Ratcliffe, John M

    2012-09-01

    Duty cycle describes the relative 'on time' of a periodic signal. In bats, we argue that high duty cycle (HDC) echolocation was selected for and evolved from low duty cycle (LDC) echolocation because increasing call duty cycle enhanced the ability of echolocating bats to detect, lock onto and track fluttering insects. Most echolocators (most bats and all birds and odontocete cetaceans) use LDC echolocation, separating pulse and echo in time to avoid forward masking. They emit short duration, broadband, downward frequency modulated (FM) signals separated by relatively long periods of silence. In contrast, bats using HDC echolocation emit long duration, narrowband calls dominated by a single constant frequency (CF) separated by relatively short periods of silence. HDC bats separate pulse and echo in frequency by exploiting information contained in Doppler-shifted echoes arising from their movements relative to background objects and their prey. HDC echolocators are particularly sensitive to amplitude and frequency glints generated by the wings of fluttering insects. We hypothesize that narrowband/CF calls produced at high duty cycle, and combined with neurobiological specializations for processing Doppler-shifted echoes, were essential to the evolution of HDC echolocation because they allowed bats to detect, lock onto and track fluttering targets. This advantage was especially important in habitats with dense vegetation that produce overlapping, time-smeared echoes (i.e. background acoustic clutter). We make four specific, testable predictions arising from this hypothesis.

  4. Pattern recognition model to estimate intergranular stress corrosion cracking (IGSCC) at crevices and pit sites of 304 SS in BWR environments

    Energy Technology Data Exchange (ETDEWEB)

    Urquidi-Macdonald, Mirna [Penn State University, 212 Earth-Engineering Science Building, University Park, PA 16801 (United States)

    2004-07-01

    Many publications have shown that crack growth rates (CGR) due to intergranular stress corrosion cracking (IGSCC) of metals is dependent on many parameters related to the manufacturing process of the steel and the environment to which the steel is exposed. Those parameters include, but are not restricted to, the concentration of chloride, fluoride, nitrates, and sulfates, pH, fluid velocity, electrochemical potential (ECP), electrolyte conductivity, stress and sensitization applied to the steel during its production and use. It is not well established how combinations of each of these parameters impact the CGR. Many different models and beliefs have been published, resulting in predictions that sometimes disagree with experimental observations. To some extent, the models are the closest to the nature of IGSCC, however, there is not a model that fully describes the entire range of observations, due to the difficulty of the problem. Among the models, the Fracture Environment Model, developed by Macdonald et al., is the most physico-chemical model, accounting for experimental observations in a wide range of environments or ECPs. In this work, we collected experimental data on BWR environments and designed a data mining pattern recognition model to learn from that data. The model was used to generate CGR estimations as a function of ECP on a BWR environment. The results of the predictive model were compared to the Fracture Environment Model predictions. The results from those two models are very close to the experimental observations of the area corresponding to creep and IGSCC controlled by diffusion. At more negative ECPs than the potential corresponding to creep, the pattern recognition predicts an increase of CGR with decreasing ECP, while the Fracture Environment Model predicts the opposite. The results of this comparison confirm that the pattern recognition model covers 3 phenomena: hydrogen embrittlement at very negative ECP, creep at intermediate ECP, and IGSCC

  5. 75 FR 81952 - Greenhouse Gas Emissions Standards and Fuel Efficiency Standards for Medium- and Heavy-Duty...

    Science.gov (United States)

    2010-12-29

    ... Greenhouse Gas Emissions Standards and Fuel Efficiency Standards for Medium- and Heavy-Duty Engines and...--Proposed Vocational Diesel Engine Standards Over the Heavy-Duty FTP Cycle Medium Model year Standard Light... Light heavy-duty Medium heavy-duty Heavy heavy-duty Fuel Consumption Baseline (gallon/1,000 ton-mile...

  6. Application of the FFTBM method and the power relative contribution to the discharge transitory of the recirculation pumps of a BWR; Aplicacion del metodo FFTBM y de la contribucion relativa de potencia al transitorio de disparo de las bombas de recirculacion de un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Castillo D, R.; Ortiz V, J.; Fuentes M, L., E-mail: rogelio.castillo@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    In this work was realized the simulation of the discharge transitory of both recirculation pumps of a BWR with the Simulate-3K code. This type of transitory is used in the stability analyses for the licensing of the fuel reload. An analysis of the precision of the simulation is also presented, using the FFTBM method jointly with the power relative contribution. This way, instead of determining the total precision of the calculation, a weighed precision is obtained by the contribution of each relevant parameter of the transitory. The results show that the precision of the simulation is acceptable due to the small magnitude of the merit figure of the width total average. The error in the merit figure comes mainly from the parameters total flow in the core and temperature of the fuel in the core. (Author)

  7. Medium and Heavy Duty Vehicle Field Evaluations (Presentation)

    Energy Technology Data Exchange (ETDEWEB)

    Walkowicz, K.

    2014-06-01

    This presentation discusses field evaluations of medium- and heavy-duty vehicles performed by NREL. The project provides medium-duty (MD) and heavy-duty (HD) test results, aggregated data, and detailed analysis, including 3rd party unbiased data (data that would not normally be shared by industry in an aggregated and detailed manner). Over 5.6 million miles of advanced technology MD and HD truck data have been collected, documented, and analyzed on over 240 different vehicles since 2002. Data, analysis, and reports are shared within DOE, national laboratory partners, and industry for R&D planning and strategy. The results help guide R&D for new technology development, help define intelligent usage of newly developed technology, and help fleets/users understand all aspects of advanced technology.

  8. Family victim advocates: the importance of critical job duties

    Directory of Open Access Journals (Sweden)

    Teresa H. Young

    Full Text Available Child advocacy centers across the United States intervened in more than 250,000 child abuse cases in 2011(National Children's Alliance, 2012. Understanding the work of family victim advocates is imperative to helping children and families in child abuse cases. In this exploratory study, we surveyed advocates and program directors from child advocacy centers (CACs across the United States to compare their perceptions of the critical job duties of family victim advocates. Data analysis revealed that CAC directors rated the importance of these duties significantly higher than family victim advocates. Results suggest the need for additional training to ensure that family victim advocates understand the importance of critical job duties to meet the needs of children and families in child abuse cases.

  9. Catalog of selected heavy duty transport energy management models

    Science.gov (United States)

    Colello, R. G.; Boghani, A. B.; Gardella, N. C.; Gott, P. G.; Lee, W. D.; Pollak, E. C.; Teagan, W. P.; Thomas, R. G.; Snyder, C. M.; Wilson, R. P., Jr.

    1983-01-01

    A catalog of energy management models for heavy duty transport systems powered by diesel engines is presented. The catalog results from a literature survey, supplemented by telephone interviews and mailed questionnaires to discover the major computer models currently used in the transportation industry in the following categories: heavy duty transport systems, which consist of highway (vehicle simulation), marine (ship simulation), rail (locomotive simulation), and pipeline (pumping station simulation); and heavy duty diesel engines, which involve models that match the intake/exhaust system to the engine, fuel efficiency, emissions, combustion chamber shape, fuel injection system, heat transfer, intake/exhaust system, operating performance, and waste heat utilization devices, i.e., turbocharger, bottoming cycle.

  10. Seismic resistance design of nuclear power plant building structures in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kitano, Takehito [Kansai Electric Power Co., Inc., Osaka (Japan)

    1997-03-01

    Japan is one of the countries where earthquakes occur most frequently in the world and has incurred a lot of disasters in the past. Therefore, the seismic resistance design of a nuclear power plant plays a very important role in Japan. This report describes the general method of seismic resistance design of a nuclear power plant giving examples of PWR and BWR type reactor buildings in Japan. Nuclear facilities are classified into three seismic classes and is designed according to the corresponding seismic class in Japan. Concerning reactor buildings, the short-term allowable stress design is applied for the S1 seismic load and it is confirmed that the structures have a safety margin against the S2 seismic load. (J.P.N.)

  11. Wi-Fi Coexistence with Duty Cycled LTE-U

    OpenAIRE

    Pang, Yimin; Babaei, Alireza; Andreoli-Fang, Jennifer; Hamzeh, Belal

    2016-01-01

    Coexistence of Wi-Fi and LTE-Unlicensed (LTE-U) technologies has drawn significant concern in industry. In this paper, we investigate the Wi-Fi performance in the presence of duty cycle based LTE-U transmission on the same channel. More specifically, one LTE-U cell and one Wi-Fi basic service set (BSS) coexist by allowing LTE-U devices transmit their signals only in predetermined duty cycles. Wi-Fi stations, on the other hand, simply contend the shared channel using the distributed coordinati...

  12. An Ethical Climate is a Duty of Care

    Directory of Open Access Journals (Sweden)

    Anona Armstrong

    2014-09-01

    Full Text Available The current emergence, once again, of corporate collapses due in no small way to unethical behaviour raises questions about the duties and responsibilities of boards of major organisations for building an ethical organisation. This paper argues that the legal duty of care to employees extends to creating an ethical work environment. It describes different types of ethical climates, how they are recognised and the consequences of their impact on the behaviours of their members. It illustrates this with some of the findings from our research into measuring ethics and ethical decision making. In conclusion, it identifies the key factors that boards should address to promote a desirable ethical climate.

  13. Light Duty Utility Arm computer software configuration management plan

    Energy Technology Data Exchange (ETDEWEB)

    Philipp, B.L.

    1998-09-14

    This plan describes the configuration management for the Light Duty Utility Arm robotic manipulation arm control software. It identifies the requirement, associated documents, and the software control methodology. The Light Duty Utility Ann (LDUA) System is a multi-axis robotic manipulator arm and deployment vehicle, used to perform surveillance and characterization operations in support of remediation of defense nuclear wastes currently stored in the Hanford Underground Storage Tanks (USTs) through the available 30.5 cm (12 in.) risers. This plan describes the configuration management of the LDUA software.

  14. Light-duty diesel engine development status and engine needs

    Energy Technology Data Exchange (ETDEWEB)

    1980-08-01

    This report reviews, assesses, and summarizes the research and development status of diesel engine technology applicable to light-duty vehicles. In addition, it identifies specific basic and applied research and development needs in light-duty diesel technology and related health areas where initial or increased participation by the US Government would be desirable. The material presented in this report updates information provided in the first diesel engine status report prepared by the Aerospace Corporation for the Department of Energy in September, 1978.

  15. How Should We Interpret Institutional Duty-Claims?

    OpenAIRE

    Lammer-Heindel, Christoffer S.

    2013-01-01

    It is rather natural to suppose that what we mean when we say that an institutional organization has a moral duty is parallel to whatever it is that we mean when we say that an individual has a duty. I challenge this interpretation on the grounds that it assumes that institutional organizations possess those characteristics or abilities requisite for moral agency—an assumption which I argue is highly suspicious. Against such an interpretation, I argue that we have very good reasons to suppose...

  16. Why participating in (certain) scientific research is a moral duty.

    Science.gov (United States)

    Stjernschantz Forsberg, Joanna; Hansson, Mats G; Eriksson, Stefan

    2014-05-01

    Our starting point in this article is the debate between John Harris and Iain Brassington on whether or not there is a duty to take part in scientific research. We consider the arguments that have been put forward based on fairness and a duty to rescue, and suggest an alternative justification grounded in a hypothetical agreement: that is, because effective healthcare cannot be taken for granted, but requires continuous medical research, and nobody knows what kind of healthcare they will need, participating in research should be viewed from the perspective of a social contract, based on our mutual need for medical advances.

  17. The case for a duty to research: not yet proven.

    Science.gov (United States)

    Brassington, Iain

    2014-05-01

    In this commentary on 'Why participating in (certain) scientific research is a moral duty', I take issue with a number of Stjernschantz Forsberg et al's claims. Though abiding by the terms of a contract might be obligatory, this won't show that those terms themselves indicate a duty--even allowing that there's a contract to begin with. Meanwhile, though we might have reasons to participate, not all reasons are moral reasons, and the paper does not establish that the reasons here are moral in character.

  18. Childhood obesity, parental duties of care and strategies for intervention.

    Science.gov (United States)

    Nolan, Elise Jane

    2012-09-01

    Childhood obesity is an increasingly serious issue which causes significant health problems among children. There are numerous causes of childhood obesity. However, the ultimate responsibility for the problems and costs associated with an obese child should be attributed to that child's parents. Parents owe a duty of care to their child and, when their child is obese, have arguably breached that duty. However, if parents were required to pay their child damages, this would arguably be problematic and of little utility. Rather, intervention strategies should be implemented which seek to treat and prevent childhood obesity and to address the identified causes of childhood obesity.

  19. Improving the action requirements of technical specifications: A risk-comparison of continued operation and plant shutdown

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-04-01

    When the systems needed to remove decay heat are inoperable or degraded, the risk of shutting down the plant may be comparable to, or even higher than, that of continuing power operation with the equipment inoperable while giving priority to repairs. This concern arises because the plant may not have sufficient capability for removing decay heat during the shutdown. However, Technical Specifications (TSs) often require {open_quotes}immediate{close_quotes} shutdown of the plant. In this paper, we present risk-based analyses of the various operational policy alternatives available in such situations, with an example application to the standby service water (SSW) system of a BWR. These analyses can be used to define risk-effective requirements for those standby safety systems under discussion.

  20. Aging Management Guideline for commercial nuclear power plants: Battery chargers, inverters and uninterruptible power supplies. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Berg, R.; Stroinski, M.; Giachetti, R. [Multiple Dynamics Corp., Southfield, MI (United States)

    1994-02-01

    This Aging Management Guideline (AMG) describes recommended methods for effective detection and mitigation of age-related degradation mechanisms in BWR and PWR commercial nuclear power plant battery chargers, inverters and uninterruptible power supplies important to license renewal. The intent of this AMG is to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR Part 54. This AMG is presented in a manner that allows personnel responsible for performance analysis and maintenance to compare their plant-specific aging mechanisms (expected or already, experienced) and aging management program activities to the more generic results and recommendations presented herein.