GEM, Fuel Cycle Cost and Economics for Thermal Reactor, Present Worth Analysis

International Nuclear Information System (INIS)

Hughes, J.A.; Hang, D.F.

1974-01-01

1- Description of problem or function: GEM is used to predict fuel cycle costs for any type nuclear system (i.e., BWR, HTGR, PWR, LMFBR, GCFR,... ). The current version is limited to thermal reactors. GEM is designed for production use by large utilities which have several reactor types on their system. GEM has been written so as to accommodate all major fuel management activities undertaken by a utility - (1) fuel bid analysis, (2) evaluation of actual day to day operation, and (3) system simulation and optimization studies. 2 - Method of solution: Costs are calculated using present-worth techniques and continuous compounding. The equations are based on an investor-owned utility capitalization structure which easily covers the range of industrial, private, and public (government) owned utilities. Three distinct types of analysis (cash flow, allocated costs, yearly cash flow) are performed, each yielding identical results. Using these as a basis many other analyses are undertaken. 3 - Restrictions on the complexity of the problem: Dimensions of all arrays are carried as variables throughout the analysis. The maximum size of each array is set by the user in program MAIN. Current values are set so that maxima are: 50 batches per case study, 20 year batch life, 30 year case study, 120 batch burn time-steps, 20 individual payments (sales) associated with each cost component

LMR design concepts for transuranic management in low sodium void worth cores

International Nuclear Information System (INIS)

Hill, R.N.

1991-01-01

The fuel cycle processing techniques and hard neuron spectrum of the Integral Fast Reactor (IFR) metal fuel cycle have favorable characteristics for the management of transuranics; and the wide range of breeding characteristics available in metal fuelled cores provides for flexibility in transuranic management strategy. Previous studies indicate that most design options which decrease the breeding ratio also show a decrease in sodium void worth; therefore, low void worths are achievable in transuranic burning (low breeding ratio) core designs. This paper describes numerous trade studies assessing various design options for a low void worth transuranic burner core. A flat annular core design appears to be a promising concept; the high leakage geometry yields a low breeding ratio and small sodium void worth. To allow flexibility in breeding characteristics, alternate design options which achieve fissile self-sufficiency are also evaluated. A self-sufficient core design which is interchangeable with the burner core and maintains a low sodium void worth is developed. 13 refs., 1 fig., 4 tabs

LMR design concepts for transuranic management in low sodium void worth cores

International Nuclear Information System (INIS)

Hill, R.N.

1991-01-01

The fuel cycle processing techniques and hard neutron spectrum of the integral Fast Reactor (IFR) metal fuel cycle have favorable characteristics for the management of transuranics; and the wide range of breeding characteristics available in metal fuelled cores provides for flexibility in transuranic management strategy. Previous studies indicate that most design options which decrease the breeding ratio also allow a decrease in sodium void worth; therefore, low void worths are achievable in transuranic burning (low breeding ratio) core designs. This paper describes numerous trade studies assessing various design options for a low void worth transuranic burner core. A flat annular core design appears to be a promising concept; the high leakage geometry yields a low breeding ratio and small sodium void worth. To allow flexibility in breeding characteristics, alternate design options which achieve fissile self-sufficiency are also evaluated. A self-sufficient core design which is interchangeable with the burner core and maintains a low sodium void worth is developed. (author)

Passive safety features of low sodium void worth metal fueled cores in a bottom supported reactor vessel

International Nuclear Information System (INIS)

Chang, Y.I.; Marchaterre, J.F.; Wade, D.C.; Wigeland, R.A.; Kumaoka, Yoshio; Suzuki, Masao; Endo, Hiroshi; Nakagawa, Hiroshi

1991-01-01

A study has been performed on the passive safety features of low-sodium-void-worth metallic-fueled reactors with emphasis on using a bottom-supported reactor vessel design. The reactor core designs included self-sufficient types as well as actinide burners. The analyses covered the reactor response to the unprotected, i.e. unscrammed, transient overpower accident and the loss-of-flow accident. Results are given demonstrating the safety margins that were attained. 4 refs., 4 figs., 2 tabs

Intercomparison of rod-worth measurement techniques in a LEU-HTR assembly

International Nuclear Information System (INIS)

Williams, T.; Chawla, R.

1994-01-01

The measurement of absorber-rod worths in the radial reflector of a LEU-HTR pebble bed system is described. Particular emphasis is placed on the choice of complementary measurement techniques to ensure that sensitivities to systematic errors in the calculated parameters used in the analysis are minimised. (author) 3 figs., 3 tabs., 8 refs

Benchmark physics experiment of metallic-fueled LMFBR at FCA. 2

International Nuclear Information System (INIS)

Iijima, Susumu; Oigawa, Hiroyuki; Ohno, Akio; Sakurai, Takeshi; Nemoto, Tatsuo; Osugi, Toshitaka; Satoh, Kunio; Hayasaka, Katsuhisa; Bando, Masaru.

1993-10-01

An availability of data and method for a design of metallic-fueled LMFBR is examined by using the experiment results of FCA assembly XVI-1. Experiment included criticality and reactivity coefficients such as Doppler, sodium void, fuel shifting and fuel expansion. Reaction rate ratios, sample worth and control rod worth were also measured. Analysis was made by using three-dimensional diffusion calculations and JENDL-2 cross sections. Predictions of assembly XVI-1 reactor physics parameters agree reasonably well with the measured values, but for some reactivity coefficients such as Doppler, large zone sodium void and fuel shifting further improvement of calculation method was need. (author)

Critical experiment and analysis for nitride fuel fast reactor using FCA

International Nuclear Information System (INIS)

Andoh, Masaki; Iijima, Susumu; Okajima, Shigeaki; Sakurai, Takeshi; Oigawa, Hiroyuki

2000-03-01

As a research on FBR with new types of fuel, a series of experiments on a nitride fuel fast reactor was carried out at Fast Critical Assembly (FCA) to evaluate the calculation accuracy on the neutronic characteristics of the reactor. In this study, criticality, sample reactivity worth and sodium void reactivity worth were measured in the FCA XIX-2 core simulating a nitride fuel fast reactor and were analyzed using the standard analysis method for FCA fast reactor cores. The accuracy of the analysis on the effective multiplication factor was the same as those of the other FCA cores. For the plate sample reactivity worth, the calculation on the radial distribution of plutonium plate reactivity worth overestimated the measurement depending on the distance from the center of the core. For the sodium void reactivity worth, the calculation overestimated the experimental value 10 to 20% at the core center, while the overestimation was improved as the voided position was located at the core boundary. It was found that the transport effect was considerable even at the center of the core. It was considered that the calculation accuracy on the non-leakage term of the void reactivity worth and transport correction should be improved. (author)

Microcomputer-based equipment-control and data-acquisition system for fission-reactor reactivity-worth measurements

International Nuclear Information System (INIS)

McDowell, W.P.; Bucher, R.G.

1980-01-01

Material reactivity-worth measurements are one of the major classes of experiments conducted on the Zero Power research reactors (ZPR) at Argonne National Laboratory. These measurements require the monitoring of the position of a servo control element as a sample material is positioned at various locations in a critical reactor configuration. In order to guarantee operational reliability and increase experimental flexibility for these measurements, the obsolete hardware-based control unit has been replaced with a microcomputer based equipment control and data acquisition system. This system is based on an S-100 bus, dual floppy disk computer with custom built cards to interface with the experimental system. To measure reactivity worths, the system accurately positions samples in the reactor core and acquires data on the position of the servo control element. The data are then analyzed to determine statistical adequacy. The paper covers both the hardware and software aspects of the design

Microcomputer-based equipment-control and data-acquisition system for fission-reactor reactivity-worth measurements

Energy Technology Data Exchange (ETDEWEB)

McDowell, W.P.; Bucher, R.G.

1980-01-01

Material reactivity-worth measurements are one of the major classes of experiments conducted on the Zero Power research reactors (ZPR) at Argonne National Laboratory. These measurements require the monitoring of the position of a servo control element as a sample material is positioned at various locations in a critical reactor configuration. In order to guarantee operational reliability and increase experimental flexibility for these measurements, the obsolete hardware-based control unit has been replaced with a microcomputer based equipment control and data acquisition system. This system is based on an S-100 bus, dual floppy disk computer with custom built cards to interface with the experimental system. To measure reactivity worths, the system accurately positions samples in the reactor core and acquires data on the position of the servo control element. The data are then analyzed to determine statistical adequacy. The paper covers both the hardware and software aspects of the design.

Calculational benchmark comparisons for a low sodium void worth actinide burner core design

International Nuclear Information System (INIS)

Hill, R.N.; Kawashima, M.; Arie, K.; Suzuki, M.

1992-01-01

Recently, a number of low void worth core designs with non-conventional core geometries have been proposed. Since these designs lack a good experimental and computational database, benchmark calculations are useful for the identification of possible biases in performance characteristics predictions. In this paper, a simplified benchmark model of a metal fueled, low void worth actinide burner design is detailed; and two independent neutronic performance evaluations are compared. Calculated performance characteristics are evaluated for three spatially uniform compositions (fresh uranium/plutonium, batch-averaged uranium/transuranic, and batch-averaged uranium/transuranic with fission products) and a regional depleted distribution obtained from a benchmark depletion calculation. For each core composition, the flooded and voided multiplication factor, power peaking factor, sodium void worth (and its components), flooded Doppler coefficient and control rod worth predictions are compared. In addition, the burnup swing, average discharge burnup, peak linear power, and fresh fuel enrichment are calculated for the depletion case. In general, remarkably good agreement is observed between the evaluations. The most significant difference is predicted performance characteristics is a 0.3--0.5% Δk/(kk) bias in the sodium void worth. Significant differences in the transmutation rate of higher actinides are also observed; however, these differences do not cause discrepancies in the performing predictions

Local heterogeneity effects on small-sample worths

International Nuclear Information System (INIS)

Schaefer, R.W.

1986-01-01

One of the parameters usually measured in a fast reactor critical assembly is the reactivity associated with inserting a small sample of a material into the core (sample worth). Local heterogeneities introduced by the worth measurement techniques can have a significant effect on the sample worth. Unfortunately, the capability is lacking to model some of the heterogeneity effects associated with the experimental technique traditionally used at ANL (the radial tube technique). It has been suggested that these effects could account for a large portion of what remains of the longstanding central worth discrepancy. The purpose of this paper is to describe a large body of experimental data - most of which has never been reported - that shows the effect of radial tube-related local heterogeneities

Core concepts for ''zero-sodium-void-worth core'' in metal fuelled fast reactor

International Nuclear Information System (INIS)

Chang, Y.I.; Hill, R.N.; Fujita, E.K.; Wade, D.C.; Kumaoka, Y.; Suzuki, M.; Kawashima, M.; Nakagawa, H.

1991-01-01

Core design options to reduce the sodium void worth in metal fueled LMRs are investigated. Two core designs which achieve a zero sodium void worth are analyzed in detail. The first design is a ''pancaked'' and annular core with enhanced transuranic burning capabilities; the high leakage in this design yields a low breeding ratio and small void worth. The second design is an axially multilayered annular core which is fissile self-sufficient; in this design, the upper and lower core regions are neutronically decoupled for reduced void worth while fissile self-sufficiency is achieved using internal axial blankets plus external radial and axial blanket zones. The neutronic performance characteristics of these low void worth designs are assessed here; their passive safety properties are discussed in a companion paper. 16 refs., 2 figs., 3 tabs

Dynamic rod worth measurements (''Rod Insertion''). Final report for the period 01 December 1994 - 30 November 1996

International Nuclear Information System (INIS)

Bogdan, G.

1996-12-01

Reload startup physics tests are performed for pressurized water reactors (PWR power plant) following a refuelling or other significant core alteration for which nuclear design calculations are required. Part of the reload startup physics tests are control rod group worths measurements. for this purpose a new so-called method ''Rod-Insertion'' was developed. It can also be used as an additional measuring instrument on the research reactor for education purposes. The principle of the rod-insertion method is to start from a critical reactor operating at low power and to measure the time-dependent reactivity change while a control rod is inserted into the core. Unlike in the rod-drop method, the measured control rod is inserted with the drive mechanism at normal speed. By analyzing the flux trace using point-kinetics, not only the total rod worth but also the differential and the integral rod worth curves are obtained. A high-quality electrometer is required for monitoring the neutron flux. The analysis is performed by transferring the data to an IBM PC compatible with some additional standard electronic board and the associated software. The new reactivity meter has been validated on the TRIGA Mark II reactors in Ljubljana and Vienna and at the Krsko Nuclear Power Plant during physics startup tests after reload. The results proved the high performance of the reactivity meter in the standard applications according to the existing procedures, as well as in the new rod-insertion technique of measuring the control rod group worths. This method drastically differs from others such as absence of any chemical control of reactivity (like boron exchange method), and minimizing a testing time and waste coolant production

Estimation of irradiated control rod worth

International Nuclear Information System (INIS)

Varvayanni, M.; Catsaros, N.; Antonopoulos-Domis, M.

2009-01-01

When depleted control rods are planned to be used in new core configurations, their worth has to be accurately predicted in order to deduce key design and safety parameters such as the available shutdown margin. In this work a methodology is suggested for the derivation of the distributed absorbing capacity of a depleted rod, useful in the case that the level of detail that is known about the irradiation history of the control rod does not allow an accurate calculation of the absorber's burnup. The suggested methodology is based on measurements of the rod's worth carried out in the former core configuration and on corresponding calculations based on the original (before first irradiation) absorber concentration. The methodology is formulated for the general case of the multi-group theory; it is successfully tested for the one-group approximation, for a depleted control rod of the Greek Research Reactor, containing five neutron absorbers. The computations reproduce satisfactorily the irradiated rod worth measurements, practically eliminating the discrepancy of the total rod worth, compared to the computations based on the nominal absorber densities.

Application of the Modified Source Multiplication (MSM) technique to subcritical reactivity worth measurements in thermal and fast reactor systems

International Nuclear Information System (INIS)

Blaise, P.; Fougeras, P.; Mellier, F.

2009-01-01

The Amplified Source Multiplication (ASM) method and its improved Modified Source Multiplication (MSM) method have been widely used in the CEA's EOLE and MASURCA critical facilities over the past decades for the determination of reactivity worths by using fission chambers in subcritical configurations. They have been successfully applied to absorber (single or clusters) worth measurement in both thermal and fast spectra, or for (sodium or water) void reactivity worths. The ASM methodology, which is the basic technique to estimate a reactivity worth, uses relatively simple relationships between count rates of efficient miniature fission chambers located in slightly subcritical reference and perturbed configurations. If this method works quite well for small reactivity variation (a few effective delayed neutron fraction), its raw results needs to be corrected to take into account the flux perturbation in the fission chamber. This is performed by applying to the measurement a correction factor called MSM. Its characteristics is to take into account the local space and energy variation of the spectrum in the fission chamber, through standard perturbation theory applied to neutron transport calculation in the perturbed configuration. The proposed paper describes in details both methodologies, with their associated uncertainties. Applications on absorber cluster worth in the MISTRAL-4 full MOX mock-up core and the last core loaded in MASURCA show the importance of the MSM correction on raw data. (authors)

Comparison of the worth of control and protection system rods of different design on the basis of the measurements in BN-600 reactor

International Nuclear Information System (INIS)

Vasilyev, B.A.; Roslyakov, V.F.; Farakshin, M.R.

1988-01-01

The results of the worth measurements of the basic and experimental absorbing rods of BN-600 reactor are presented. The procedure used for the rods worth comparison on the basis of calculated and experimental data interpretation is described here. Basic and experimental rods relative worth is also presented. (author). 5 refs, 3 figs, 2 tabs

Analysis of control rod worth in experimental fast reactor JOYO

International Nuclear Information System (INIS)

Arii, Y.; Aoyama, T.; Okimoto, Y.; Yoshida, A.; Mizoo, N.

1988-01-01

In JOYO, the measurement of control rod worths have been carried out in the beginning of the each cycle, using both period method and neutron source multiplication method. In this paper, the calculational method of control rod worths in the design stage and the comparison with the design values and measured ones are shown. The reasons that the control rod worths change slightly in each cycle, are also investigated. (author). 13 figs, 12 tabs

Conversion and start up of Tehran Research Reactor with LEU fuel

International Nuclear Information System (INIS)

Zaker, M.

2004-01-01

The MW Tehran Research Reactor, Highly Enriched Uranium (HEU) fuel has been converted to Low Enriched Uranium (LEU) fuel using U 3 0 8 -Al with less than 20% enriched uranium. Measured value of excess reactivity, control rod worth and other parameters indicate good agreement with computational predictions. (author)

Quantifying data worth toward reducing predictive uncertainty

Science.gov (United States)

Dausman, A.M.; Doherty, J.; Langevin, C.D.; Sukop, M.C.

2010-01-01

The present study demonstrates a methodology for optimization of environmental data acquisition. Based on the premise that the worth of data increases in proportion to its ability to reduce the uncertainty of key model predictions, the methodology can be used to compare the worth of different data types, gathered at different locations within study areas of arbitrary complexity. The method is applied to a hypothetical nonlinear, variable density numerical model of salt and heat transport. The relative utilities of temperature and concentration measurements at different locations within the model domain are assessed in terms of their ability to reduce the uncertainty associated with predictions of movement of the salt water interface in response to a decrease in fresh water recharge. In order to test the sensitivity of the method to nonlinear model behavior, analyses were repeated for multiple realizations of system properties. Rankings of observation worth were similar for all realizations, indicating robust performance of the methodology when employed in conjunction with a highly nonlinear model. The analysis showed that while concentration and temperature measurements can both aid in the prediction of interface movement, concentration measurements, especially when taken in proximity to the interface at locations where the interface is expected to move, are of greater worth than temperature measurements. Nevertheless, it was also demonstrated that pairs of temperature measurements, taken in strategic locations with respect to the interface, can also lead to more precise predictions of interface movement. Journal compilation ?? 2010 National Ground Water Association.

Reliability worth assessment of radial systems with distributed generation

OpenAIRE

Bellart Llavall, Francesc Xavier

2010-01-01

With recent advances in technology, utilities generation (DG) on the distribution systems. Reliability worth is very important in power system planning and operation. Having a DG ensures reli increase the reliability worth. This research project presents the study of a radial distribution system and the impact of placing DG in order to increase the reliability worth. where a DG have to be placed. The reliability improvement is measured by different reliability indices tha...

Analysis of sodium-void-worths in ZPPR-3 modified phase 3 core

Energy Technology Data Exchange (ETDEWEB)

Takeda, T.; Arai, K.; Otake, I. [Osaka Univ. (JP)

1980-09-15

The sodium-void-worths in the ZPPR-3 modified phase 3 core, in which singularities such as control-rods and sodium-followers were voided, have been analyzed using a unified diffusion coefficient. The unified diffusion coefficient is obtained by applying the Benoist formula to a super-cell consisting of different drawers, and is applicable not only to fuel drawers but also to control-rod drawers or sodium-followers. Using the coefficient the interference effect of neutron streaming between different drawers can be taken into account. The applicability of the unified diffusion coefficient to sodium-void-worth calculations has been checked in a slab model and a RZ model. The sodium-void-worths in the ZPPR-3 modified phase 3 core have been analyzed by carrying out 16-group three-dimensional diffusion calculations using the unified diffusion coefficient and the results have been compared with experimental data. The comparison indicates that the unified diffusion coefficient is useful in calculating the sodium-void-worth in a region including sodium-voided singularities.

Division I men and women athletes do not differ on perceptions of worth.

Science.gov (United States)

Lockhart, Barbara D; Black, Nate; Vincent, William J

2012-04-01

Historically, especially prior to the increased interest in women's athletics with the passage of Title IX in 1972, there have been negative perceptions of women as athletes. If these social perceptions still hold in part today, as is indirectly suggested by unequal press coverage and less basic support for women athletes, one might predict that collegiate female athletes would rate themselves lower on self-esteem and worth than collegiate male athletes. 176 Division I male (n = 90) and female (n = 86) athletes rated their perceptions of self on the Worth Index which measures basic human worth, personal security, performance, and physical self; these are divided into intrinsic (unconditional worth) measures and behavior or performance (conditional worth) measures. There were no significant sex differences in the ratings of any aspect of perceived worth, in contrast to prior results among non-athletes. In spite of less support given to women athletes, perhaps the long-term high-intensity competition that is required to reach Division I status tends to eliminate sex differences in self-worth.

Reactivity worth measurements on the CALIBAN reactor: interpretation of integral experiments for the nuclear data validation

International Nuclear Information System (INIS)

Richard, B.

2012-01-01

The good knowledge of nuclear data, input parameters for the neutron transport calculation codes, is necessary to support the advances of the nuclear industry. The purpose of this work is to bring pertinent information regarding the nuclear data integral validation process. Reactivity worth measurements have been performed on the Caliban reactor, they concern four materials of interest for the nuclear industry: gold, lutetium, plutonium and uranium 238. Experiments which have been conducted in order to improve the characterization of the core are also described and discussed, the latter are necessary to the good interpretation of reactivity worth measurements. The experimental procedures are described with their associated uncertainties, measurements are then compared to numerical results. The methods used in numerical calculations are reported, especially the multigroup cross sections generation for deterministic codes. The modeling of the experiments is presented along with the associated uncertainties. This comparison led to an interpretation concerning the qualification of nuclear data libraries. Discrepancies are reported, discussed and justify the need of such experiments. (author) [fr

Analysis of Doppler effect measurement in FCA cores using JENDL-3.2 library

International Nuclear Information System (INIS)

Okajima, Shigeaki

1996-01-01

For the evaluation of the calculation accuracy of the 238 U Doppler effect using JENDL-3.2 library, the previously measured Doppler reactivity worths in the FCA were systematically analyzed. In the analysis the Doppler reactivity worth was calculated by a first order perturbation theory. The calculated results were compared with those using JENDL-3.1 library. The JENDL-3.2 calculation in MOX fuel mock-up cores agrees well with the experimental values within the experimental error. In U-235/Pu fuel cores, the JENDL-3.2 calculation gives 12-15% larger Doppler reactivity worths than the JENDL-3.1 calculation. (author)

Feasibility of reactivity worth measurements by perturbation method with Caliban and Silene experimental reactors

Energy Technology Data Exchange (ETDEWEB)

Casoli, Pierre; Authier, Nicolas [Commissariat a l' Energie Atomique, Centre d' Etudes de Valduc, 21120 Is-Sur-Tille (France)

2008-07-01

Reactivity worth measurements of material samples put in the central cavities of nuclear reactors allow to test cross section nuclear databases or to extract information about the critical masses of fissile elements. Such experiments have already been completed on the Caliban and Silene experimental reactors operated by the Criticality and Neutronics Research Laboratory of Valduc (CEA, France) using the perturbation measurement technique. Calculations have been performed to prepare future experiments on new materials, such as light elements, structure materials, fission products or actinides. (authors)

The spent fuel safety experiment

International Nuclear Information System (INIS)

Harmms, G.A.; Davis, F.J.; Ford, J.T.

1995-01-01

The Department of Energy is conducting an ongoing investigation of the consequences of taking fuel burnup into account in the design of spent fuel transportation packages. A series of experiments, collectively called the Spent Fuel Safety Experiment (SFSX), has been devised to provide integral benchmarks for testing computer-generated predictions of spent fuel behavior. A set of experiments is planned in which sections of unirradiated fuel rods are interchanged with similar sections of spent PWR fuel rods in a critical assembly. By determining the critical size of the arrays, one can obtain benchmark data for comparison with criticality safety calculations. The integral reactivity worth of the spent fuel can be assessed by comparing the measured delayed critical fuel loading with and without spent fuel. An analytical effort to model the experiments and anticipate the core loadings required to yield the delayed critical conditions runs in parallel with the experimental effort

Verification study of thorium cross section in MVP calculation of thorium based fuel core using experimental data

International Nuclear Information System (INIS)

Mai, V. T.; Fujii, T.; Wada, K.; Kitada, T.; Takaki, N.; Yamaguchi, A.; Watanabe, H.; Unesaki, H.

2012-01-01

Considering the importance of thorium data and concerning about the accuracy of Th-232 cross section library, a series of experiments of thorium critical core carried out at KUCA facility of Kyoto Univ. Research Reactor Inst. have been analyzed. The core was composed of pure thorium plates and 93% enriched uranium plates, solid polyethylene moderator with hydro to U-235 ratio of 140 and Th-232 to U-235 ratio of 15.2. Calculations of the effective multiplication factor, control rod worth, reactivity worth of Th plates have been conducted by MVP code using JENDL-4.0 library [1]. At the experiment site, after achieving the critical state with 51 fuel rods inserted inside the reactor, the measurements of the reactivity worth of control rod and thorium sample are carried out. By comparing with the experimental data, the calculation overestimates the effective multiplication factor about 0.90%. Reactivity worth of the control rods evaluation using MVP is acceptable with the maximum discrepancy about the statistical error of the measured data. The calculated results agree to the measurement ones within the difference range of 3.1% for the reactivity worth of one Th plate. From this investigation, further experiments and research on Th-232 cross section library need to be conducted to provide more reliable data for thorium based fuel core design and safety calculation. (authors)

Spectral fine structure effects on material and doppler reactivity worth

International Nuclear Information System (INIS)

Greenspan, E.; Karni, Y.

1975-01-01

New formulations concerning the fine structure effects on the reactivity worth of resonances are developed and conclusions are derived following the extension to more general types of perturbations which include: the removal of resonance material at finite temperatures and the temperature variation of part of the resonance material. It is concluded that the flux method can overpredict the reactivity worth of resonance materials more than anticipated. Calculations on the Doppler worth were carried out; the results can be useful for asessing the contribution of the fine structure effects to the large discrepancy that exists between the calculated and measured small sample Doppler worths. (B.G.)

Optical fuel spray measurements

Energy Technology Data Exchange (ETDEWEB)

Hillamo, H.

2011-07-01

Diesel fuel sprays, including fuel/air mixing and the physics of two-phase jet formation, are discussed in the thesis. The fuel/air mixing strongly affects emissions formation in spray combustion processes where the local combustion conditions dictate the emission formation. This study comprises optical measurements both in pressurized spray test rigs and in a running engine.The studied fuel injection was arranged with a common rail injection system and the injectors were operated with a solenoid-based injection valve. Both marine and heavy-duty diesel engine injectors were used in the study. Optical fuel spray measurements were carried out with a laser-based double-framing camera system. This kind of equipments is usually used for flow field measurements with Particle Image Velocimetry technique (PIV) as well as for backlight imaging. Fundamental fuel spray properties and spray formation were studied in spray test rigs. These measurements involved studies of mixing, atomization, and the flow field. Test rig measurements were used to study the effect of individual injection parameters and component designs. Measurements of the fuel spray flow field, spray penetration, spray tip velocity, spray angle, spray structure, droplet accumulation, and droplet size estimates are shown. Measurement campaign in a running optically accessible large-bore medium-speed engine was also carried out. The results from engine tests were compared with equivalent test rig measurements, as well as computational results, to evaluate the level of understanding of sprays. It was shown that transient spray has an acceleration and a deceleration phase. Successive flow field measurements (PIV) in optically dense diesel spray resulted in local and average velocity data of diesel sprays. Processing fuel spray generates a flow field to surrounding gas and entrainment of surrounding gas into fuel jet was also seen at the sides of the spray. Laser sheet imaging revealed the inner structure of diesel

Measurement and analysis of reactivity worth of 237Np sample in cores of TCA and FCA

International Nuclear Information System (INIS)

Sakurai, Takeshi; Mori, Takamasa; Okajima, Shigeaki; Tani, Kazuhiro; Suzaki, Takenori; Saito, Masaki

2009-01-01

The reactivity worth of 22.87 grams of 237 Np oxide sample was measured and analyzed in seven uranium cores in the Tank-Type Critical Assembly (TCA) and two uranium cores in the Fast Critical Assembly (FCA) at the Japan Atomic Energy Agency. The TCA cores provided a systematic variation in the neutron spectrum between the thermal and resonance energy regions. The FCA cores, XXI and XXV, provided a hard neutron spectrum of the fast reactor and a soft one of the resonance energy region, respectively. Analyses were carried out using the JENDL-3.3 nuclear data library with a Monte Carlo method for the TCA cores and a deterministic method for the FCA cores. The ratios of calculated to experimental (C/E) reactivity worth were between 0.97 and 0.91, and showed no apparent dependence on the neutron spectrum. (author)

Collegiate misuse of prescription stimulants: examining differences in self-worth.

Science.gov (United States)

Giordano, Amanda L; Prosek, Elizabeth A; Reader, Emily A; Bevly, Cynthia M; Turner, Kori D; LeBlanc, Yvette N; Vera, Ryan A; Molina, Citlali E; Garber, Sage Ann

2015-02-01

Prescription stimulant medication is commonly used to treat attention-deficit hyperactivity disorder (ADHD). However, stimulant medication misuse is a prevalent problem among the college population. There is limited research on psychological factors associated with collegiate nonmedical stimulant misuse. To examine the association between college students' self-worth and stimulant medication misuse. A quantitative study implemented during the 2013-2014 academic year in which we utilized a convenience sample of undergraduate students at a public university. College students (N = 3,038) completed an electronic survey packet including a stimulant use index and the Contingencies of Self-Worth Scale. We conducted descriptive discriminant analysis (DDA) to measure the associations between four groups: Nonusers, Appropriate Users, Nonprescribed Misusuers, and Prescribed Users. Significant differences in contingencies of self-worth existed between the four groups of students. Specifically, external contingencies of self-worth, such as appearance and approval, were associated with stimulant medication misuse, whereas, internal contingencies of self-worth, such as God's love and virtue, were associated with nonuse and appropriate prescribed use. Conclusions/Importance: The findings of the current study suggested contingencies of self-worth partially explain prescription stimulant misuse among the collegiate population. Addressing self-worth may be helpful in the treatment of stimulant misuse with college students.

Puerto Rican kindergartners' self-worth as coded from the Attachment Story Completion Task: correlated with other self-evaluation measures and ratings of child behavior toward mothers and peers.

Science.gov (United States)

Gullón-Rivera, Ángel L

2013-01-01

This multi-method multi-informant study assessed 105 Puerto Rican kindergartners' sense of self-worth in family relationships as coded from their responses to the Attachment Story Completion Task (ASCT). The ASCT scores were compared with responses to two other age-appropriate self-evaluation measures (the Cassidy Puppet Interview and the Pictorial Scales of Social Acceptance). Correlations of children's scores on the three self-measures with maternal ratings of the mother-child relationship and teacher ratings of the child's prosocial behavior with peers were then compared. ASCT self-worth and Puppet Interview scores were strongly correlated with each other and both were modestly related to the pictorial social acceptance scales. All three measures were significantly associated with maternal and teacher reports of child behavior, but the strongest correlations were obtained with the ASCT. Coding the ASCT in terms of self-worth appears to be a promising approach for evaluating young children's (vicariously expressed) self-worth in family relationships.

Self-worth, perceived competence, and behaviour problems in children with cerebral palsy.

Science.gov (United States)

Schuengel, Carlo; Voorman, Jeanine; Stolk, Joop; Dallmeijer, Annet; Vermeer, Adri; Becher, Jules

2006-10-30

To examine the relevance of physical disabilities for self-worth and perceived competence in children with cerebral palsy (CP), and to examine associations between behaviour problems and self-worth and perceived competence. The Harter scales for self-worth and perceived competence and a new scale for perceived motor competence were used in a sample of 80 children with CP. Their motor functioning was assessed with the Gross Motor Functioning Measure (GMFM) and behaviour problems with the Child Behaviour Check List administered to parents. Self-worth and perceived competence for children with CP were comparable to the Dutch norm sample, except for perceived athletic competence. Within the CP sample, the GMFM showed a domain-specific effect on perceived motor competence. In the multivariate analysis, internalizing problems were associated negatively with all perceived competence scales and self-worth, whereas aggression was positively associated with perceived motor competence, physical appearance, and self-worth. Children with CP appear resilient against challenges posed to their self-worth caused by their disabilities. The relevance of the physical disability appears to be domain-specific. For internalizing problems and aggression, different theoretical models are needed to account for their associations with self-worth and perceived competence.

Analytical analyses of startup measurements associated with the first use of LEU fuel in Romania's 14-MW TRIGA reactor

International Nuclear Information System (INIS)

Bretscher, M.M.; Snelgrove, J.L.; Ciocanescu, M.

1992-01-01

The 14-MW TRIGA steady state reactor (SSR) is located in Pitesti, Romania. Beginning with an HEU core (10 wt% U), the reactor first went critical in November 1979 but was shut down ten years later because of insufficient excess reactivity. Last November the Institute for Nuclear Research (INR), which operates the SSR, received from the ANL RERTR program a shipment of 125 LEU pins fabricated by General Atomics and of the same geometry as the original fuel but with an enrichment of 19.7% 235U and a loading of 45 wt% U. Using 100 of these pins, four LEU clusters, each containing a 5 x 5 square array of fuel rods, were assembled. These four LEU clusters replaced the four most highly burned HEU elements in the SSR. The reactor resumed operations last February with a 35-element mixed HEU/LEU core configuration. In preparation for full power operation of the SSR with this mixed HEU/LEU core, a number of measurements were made. These included control rod calibrations, excess reactivity determinations, worths of experiment facilities, reaction rate distributions, and themocouple measurements of fuel temperatures as a function of reactor power. This paper deals with a comparison of some of these measured reactor parameters with corresponding analytical calculations

Analysis of subcritical control rod worth measurements in assembly BZB/3

International Nuclear Information System (INIS)

Giese, H.

1981-07-01

A series of subcritical absorber array measurements was performed in version three of the BIZET assembly BZB in order to check the ability of standard reactor computational codes used by the BIZET participants in predicting control rod worths in large fast reactors. Assembly BZB/3 was a two-zone core with a diameter of about 2.5 m and a core height of 0.89 m, fuelled with plutonium. Fifteen control rod positions and twelve secondary shutdown rod positions were simulated in the core. The measurements comprised the insertion of single absorbers as well as various groups of absorbers and were based on the modified source multiplication method. The KfK analysis was confined to the calculation of eigenvalues for different absorber arrays, also with a view to a comparison with the results of a former BZA evaluation with calculation-to-experiment values of up to C/E ∼ 1.10. The C/E-values found for BZB/3 ranged from 1.02 to 1.10 and did not show a systematic variation at different radial positions or different degrees of absorber asymmetry

Self-worth, perceived competence, and behaviour problems in children with cerebral palsy

OpenAIRE

Schuengel, C.; Voorman, J.; Stolk, J.; Dallmeijer, A.J.; Vermeer, A; Becher, J.

2006-01-01

Purpose. To examine the relevance of physical disabilities for self-worth and perceived competence in children with cerebral palsy (CP), and to examine associations between behaviour problems and self-worth and perceived competence. Methods. The Harter scales for self-worth and perceived competence and a new scale for perceived motor competence were used in a sample of 80 children with CP. Their motor functioning was assessed with the Gross Motor Functioning Measure (GMFM) and behaviour probl...

The development of the measurement technique of the control rod worth with the inverse kinetics method considering the influence of the steady neutron source

International Nuclear Information System (INIS)

Takeuchi, Mitsuo; Wada, Shigeru; Takahashi, Hiroyuki; Hayashi, Kazuhiko; Murayama, Yoji

2000-09-01

At the research reactor such as JRR-3M, the operation management is carried out in order to ensure safe operation, for example, the excess reactivity is measured regularly and confirmed that it satisfies a safety condition. The excess reactivity is calculated using control rod position in criticality and control rod worth measured by a positive period method (P.P method), the conventional inverse kinetic method (IK method) and so on. The neutron source, however, influences measurement results and brings in a measurement error. A new IK method considering the influence of the steady neutron sources is proposed and applied to the JRR-3M. This report shows that the proposed IK method measures control rod worth more precisely than a conventional IK method. (author)

Description of reactor fuel breeding with three integral concepts

International Nuclear Information System (INIS)

Ott, K.O.; Hanan, N.A.; Maudlin, P.J.; Borg, R.C.

1979-01-01

The time-dependent breeding of fuel in a growing system of breeder reactors can be characterized by the transitory (instantaneous) growth rate, γ(t). The three most important aspects of γ(t) can be expressed by time-independent integral concepts. Two of these concepts are in widespread use. A third integral concept that links the two earlier ones is introduced. The time-dependent growth rate has an asymptotic value, γ/sup infinity/, the equilibrium growth rate, which is the basis for the calculation of the doubling time. The equilibrium growth rate measures the breeding capability and represents a reactor property. Maximum deviation of γ(t) and γ/sup infinity/ generally appears at the initial startup of the reactor, where γ(t = 0) = γ 0 . This deviation is due to the difference between the initial and asymptotic fuel inventory composition. The initial growth rate can be considered a second integral concept; it characterizes the breeding of a particular fuel in a given reactor. Growth rates are logarithmic derivatives of the growing mass of fuel in breeder reactors, especially γ/sup infinity/, which describes the asymptotic growth by exp(γ/sup infinity/t). There is, however, a variation in the fuel-mass factor in front of this exponential function during the transition from γ 0 to γ/sup infinity/. It is shown that this variation of the fuel mass during transitioncan be described by a third integral concept, termed the breeding bonus, b. The breeding bonus measures the quality of a fuel for its use in a given reactor in terms of its impact on the magnitude of the asymptotically growing fuel mass. The calculation of γ 0 and γ/sup infinity/ is facilitated by use of the critical mass (CM) worths and the breeding worth factors, respectively

Polynomial curve fitting for control rod worth using least square numerical analysis

International Nuclear Information System (INIS)

Muhammad Husamuddin Abdul Khalil; Mark Dennis Usang; Julia Abdul Karim; Mohd Amin Sharifuldin Salleh

2012-01-01

RTP must have sufficient excess reactivity to compensate the negative reactivity feedback effects such as those caused by the fuel temperature and power defects of reactivity, fuel burn-up and to allow full power operation for predetermined period of time. To compensate this excess reactivity, it is necessary to introduce an amount of negative reactivity by adjusting or controlling the control rods at will. Control rod worth depends largely upon the value of the neutron flux at the location of the rod and reflected by a polynomial curve. Purpose of this paper is to rule out the polynomial curve fitting using least square numerical techniques via MATLAB compatible language. (author)

Whole-core LEU fuel demonstration in the ORR

International Nuclear Information System (INIS)

Snelgrove, J.L.; Bretscher, M.M.; Cornella, R.J.; Hobbs, R.W.

1985-01-01

A whole-core demonstration of LEU fuel in the ORR is expected to begin during November 1985. Fuel elements will contain U 3 Si 2 at 4.8 Mg U/m 3 and shim rod fuel followers will contain U 3 Si 2 at 3.5 Mg U/m 3 . Fuel fabrication is underway at B and W, CERCA, and NUKEM, with shipments scheduled to commence in October. The primary objectives of the demonstration are to provide data for validation of LEU and mixed-core fuel cycle calculations and to provide a large-scale demonstration of the acceptable performance of production-line U 3 Si 2 fuel elements. It is planned to approach the full LEU core through a series of mixed cores. Measurements to be made include flux distribution, reactivity swing, control rod worths, cycle length, fuel discharge burnup, gamma heating rates, β/sub eff/l, and isothermal temperature coefficient. Measurements will also be made on fresh LEU and fresh HEU critical configurations. Preliminary safety approval has been received and the final safety assessment is being reviewed

Method of measuring distance between fuel element

International Nuclear Information System (INIS)

Urata, Megumu.

1991-01-01

The distance between fuel elements contained in a pool is measured in a contactless manner even for a narrow distance less than 1 mm. That is, the equipment for measuring the distance between spent fuel elements of a spent fuel assembly in a nuclear reactor comprises a optical fiber scope, a lens, an industrial TV camera and a monitor TV. The top end of the optical fiber scope is inserted between fuel elements to be measured. The state thereof is displayed on the TV screen to measure the distance between the fuel elements. The measured results are compared with a previously formed calibration curve to determine the value between the fuel elements. Then, the distance between the fuel elements can be determined in the pool of a power plant without dismantling the fuel assembly, to investigate the state of the bending and estimate the fuel working life. (I.S.)

High-240Pu fuel worth in the Fast Test Reactor Engineering Mockup

International Nuclear Information System (INIS)

Daughtry, J.W.; Dobbin, K.D.

1975-01-01

Reactivity effects associated with the replacement of low- 240 Pu fuel with high- 240 Pu fuel were calculated and compared to measurements made in the FTR Engineering Mockup Critical (EMC). When the Pu and U isotopic compositions were changed in a way that increased the amounts of 240 Pu and 241 Pu and reduced the amounts of 239 Pu and 238 U while conserving total fissile mass and total fertile mass, the reactivity effect was positive. Calculation-to-experiment bias factors were obtained for this type of change and for the replacement of Fe 2 O 3 with U 3 O 8 in subassembly-size zones of the EMC. The k/sub e/--k/sub c/ bias decreased when high- 240 Pu fuel was introduced and increased when Fe 2 O 3 was replaced with U 3 O 8 . When the two changes were combined, their effects on the k/sub e/ --k/sub c/ bias tended to cancel out. The work described is related to plans for the utilization of light water reactor discharge Pu in the FTR

Significant others and contingencies of self-worth: activation and consequences of relationship-specific contingencies of self-worth.

Science.gov (United States)

Horberg, E J; Chen, Serena

2010-01-01

Three studies tested the activation and consequences of contingencies of self-worth associated with specific significant others, that is, relationship-specific contingencies of self-worth. The results showed that activating the mental representation of a significant other with whom one strongly desires closeness led participants to stake their self-esteem in domains in which the significant other wanted them to excel. This was shown in terms of self-reported contingencies of self-worth (Study 1), in terms of self-worth after receiving feedback on a successful or unsatisfactory performance in a relationship-specific contingency domain (Study 2), and in terms of feelings of reduced self-worth after thinking about a failure in a relationship-specific contingency domain (Study 3). Across studies, a variety of contingency domains were examined. Furthermore, Study 3 showed that failing in an activated relationship-specific contingency domain had negative implications for current feelings of closeness and acceptance in the significant-other relationship. Overall, the findings suggest that people's contingencies of self-worth depend on the social situation and that performance in relationship-specific contingency domains can influence people's perceptions of their relationships.

The ORR Whole-Core LEU Fuel Demonstration

International Nuclear Information System (INIS)

Bretscher, M.M.; Snelgrove, J.L.

1990-01-01

The ORR Whole-Core LEU Fuel Demonstration, conducted as part of the US Reduced Enrichment Research and Test Reactor Program, has been successfully completed. Using commercially-fabricated U 3 Si 2 -Al 20%-enriched fuel elements (4.8 g U/cc) and fuel followers (3.5 g U/cc), the 30-MW Oak Ridge Research Reactor was safely converted from an all-HEU core, through a series of HEU/LEU mixed transition cores, to an all-LEU core. There were no fuel element failures and average discharge burnups were measured to be as high as 50% for the standard elements and 75% for the fuel followers. Experimental results for burnup-dependent critical configurations, cycle-averaged fuel element powers, and fuel-element-averaged 235 U burnups validated predictions based on three-dimensional depletion calculations. Calculated values for plutonium production and isotopic mass ratios as functions of 235 U burnup support the corresponding measured quantities. In general, calculations for reaction rate distributions, control rod worths, prompt neutron decay constants, and isothermal temperature coefficients were found to agree with corresponding measured values. Experimentally determined critical configurations for fresh HEU and LEU cores radially reflected with water and with beryllium are well-predicted by both Monte Carlo and diffusion calculations. 17 refs

A longitudinal assessment of the links between physical activity and physical self-worth in adolescent females.

Science.gov (United States)

Raudsepp, Lennart; Neissaar, Inga; Kull, Merike

2013-01-01

A longitudinal framework was used to examine the hypotheses of (1) whether physical activity predicts changes in physical self-worth or (2) whether physical self-worth predicts changes in physical activity in adolescent girls. Participants (n=272) completed measures of physical self-worth and participation in physical activities at three different points spanning a two-year interval. A cross-lagged panel model using structural equation modelling analyses indicated that physical self-worth predicted subsequent physical activity and physical activity in turn predicted subsequent physical self-worth across time. Findings demonstrated a reciprocal relationship between physical self-worth and physical activity during early adolescence. This study supports the use of the reciprocal effects model (REM) in gaining an understanding of the cross-lagged relationships between physical self-worth and participation in physical activities amongst adolescent girls.

Comparable Worth Theory and Policy.

Science.gov (United States)

Wittig, Michele Andrisin; Lowe, Rosemary Hays

1989-01-01

Provides different perspectives on comparable worth issues. Covers the following topics: (1) competing explanations for the wage gap; (2) indirect approaches to wage equity; (3) the need for a direct approach to wage equity; (4) job evaluation; (5) application of comparable worth principles to compensation systems; and (6) strategies for adopting…

Application of a spatial modal kinetic model for determination of control rod worths

International Nuclear Information System (INIS)

Gomez, A.; Waldman, R.M.

1993-01-01

A high-precision rod drop method based on a modal kinetic model, with low dependence on detector location, is proposed to measure the reactivity worth of control rods. This value is obtained from data adjustment for the delayed evolution. It is necessary to maintain the experimental data fluctuation in a small value so that the error of the control rod worth should not be large. A model was developed in order to relate the fluctuation with some parameters which may be modified in the measuring process. The method was applied in the RA-6 reactor to measure control rod worth. For practical purpose it was found that the method can be applied to 15 dollars and it does not depend on relative detector and control rod locations, as the method based on the Point Reactor Model does. (author). 2 refs

Adjusted Money's Worth Ratios in Life Annuities

OpenAIRE

Jaime Casassus; Eduardo Walker

2013-01-01

The Money's Worth Ratio (MWR) measures an annuity's actuarial fairness. It is calculated as the discounted present value of expected future payments divided by its cost. We argue that from the perspective of annuitants, this measure may overestimate the value-for-money obtained, since it does not adjust for liquidity or risk factors. Measuring these factors is challenging, requiring detailed knowledge of assets, liabilities, and of the stochastic processes followed by them. Using a multi-fact...

Construction and initial validation of the self-worth protection scale.

Science.gov (United States)

Thompson, Ted; Dinnel, Dale L

2003-03-01

The self-worth theory of achievement motivation holds that in certain circumstances students stand to gain by deliberately withdrawing effort. When failure occurs despite effort, students are likely to conclude that failure resulted from lack of ability. Thus, withdrawing effort offers a defence against conclusions of low ability, thereby protecting self-worth. We undertook to assess the psychometric properties of the Self-Worth Protection Scale (SWPS). Data were obtained from 243 participants (Study 1) and 411 participants (Study 2) enrolled in undergraduate psychology courses at a university in the United States. We administered a number of scales, including the SWPS and scales assessing a fear of negative evaluation, academic self-esteem, uncertain global self-evaluations, self-handicapping, and causal uncertainty. Exploratory factor analysis indicated a three-factor solution (ability doubts, the importance of ability as a criterion of self-worth, and an avoidance orientation) utilising 33 of the original 44 items. A confirmatory factor analysis indicated that this three-factor solution was a poor fit of the data. After modifying the model, a confirmatory factor analysis indicated that a three-factor solution with 26 of the original items and a higher order factor of self-worth protection was an adequate fit of the data. Reliability measures were acceptable for the three subscales and total score. The total score of the SWPS was positively correlated with theoretically related constructs, demonstrating construct validity. The SWPS appears to be a psychometrically sound scale to assist in identifying individuals who manifest self-worth protection in achievement situations.

Attachment styles and contingencies of self-worth.

Science.gov (United States)

Park, Lora E; Crocker, Jennifer; Mickelson, Kristin D

2004-10-01

Previous research on attachment theory has focused on mean differences in level of self-esteem among people with different attachment styles. The present study examines the associations between attachment styles and different bases of self-esteem, or contingencies of self-worth, among a sample of 795 college students. Results showed that attachment security was related to basing self-worth on family support. Both the preoccupied attachment style and fearful attachment style were related to basing self-worth on physical attractiveness. The dismissing attachment style was related to basing self-worth less on others' approval, family support, and God's love.

Measuring deformation of Fuel pin in a Nuclear Fuel Test Rig

Energy Technology Data Exchange (ETDEWEB)

Heo, S. H.; Yang, T. H.; Hong, J. T.; Joung, C. Y.; Ahn, S. H.; Jang, S. Y.; Kim, J. H. [KAERI, Daejeon (Korea, Republic of)

2016-05-15

In this study, an LVDT core for measuring the longitudinal displacement of fuel pellets and clad was designed and produced. A signal processing method for the prepared core was investigated. The Nuclear Fuel Test Rig is used to observe changes in the characteristics of the fuel according to the neutron irradiation at HANARO (High-flux Advanced Neutron Application Reactor), which is a research reactor. Which are the strain and internal temperature of the irradiated nuclear fuel and the internal pressure of fuel due to fission gas, the characteristics of the fuel are measured using various sensors such as a thermocouple, SPND and LVDT. In this study, two shaped LVDT (Linear Variable Differential Transformer) cores for displacement measurements were designed and manufactured in order to measure the displacement of a fuel pellet and cladding tube using LVDT sensors for measuring electrical signals by converting the physical variation such as the force and displacement into a linear motion. In addition, signals from the manufactured LVDT sensor were collected and calibrated. Moreover, a method for obtaining the displacement in the core according to the sensing signal was planned. A derived equation can used to predict the change in the position of core. A following study should be conducted to test the output signal and real variation of out-pile system. For further work, a performance verification is required for an in-pile irradiation test.

40 CFR 610.42 - Fuel economy measurement.

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 29 2010-07-01 2010-07-01 false Fuel economy measurement. 610.42... ECONOMY RETROFIT DEVICES Test Procedures and Evaluation Criteria General Vehicle Test Procedures § 610.42 Fuel economy measurement. (a) Fuel consumption will be measured by: (1) The carbon balance method, or...

Fuel assembly for use in BWR type reactor

International Nuclear Information System (INIS)

Inaba, Yuzo.

1988-01-01

Purpose: To attain the reduction of neutron irradiation amount to control rods by the improvement in the reactor shutdown margin and the improvement of the control rod worth, by enhancing the arrangement of burnable poisons. Constitution: The number of burnable poison-incorporated fuel rods present in the outer two rows along the sides in adjacent with a control rod among the square lattice arrangement in a fuel assembly is decreased to less than 1/4 for that of total burnable poison-incorporated fuel rods, while the remaining burnable posion-incorporated fuel rods are arranged in the region other than above (that is, those regions not nearer to the control rod). Thus, even if a sufficient number of burnable poison to prolong the controlling effect for the reactivity with the burnable contents as the fuel assembly are disposed, only the burnable poison -incorporated fuel rods by the number less than 1/4 for that of the total burnable poison-incorporated fuel rods are present near the control rod of the fuel assembly. Accordingly, the control rod worth at the initial stage of the burning is increased at both high and normal temperatures. (Kawakami, Y.)

Calibration of spent fuel measurement assembly

International Nuclear Information System (INIS)

Koleska, Michal; Viererbl, Ladislav; Marek, Milan

2014-01-01

The LVR-15 research reactor (Czech Republic) had been converted from the highly enriched IRT-2M to the low enriched IRT-4M fuel. For the possibility of the independent pre-transport evaluation of IRT-2M burnup, a spectrometric system was developed. This spectrometric system consists of the fuel holder, the collimator and the portable Canberra Big MAC HPGe (High Purity Germanium) detector. In order to have well reproducible and reliable experimental data for modeling of the measurement system, calibration with the 110m Ag isotope with known activity was performed. This isotope was chosen for having energies similar to isotopes measured in fuel assemblies. The 110m Ag isotope was prepared by irradiating of the silver foil in LVR-15 research reactor; its activity was evaluated in the LVR-15's spectrometric laboratory. From the measured data, an efficiency curve of the spectrometric system has been determined. The experimental data were compared to the calculation results with the MCNPX model of the spectrometric system. - Highlights: • Calibration of research reactor spent fuel measurement assembly. • On-site prepared 110m Ag isotope used for the measurement. • Calculated self-shielding factor for the IRT-2M fuel. • Applicable to other research reactor fuel geometries

Objective and subjective measures of fuel poverty

International Nuclear Information System (INIS)

Waddams Price, Catherine; Brazier, Karl; Wang, Wenjia

2012-01-01

As energy prices continue to rise to reflect the real cost of carbon, the numbers of households in fuel poverty is increasing rapidly. This paper uses a unique data set to explore one alternative measure of fuel poverty based on whether consumers feel able to afford their energy, similar to one introduced by the government a few years ago. We explore the links between an expenditure based and our subjective measure of fuel poverty and compare these differences with those between official measures. Amongst low income households, 28% spent more than a tenth of their income on energy in the home, and so would be likely to qualify as fuel poor under the most usual definition; but only 16% felt unable to afford sufficient energy to keep their homes warm. Amongst this group who felt they had difficulty in affording sufficient energy, less than half were ‘expenditure’ fuel poor. The paper argues that reintroduction of a self-reported measure by the government would be a valuable aid to policy development. - Highlights: ► Defines a subjective measure of fuel poverty (unable to afford heating). ► Fewer households feel fuel poor than meet the official definition of fuel poverty. ► Though they are positively related, different factors affect the two measures. ► Those on standard and prepayment metres more likely to feel fuel poor. ► Large families spend a higher proportion of income on energy but are no more likely to feel fuel poor.

Transient feedback from fuel motion in metal IFR [Integral Fast Reactor] fuel

International Nuclear Information System (INIS)

Rhodes, E.A.; Stanford, G.S.; Regis, J.P.; Bauer, T.H.; Dickerman, C.E.

1990-01-01

Results from hodoscope data analyses are presented for TREAT transient-overpower tests M5 through M7 with emphasis on transient feedback mechanisms, including prefailure expansion at the tops of the fuel pins, subsequent dispersive axial fuel motion, and losses in relative worth of the fuel pins during the tests. Tests M5 and M6 were the first TOP tests of margin to cladding branch and prefailure elongation of D9-clad ternary (U-Pu-Zr) IFR-type fuel. Test M7 extended these results to high-burnup fuel and also initiated transient testing of HT9-clad binary (U-Zr) FFTF-driver fuel. Results show significant prefailure negative reactivity feedback and strongly negative feedback from fuel driven to failure. 4 refs., 6 figs

Dimensional measurement of fresh fuel bundle for CANDU reactor

International Nuclear Information System (INIS)

Jo, Chang Keun; Cho, Moon Sung; Suk, Ho Chun; Koo, Dae Seo; Jun, Ji Su; Jung, Jong Yeob

2005-01-01

This report describes the results of the dimensional measurement of fresh fuel bundles for the CANDU reactor in order to estimate the integrity of fuel bundle in two-phase flow in the CANDU-6 fuel channel. The dimensional measurements of fuel bundles are performed by using the 'CANDU Fuel In-Bay Inspection and Dimensional Measurement System', which was developed by this project. The dimensional measurements are done from February 2004 to March 2004 in the CANDU fuel storage of KNFC for the 36 fresh fuel bundles, which are produced by KNFC and are waiting for the delivery to the Wolsong-3 plant. The detail items of dimensional measurements are included fuel rod and bearing pad profiles of the outer ring in fuel bundle, diameter of fuel bundle, bowing of fuel bundle, fuel rod length, and surface profile of end plate profile. The measurement data will be compared with those of the post-irradiated bundles cooled in Wolsong-3 NPP spent fuel pool by using the same bundles and In-Bay Measurement System. So, this analysis of data will be applied for the evaluation of fuel bundle integrity in two-phase flow of the CANDU-6 fuel channel

76 FR 16345 - Net Worth and Equity Ratio

Science.gov (United States)

2011-03-23

... acquisition must be measured under generally accepted accounting principles as referenced in the Act. 12 U.S.C... equity or member interest in the acquirer. Generally accepted accounting principles require this excess... generally accepted accounting principles. For low income-designated credit unions, net worth also includes...

Application of the Modified Source Multiplication (MSM) Technique to Subcritical Reactivity Worth Measurements in Thermal and Fast Reactor Systems

International Nuclear Information System (INIS)

Blaise, P.; Fougeras, Ph.; Mellier, F.

2011-01-01

The Amplified Source Multiplication (ASM) method and its improved Modified Source Multiplication (MSM) method have been widely used in the CEA's EOLE and MASURCA critical facilities over the past decades for the determination of reactivity worths by using fission chambers in subcritical configurations. The ASM methodology uses relatively simple relationships between count rates of efficient miniature fission chambers located in slightly subcritical reference and perturbed configurations. While this method works quite well for small reactivity variations, the raw results need to be corrected to take into account the flux perturbation at the fission chamber location. This is performed by applying to the measurement a correction factor called MSM. This paper describes in detail both methodologies, with their associated uncertainties. Applications on absorber cluster worth in the MISTRAL-4 full MOX mock-up core and the last core loaded in MASURCA show the importance of the MSM correction on raw ASM data. (authors)

Measurement of fuel importance distribution in non-uniformly distributed fuel systems

International Nuclear Information System (INIS)

Yamane, Yoshihiro; Hirano, Yasushi; Yasui, Hazime; Izima, Kazunori; Shiroya, Seiji; Kobayashi, Keiji.

1995-01-01

A reactivity effect due to a spatial variation of nuclear fuel concentration is an important problem for nuclear criticality safety in a reprocessing plant. As a theory estimating this reactivity effect, the Goertzel and fuel importance theories are well known. It has been shown that the Goertzel's theory is valid in the range of our experiments based on measurements of reactivity effect and thermal neutron flux in non-uniformly distributed fuel systems. On the other hand, there have been no reports concerning systematic experimental studies on the flatness of fuel importance which is a more general index than the Goertzel's theory. It is derived from the perturbation theory that the fuel importance is proportional to the reactivity change resulting from a change of small amount of fuel mass. Using a uniform and three kinds of nonuniform fuel systems consisting of 93.2% enriched uranium plates and polyethylene plates, the fuel importance distributions were measured. As a result, it was found experimentally that the fuel importance distribution became flat, as its reactivity effect became large. Therefore it was concluded that the flatness of fuel importance distribution is the useful index for estimating reactivity effect of non-uniformly distributed fuel system. (author)

MOX fuel design and development consideration

International Nuclear Information System (INIS)

Yamate, K.; Abeta, S.; Suzuki, K.; Doi, S.

1997-01-01

Pu thermal utilization in Japan will be realized in several plants in late 1990's, and will be expanded gradually. For this target, adequacy of methods for MOX fuel design, nuclear design, and safety analysis has been evaluated by the committee of competent authorities organized by government in advance of the licensing application. There is no big difference of physical properties and irradiation behaviors between MOX fuel and UO 2 fuel, because Pu content of MOX fuel for Pu thermal utilization is low. The fuel design code for UO 2 fuel will be applied with some modifications, taking into account of characteristic of MOX fuel. For nuclear design, new code system is to be applied to treat the heterogeneity in MOX fuel assembly and the neutron spectrum interaction with UO 2 fuel more accurately. For 1/3 MOX fueled core in three loop plant, it was confirmed that the fuel rod mechanical design could meet the design criteria, with slight reduction of initial back-fitting pressure, and with appropriate fuel loading patterns in the core to match power with UO 2 fuel. With the increase of MOX fuel fraction in the core, control rod worth and boron worth decrease. Compensating the decrease by adding control rod and utilizing enriched B-10 in safety injection system, 100% MOX fueled core could be possible. Up to 1/3 MOX fueled core in three loop plant, no such modifications of the plant is necessary. The fraction of MOX fuel in PWR is designed to less than 1/3 in the present program. In order to improve Pu thermal utilization in future, various R and D program on fuel design and nuclear design are being performed, such as the irradiation program of MOX fuel manufactured through new process to the extent of high burnup. (author). 8 refs, 9 figs, 2 tabs

Monte Carlo verification of control-rod worth for the Savannah River K reactor

International Nuclear Information System (INIS)

Mosteller, R.D.

1992-01-01

The Savannah River K Reactor is a heavy-water reactor that relies on control-rod movement to control its reactivity and power distribution during normal operations. It is necessary, therefore, to have an accurate estimate of the reactivity worth of its control rods in order to predict the behavior of the reactor. Westinghouse Savannah River Company (WSRC) uses the GLASS lattice-physics code to calculate few-group cross sections for fuel and control-rod assemblies in the K reactor. This paper compares the control-rod worth calculated by GLASS to that calculated by the MCNP Monte Carlo program. The GLASS calculations utilize its standard 37-group cross-section library, while the MCNP calculations employ continuous-energy isotopic cross-section libraries derived from ENDF/B-V. The MCNP calculations therefore combine the most rigorous analytical model and the most accurate cross sections currently available for thermal-reactor analysis. Consequently, the MCNP results comprise a computational benchmark against which the accuracy of the GLASS code can be evaluated

The whole-core LEU silicide fuel demonstration in the JMTR

Energy Technology Data Exchange (ETDEWEB)

Aso, Tomokazu; Akashi, Kazutomo; Nagao, Yoshiharu [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)] [and others

1997-08-01

The JMTR was fully converted to LEU silicide (U{sub 3}Si{sub 2}) fuel with cadmium wires as burnable absorber in January, 1994. The reduced enrichment program for the JMTR was initiated in 1979, and the conversion to MEU (enrichment ; 45%) aluminide fuel was carried out in 1986 as the first step of the program. The final goal of the program was terminated by the present LEU conversion. This paper describes the results of core physics measurement through the conversion phase from MEU fuel core to LEU fuel core. Measured excess reactivities of the LEU fuel cores are mostly in good agreement with predicted values. Reactivity effect and burnup of cadmium wires, therefore, were proved to be well predicted. Control rod worth in the LEU fuel core is mostly less than that in the MEU fuel core. Shutdown margin was verified to be within the safety limit. There is no significant difference in temperature coefficient of reactivity between the MEU and LEU fuel cores. These results verified that the JMTR was successfully and safely converted to LEU fuel. Extension of the operating cycle period was achieved and reduction of spend fuel elements is expected by using the fuel with high uranium density.

Fuel motion in overpower tests of metallic integral fast reactor fuel

International Nuclear Information System (INIS)

Rhodes, E.A.; Bauer, T.H.; Stanford, G.S.; Regis, J.P.; Dickerman, C.E.

1992-01-01

In this paper results from hodoscope data analyses are presented for transient overpower (TOP) tests M5, M6, and M7 at the Transient Reactor Test Facility, with emphasis on transient feedback mechanisms, including prefailure expansion at the tops of the fuel pins, subsequent dispersive axial fuel motion, and losses in relative worth of the fuel pins during the tests. Tests M5 and M6 were the first TOP tests of margin to cladding breach and prefailure elongation of D9-clad ternary (U-Pu-Zr) integral fast reactor-type fuel. Test M7 extended these results to high-burnup fuel and also initiated transient testing of HT-9-clad binary (U-Zr) Fast Flux Test Facility driver fuel. Results show significant prefailure negative reactivity feedback and strongly negative feedback from fuel driven to failure

Parent-adolescent attachment and procrastination: The mediating role of self-worth.

Science.gov (United States)

Chen, Bin-Bin

2017-01-01

Within the theoretical framework of attachment theory, the author examined associations between adolescents' procrastination and their attachment relationships with both mothers and fathers, and explored the potential mediation role of self-worth in these associations. Participants were 384 Chinese adolescents (49.6% boys, average age 15.13 years) from public schools in Shanghai, China. They completed self-report measures of 3 dimensions of parental attachment (i.e., trust, communication, and alienation), general self-worth, and procrastination. The results indicated that both paternal and maternal trust and paternal communication were negatively associated with higher levels of procrastination whereas both paternal and maternal alienation were positively associated with procrastination. In addition, self-worth mediated the associations among 3 dimensions of parental attachment and procrastination. The findings highlighted the importance of parental attachment-based intervention strategies to reduce procrastination among adolescents.

Measurement of the fuel temperature and the fuel-to-coolant heat transfer coefficient of Super Phenix 1 fuel elements

International Nuclear Information System (INIS)

Edelmann, M.

1995-12-01

A new measurement method for measuring the mean fuel temperature as well as the fuel-to-coolant heat transfer coefficient of fast breeder reactor subassemblies (SA) is reported. The method is based on the individual heat balance of fuel SA's after fast reactor shut-downs and uses only the plants normal SA outlet temperature and neutron power signals. The method was used successfully at the french breeder prototype Super Phenix 1. The mean SA fuel temperature as well as the heat transfer coefficient of all SPX SA's have been determined at power levels between 15 and 90% of nominal power and increasing fuel burn-up from 3 to 83 EFPD (Equivalent of Full Power-Days). The measurements also provided fuel and whole SA time constants. The estimated accuracy of measured fuel parameters is in the order of 10%. Fuel temperatures and SA outlet temperature transients were also calculated with the SPX1 systems code DYN2 for exactly the same fuel and reactor operating parameters as in the experiments. Measured fuel temperatures were higher than calculated ones in all cases. The difference between measured and calculated core mean values increases from 50 K at low power to 180 K at 90% n.p. This is about the double of the experimental error margins. Measured SA heat transfer coefficients are by nearly 20% lower than corresponding heat transfer parameters used in the calculations. Discrepancies found between measured and calculated results also indicate that either the transient heat transfer in the gap between fuel and cladding (gap conductance) might not be exactly reproduced in the computer code or that the gap in the fresh fuel was larger than assumed in the calculations. (orig.) [de

MEASUREMENTS MADE DURING THE COMMISSIONING OF THE WINDSCALE ADVANCED GAS- COOLED REACTOR (WAGR)

Energy Technology Data Exchange (ETDEWEB)

Gallie, R R

1963-06-15

Some measurements were made on the WAGR fuel in the APEX and HERO facilities in order to derive the loading pattern. The reactor was then loaded, and measurements of the excess reactivity were in good agreement with predictions. Other measurements made were control rod calibrations, reactivity worths of control rods, isothermal temperature coefficient, and pressure drops. (D.L.C.)

Pyrometric fuel particle measurements in pressurised reactors

Energy Technology Data Exchange (ETDEWEB)

Hernberg, R; Joutsenoja, T [Tampere Univ. of Technology (Finland)

1997-10-01

A fibre-optic two-colour pyrometric technique for fuel particle temperature and size measurement is modified and applied to three pressurised reactors of different type in Finland, Germany and France. A modification of the pyrometric method for simultaneous in situ measurement of the temperature and size of individual pulverised coal particles at the pressurised entrained flow reactor of VTT Energy in Jyvaeskylae was developed and several series of measurements were made in order to study the effects of oxygen concentration (3-30 vol%) and pressure (0.2-1.0 MPa) on the particle temperature. The fuels used in the experiments were Westerholt, Polish and Goettelborn hvb coals, Gardanne lignite and Niederberg anthracite. The initial nominal fuel particle size varied in the experiments from 70 to 250 ,{mu}m and the gas temperature was typically 1173 K. For the anthracite also the effects of gas temperature (1073-1423K) and CO{sub 2} concentration (6-80 vol%) were studied. In Orleans a fibreoptic pyrometric device was installed to a pressurised thermogravimetric reactor of CNRS and the two-colour temperatures of fuel samples were measured. The fuel in the experiments was pulverised Goettelborn char. The reliability of optical temperature measurement in this particular application was analysed. In Essen a fibre-optic pyrometric technique that is capable to measure bed and fuel particle temperatures was applied to an atmospheric fluidised bed reactor of DMT. The effects of oxygen concentration (3-8 vol%) and bed temperature (1123-1193 K) on the fuel particle temperature were studied. The fuels in these were Westerholt coal and char and EBV-coal. Some results of these measurements are presented. The project belonged to EU`s Joule 2 extension research programme (contract JOU2-CT93-0331). (orig.)

Development and Validation of the Sexual Contingent Self-Worth Scale.

Science.gov (United States)

Glowacka, Maria; Rosen, Natalie O; Vannier, Sarah; MacLellan, Margaret C

2017-01-01

Sexual contingent self-worth (CSW) refers to self-worth that is dependent on maintaining a sexual relationship, and has not been studied previously. This novel construct may have implications for sexual, relationship, and psychological well-being, because it could affect the cognitions, affect, and behaviors of individuals in sexual relationships. The purpose of this study was to develop the Sexual Contingent Self-Worth Scale and examine its reliability and validity in community samples. Two separate online studies (N = 329 and N = 282) included men and women who were in committed, sexually active relationships. The Sexual CSW Scale was adapted from a validated measure of relationship CSW. In Study 1, participants completed the Sexual CSW Scale, whereas in Study 2, participants also responded to standardized measures of related constructs. In addition, participants completed the Sexual CSW Scale again two weeks later in Study 2. Factor analysis yielded two subscales: (a) sexual CSW dependent on positive sexual events in the relationship and (b) sexual CSW dependent on negative sexual events. Results indicated good construct validity, incremental validity, internal consistency, and test-retest reliability for the Sexual CSW Scale. This research contributes to the fields of both CSW and sexuality by introducing a novel domain of CSW.

Measuring vibrations in fuel channels CNE

International Nuclear Information System (INIS)

Martín Ghiselli, A.; Fiori, J.; Sacchi, M.; Villabrille, G.

2013-01-01

This paper present a description of implementation and execution of vibration measurements made at the request of NUCLEOELECTRICA ARGENTINA S.A. on the ends of the reactor fuel channels of Embalse Nuclear Power Plant to explore possible differences between the dynamic behavior of empty fuel channel and with full charge of fuel elements inside. (author)

What the Common Economic Arguments against Comparable Worth Are Worth.

Science.gov (United States)

Bergmann, Barbara R.

1989-01-01

Reviews economists' views about how the economy works, from which conclusions opposing comparable worth are drawn. Discusses factors that have been omitted from economists' views--social and psychological factors that affect behavior in the workplace, permit and encourage discrimination, and have an effect on the distribution of jobs and wages.…

A design study of high breeding ratio sodium cooled metal fuel core without blanket fuels

International Nuclear Information System (INIS)

Kobayashi, Noboru; Ogawa, Takashi; Ohki, Shigeo; Mizuno, Tomoyasu; Ogata, Takanari

2009-01-01

The metal fuel core is superior to the mixed oxide fuel core because of its high breeding ratio and compact core size resulting from hard neutron spectrum and high heavy metal densities. Utilizing these characteristics, a conceptual design for a high breeding ratio was performed without blanket fuels. The design conditions were set so a sodium void worth of less than 8 $, a core height of less than 150 cm, the maximum cladding temperature of 650degC, and the maximum fuel pin bundle pressure drop of 0.4 MPa. The breeding ratio of the resultant core was 1.34 with 6wt% zirconium content fuel. Applying 3wt% zirconium content fuel enhanced the breeding ratio up to 1.40. (author)

Enrichment measurement in TRIGA type fuels

International Nuclear Information System (INIS)

Aguilar H, F.; Mazon R, R.

2001-05-01

The Department of Energy of the United States of North America, through the program 'Idaho Operations Nuclear Spent Fuel Program' of the Idaho National Engineering and Environmental Laboratory (INEEL), in Idaho Falls; Idaho USA, hires to Global Technologies Inc. (GTI) to develop a prototype device of detection enrichment uranium (DEU Detection of Enrichment of Uranium) to determine quantitatively the enrichment in remainder U-235 in a TRIGA fuel element at the end of it useful life. The characteristics of the prototype developed by GTI are the following ones: It allows to carry out no-destructive measurements of TRIGA type fuel. Easily transportable due to that reduced of it size. The determination of the enrichment (in grams of U-235) it is obtained with a precision of 5%. The National Institute of Nuclear Research (ININ), in its facilities of the Nuclear Center of Mexico, it has TRIGA type fuel of high and low enrichment (standard and FLIP) fresh and with burnt, it also has the infrastructure (hot cells, armor-plating of transport, etc) and qualified personnel to carry out the necessary maneuvers to prove the operation of the DEU prototype. For this its would be used standard type fuel elements and FLIP, so much fresh as with certain burnt one. In the case of the fresh fuels the measurement doesn't represent any risk, the fuels before and after the measurement its don't contain a quantity of fission products that its represent a radiological risk in its manipulation; but in the case of the fuels with burnt the handling of the same ones represents an important radiological risk reason why for its manipulation it was used the transport armor-plating and the hot cells. (Author)

Spent fuel critical masses and supportive measurements

International Nuclear Information System (INIS)

Toffer, H.; Wells, A.H.

1987-01-01

Critical masses for spent fuel are larger than for green fuel and therefore use of the increased masses could result in improved handling, storage, and transport of such materials. To apply spent fuel critical masses requires an assessment of fuel exposure and the corresponding isotopic compositions. The paper discusses several approaches at the Hanford N Reactor in establishing fuel exposure, including a direct measurement of spent to green fuel critical masses. The benefits derived from the use of spent fuel critical masses are illustrated for cask designs at the Nuclear Assurance Corporation. (author)

19 CFR 212.11 - Net worth exhibit.

Science.gov (United States)

2010-04-01

... 19 Customs Duties 3 2010-04-01 2010-04-01 false Net worth exhibit. 212.11 Section 212.11 Customs Duties UNITED STATES INTERNATIONAL TRADE COMMISSION INVESTIGATIONS OF UNFAIR PRACTICES IN IMPORT TRADE IMPLEMENTATION OF THE EQUAL ACCESS TO JUSTICE ACT Information Required From Applicants § 212.11 Net worth exhibit...

Measurements and analyses on reactivity effects of absorber rods in a light-water moderated UO2 lattices

International Nuclear Information System (INIS)

Murakami, Kiyonobu; Miyoshi, Yoshinori; Hirose, Hideyuki; Suzaki, Takenori

1985-03-01

Reactivity effects and reactivity-interference effects of absorber rods were measured with a cylindrical core aiming to obtain bench-marks for verification of the calculational methods. The core consisted of 2.6 w/o enriched UO 2 fuel rods lattice of which water-to-fuel volume ratio was 1.83. In the experiment, the critical water levels were measured changing neutron absorber content of absorber rods and the distance between two absorber rods in the core center. Monte Calro codes KENO-IV and MULTI-KENO were used to calculate reactivity worthes of absorber rods. The calculational results of effective multiplication factors ranged from 0.978 to 0.999 for the 60 cases of critical cores with inserted absorber rods. The calculational results of absorber worthes agreed to the experimental results within twice of the standerd deviation accompanied with the Monte Calro calculation. (author)

The whole-core LEU fuel demonstration in the ORR

International Nuclear Information System (INIS)

Snelgrove, J.L.; Bretscher, M.M.; Cornella, R.J.; Hobbs, R.W.

1985-01-01

A whole-core demonstration of LEU fuel in the ORR is expected to begin during November 1985. Fuel elements will contain U 3 Si 2 at 4.8 Mg U/m 3 and shim rod fuel followers will contain U 3 Si 2 at 3.5 Mg U/m 3 . Fuel fabrication is underway at B and W, CERCA, and NUKEM, with shipments scheduled to commence in October. The primary objectives of the demonstration are to provide data for validation of LEU and mixed-core fuel cycle calculations and to provide a large-scale demonstration of the acceptable performance of production-line U 3 Si 2 fuel elements. It is planned to approach the full LEU core through a series of mixed cores. Measurements to be made include flux distribution, reactivity swing, control rod worth, cycle length, fuel discharge burn-up, gamma heating rate, β eff /l, and isothermal temperature coefficient. Measurements will also be made on fresh LEU and fresh HEU critical configurations. Preliminary safety approval has been received and the final safety assessment is being reviewed. Key issues being addressed in the safety assessment are fuel performance, radiological consequences, margin to burnout and transient behavior. The LEU core is comparable in all safety aspects to the HEU core and the transition core is only marginally worse owing to higher power seeking factors. (author)

Nuclear reactor, fuel assembly and neutron measuring system

International Nuclear Information System (INIS)

Chaki, Masao; Murase, Michio; Zukeran, Atsushi; Moriya, Kimiaki

1998-01-01

The present invention provides a BWR type reactor improved with the efficiency of used fuels and fuel economy by increasing a rated power and reducing exchange fuels. Namely, in a BWR type reactor at present, a thermal limit value is determined by conducting nuclear calculation of the reactor core based on data of reactor flow rate measurement and data of neutron flux measurement. However, since the neutron calculation of the reactor core is based on fuel assemblies while the points for the neutron measurement are present at the outside of the fuel assemblies, errors are caused. A margin including the errors has been used as a thermal limit value during operation. In the present invention, neutron fluxes in the fuel assembly as a base of the nuclear calculation can be measured by the same number of neutron detector tubes, but the number of the measuring points is increased to four times. With such procedures, errors caused by the difference of the neutron calculation and values at neutron measuring points can be reduced. As a result, a margin of the thermal limit value is reduced to increase the degree of freedom of reactor operation. Then, the economical property of the reactor operation can be improved. (N.H.)

Characteristics of self-worth protection in achievement behaviour.

Science.gov (United States)

Thompson, T

1993-11-01

Two experiments are reported comprising an investigation of individual difference variables associated with self-worth protection. This is a phenomenon whereby students in achievement situations adopt one of a number of strategies, including withdrawing effort, in order to avoid damage to self-esteem which results from attributing failure to inability. Experiment 1 confirmed the adequacy of an operational definition which identified self-worth students on the basis of two criteria. These were deteriorated performance following failure, together with subsequent enhanced performance following a face-saving excuse allowing students to explain failure without implicating low ability. The results of Experiment 2 established that the behaviour of self-worth protective students in achievement situations may be understood in terms of their low academic self-esteem coupled with uncertainty about their level of global self-esteem. Investigation of the manner in which self-worth students explain success and failure outcomes failed to demonstrate a tendency to internalise failure but revealed a propensity on the part of these students to reject due credit for their successes. The implications of these findings in terms of the prevention and modification of self-worth protective reactions in achievement situations are discussed.

A comparative study to investigate burnup in research reactor fuel using two independent experimental methods

International Nuclear Information System (INIS)

Iqbal, M.; Mehmood, T.; Ayazuddin, S.K.; Salahuddin, A.; Pervez, S.

2001-01-01

Two independent experimental methods have been used for the comparative study of fuel burnup measurement in low enriched uranium, plate type research reactor. In the first method a gamma ray activity ratio method was employed. An experimental setup was established for gamma ray scanning using prior calibrated high purity germanium detector. The computer software KORIGEN gave the theoretical support. In the second method reactivity difference technique was used. At the same location in the same core configuration the fresh and burned fuel element's reactivity worth was estimated. For theoretical estimated curve, group cross-sections were generated using computer code WIMS-D/4, and three dimensional modeling was made by computer code CITATION. The measured burnup of different fuel elements using these methods were found to be in good agreement

Assessment of CANDU physics codes using experimental data - II: CANDU core physics measurements

International Nuclear Information System (INIS)

Roh, Gyu Hong; Jeong, Chang Joon; Choi, Hang Bok

2001-11-01

Benchmark calculations of the advanced CANDU reactor analysis tools (WIMS-AECL, SHETAN and RFSP) and the Monte Carlo code MCNP-4B have been performed using Wolsong Units 2 and 3 Phase-B measurement data. In this study, the benchmark calculations have been done for the criticality, boron worth, reactivity device worth, reactivity coefficient, and flux scan. For the validation of the WIMS-AECL/SHETANRFSP code system, the lattice parameters of the fuel channel were generated by the WIMS-AECL code, and incremental cross sections of reactivity devices and structural material were generated by the SHETAN code. The results have shown that the criticality is under-predicted by -4 mk. The reactivity device worths are generally consistent with the measured data except for the strong absorbers such as shutoff rod and mechanical control absorber. The heat transport system temperature coefficient and flux distributions are in good agreement with the measured data. However, the moderator temperature coefficient has shown a relatively large error, which could be caused by the incremental cross-section generation methodology for the reactivity device. For the MCNP-4B benchmark calculation, cross section libraries were newly generated from ENDF/B-VI release 3 through the NJOY97.114 data processing system and a three-dimensional full core model was developed. The simulation results have shown that the criticality is estimated within 4 mk and the estimated reactivity worth of the control devices are generally consistent with the measurement data, which implies that the MCNP code is valid for CANDU core analysis. In the future, therefore, the MCNP code could be used as a reference tool to benchmark design and analysis codes for the advanced fuels for which experimental data are not available

Physical Activity and Global Self-worth in a Longitudinal Study of Children.

Science.gov (United States)

Reddon, Hudson; Meyre, David; Cairney, John

2017-08-01

Physical activity is associated with an array of physical and mental health benefits among children and adolescents. The development of self-worth/self-esteem has been proposed as a mechanism to explain the mental health benefits derived from physical activity. Despite several studies that have analyzed the association between physical activity and self-worth, the results have been inconsistent. It is also uncertain how related physical health measures, such as sedentary behavior, body composition, and fitness, influence the relationship between physical activity and self-worth over time. In the present study, we 1) analyzed if the association between physical activity and self-worth remained constant over time and whether this relationship varied by sex and 2) investigated if changes in body composition and fitness level mediated the relationship between physical activity and self-worth. Data from the Physical Health Activity Study Team were used for this analysis. The Physical Health Activity Study Team is a prospective cohort study that included 2278 children at baseline (ages 9-10 yr) and included eight follow-up contacts for a 4-yr study period. Linear mixed-effects models were used to estimate global self-worth (GSW) over follow-up. Increased physical activity was associated with greater GSW across all waves of data collection, and this relationship did not vary significantly over time or between sexes. Aerobic fitness was positively associated with GSW, whereas body mass index (BMI) was inversely related to GSW. Both aerobic fitness and BMI appeared to mediate the association between physical activity and GSW. Sedentary behavior was not significantly associated with GSW. Physical activity is associated with greater GSW, and this relationship appears to be mediated by BMI and aerobic fitness. These findings reinforce the importance of physical behaviors and physical characteristics in shaping GSW in children.

Contingencies of self-worth and social-networking-site behavior.

Science.gov (United States)

Stefanone, Michael A; Lackaff, Derek; Rosen, Devan

2011-01-01

Social-networking sites like Facebook enable people to share a range of personal information with expansive groups of "friends." With the growing popularity of media sharing online, many questions remain regarding antecedent conditions for this behavior. Contingencies of self-worth afford a more nuanced approach to variable traits that affect self-esteem, and may help explain online behavior. A total of 311 participants completed an online survey measuring such contingencies and typical behaviors on Facebook. First, exploratory factor analyses revealed an underlying structure to the seven dimensions of self-worth. Public-based contingencies explained online photo sharing (β = 0.158, p relationship with time online (β = -0.186, p relationship with the intensity of online photo sharing (β = 0.242), although no relationship was evident for time spent managing profiles.

Reactivity worth measurement of the control blades of the University of Florida training reactor

International Nuclear Information System (INIS)

Quintero-Leyva, Barbaro

1997-01-01

A series of experiments were carried out in order to measure the reactivity worth of the safety and regulating blades of the University of Florida Training Reactor (UFTR) using the Inverse Kinetics, the Inverse Kinetics-Rod Drop method and the Power Ratio. The reactor's own instrumentation (compensated ion chamber) and an independent counting system (fission chamber) were used. A very smooth exponential decay of the flux was observed after 6s of the beginning of the transients using the reading of the reactor detector. The results of the measurements of the reactivity using both detectors were consistent and in good agreement. The compensated ion chamber showed a very smooth exponential behavior; this suggests that if we could record the power for a small sample time, say 0.1 s from the beginning of the transient, several additional research projects could be accomplished. First, precise intercomparison of the methods could be achieved if the statistics level is acceptable. Second, a precise description of the bouncing of the blades and its effects on the reactivity could be achieved. Finally, the design of a reactivity-meter could be based on such study. (author)

Coincidence measurements of FFTF breeder fuel subassemblies

International Nuclear Information System (INIS)

Eccleston, G.W.; Foley, J.E.; Krick, M.; Menlove, H.O.; Goris, P.; Ramalho, A.

1984-04-01

A prototype coincidence counter developed to assay fast breeder reactor fuel was used to measure four fast-flux test facility subassemblies at the Hanford Engineering Development Laboratory in Richland, Washington. Plutonium contents in the four subassemblies ranged between 7.4 and 9.7 kg with corresponding 240 Pu-effective contents between 0.9 and 1.2 kg. Large count rates were observed from the measurements, and plots of the data showed significant multiplication in the fuel. The measured data were corrected for deadtime and multiplication effects using established formulas. These corrections require accurate knowledge of the plutonium isotopics and 241 Am content in the fuel. Multiplication-corrected coincidence count rates agreed with the expected count rates based on spontaneous fission-neutron emission rates. These measurements indicate that breeder fuel subassemblies with 240 Pu-effective contents up to 1.2 kg can be nondestructively assayed using the shift-register electronics with the prototype counters. Measurements using the standard Los Alamos National Laboratory shift-register coincidence electronics unit can produce an assay value accurate to +-1% in 1000 s. The uncertainty results from counting statistics and deadtime-correction errors. 3 references, 8 figures, 8 tables

Software to improve spent fuel measurements using the FDET

International Nuclear Information System (INIS)

Staples, P.; Beddingfield, D.; Lestone, J.; Pelowitz, D.; Sprinkle, J.; Bytchkov, V.; Starovich, Z.; Harizanov, I.; Vellejo-Luna, J.; Lavender, C.

2001-01-01

Full text: Vast quantities of spent fuel are available for safeguard measurements, primarily in Commonwealth of Independent States (CIS) of the former Soviet Union. This spent fuel, much of which consists of long cooling time material, is going to become less unique in the world safeguards arena as reprocessing projects or permanent repositories continue to be delayed or postponed. The long cooling time of many of the spent fuel assemblies in the CIS countries being prepared for intermediate term storage promotes the possibility of increased accuracy in spent fuel assays. An important point to consider for the future that could advance safeguards measurements for re-verification and inspection measurements would be to determine what safeguards requirements should be imposed upon this 'new' class of spent fuel. Improvements in measurement capability will obviously affect the safeguards requirements. What most significantly enables this progress in spent fuel measurements is the improvement in computer processing power and software enhancements leading to user-friendly Graphical User Interfaces (GUI's). The software used for these projects significantly reduces the IAEA inspector's time both learning and operating computer and data acquisition systems. While at the same time by standardizing the spent fuel measurements it is possible to increase reproducibility and reliability of the measurement data. The inclusion of various analysis algorithms into the operating environment, which can be performed in real time upon the measurement data, can also lead to increases in safeguard reliability and improvements in efficiency to plant operations. (author)

On the problem of in-core fuel management in power reactors

International Nuclear Information System (INIS)

Marinkovic, N.; Matausek, M.V.

1985-01-01

Within the scope of in-core fuel management including refuelling schedule and reactivity control it is indispensable to define nuclear fuel worth, optimal depletion of the spent fuel assemblies as well as isotopic composition of the spent fuel. This paper shows the computed values of the mentioned parameters in case of different reactor types, PWR, WWER, HWR and BWR of 1000 MWe as well as the intensity of radiation of the spent fuel 3 and 1 years after fission.(author)

Burnup degree measuring device for spent fuel

International Nuclear Information System (INIS)

Doi, Hideo; Imaizumi, Hideki; Endo, Yasumi; Itahara, Kuniyuki.

1994-01-01

The present invention provides a small-sized and convenient device for measuring a burnup degree of spent fuels, which can be installed without remodelling an existent fuel storage pool. Namely, a gamma-ray detecting portion incorporates a Cd-Te detector for measuring intensity ratio of gamma-rays. A neutron detecting portion incorporates a fission counter tube. The Cd-Te detector comprises a neutron shielding member for reducing radiation damages and a background controlling plate for reducing low energy gamma-rays entering from a collimator. Since the Cd-Td detector for use in a gamma-ray spectroscopy can be used at a normal temperature and can measure even a relatively strong radiation field, it can measure the intensity of gamma-rays from Cs-137 and Cs-134 in spent fuels accurately at a resolving power of less than 10 keV. Further, in a case where a cooling period is less than one year, gamma-rays from Rh-106 and Nb-95 can also be measured. (I.S.)

42 CFR 422.382 - Minimum net worth amount.

Science.gov (United States)

2010-10-01

... that CMS considers appropriate to reduce, control or eliminate start-up administrative costs. (b) After... section. (c) Calculation of the minimum net worth amount—(1) Cash requirement. (i) At the time of application, the organization must maintain at least $750,000 of the minimum net worth amount in cash or cash...

A core design study for 'zero-sodium-void-worth' cores

International Nuclear Information System (INIS)

Kawashima, Masatoshi; Suzuki, Masao; Hill, R.N.

1992-01-01

Recently, a number of low sodium-void-worth metal-fueled core design concepts have been proposed; to provide for flexibility in transuranic nuclide management strategy, core designs which exhibit a wide range of breeding characteristics have been developed. Two core concepts, a flat annular (transuranic burning) core and an absorber-type parfait (transuranic self-sufficient) core, are selected for this study. In this paper, the excess reactivity management schemes applied in the two designs are investigated in detail. In addition, the transient effect of reactivity insertions on the parfait core design is assessed. The upper and lower core regions in the parfait design are neutronically decoupled; however, the common coolant channel creates thermalhydraulic coupling. This combination of neutronic and thermalhydraulic characteristics leads to unique behavior in anticipated transient overpower events. (author)

Development of CANDU fuel in-bay inspection and dimensional measurement system

Energy Technology Data Exchange (ETDEWEB)

Jo, Chang Keun; Cho, Moon Sung; Suk, Ho Chun; Koo, Dae Seo; Park, Kwang June; Jun, Ji Su; Jung, Jong Yeob

2004-09-01

In this report, we describe a spent fuel inspection and dimensional measuring system for CANDU fuel bundles in atmosphere or water. The submissible camera with radiation resistance was used to inspect a spent fuel in water. The design criteria of the dimensional measuring system was set up {+-}0.01mm(10{mu}m) for measuring accuracy. The LVDT (Linear Variable Differential Transformer) was used as measuring sensor in this dimensional measuring system. An LVDT calibration equipment was developed in order to satisfy the required accuracy of the system. Also, aluminum master and CANDU Master with same dimension of fuel bundle was made to calibrate the mechanical error of the dimensional measuring system. The accuracy of the fuel inspection system was examined. The results show that the accuracy in dimensional measurement of fuel rod profile and bearing pad profile, diameter of fuel bundle, fuel rod length, and end plate profile using standard test equipment satisfies the design criteria, i.e., maximum measurement error of {+-}0.01mm(10{mu}m)

Burnup measurements of leader fuel elements

International Nuclear Information System (INIS)

Henriquez, C; Navarro, G; Pereda, C

2000-01-01

Some time ago the CCHEN authorities decided to produce a set of 50 low enrichment fuel elements. These elements were produced in the PEC (Fuel Elements Plant), located at CCHEN offices in Lo Aguirre. These new fuel elements have basically the same geometrical characteristics of previous ones, which were British and made with raw material from the U.S. The principal differences between our fuel elements and the British ones is the density of fissile material, U-235, which was increased to compensate the reduction in enrichment. Last year, the Fuel Elements Plant (PEC) delivered the shipment's first four (4) fuel elements, called leaders, to the RECH1. A test element was delivered too, and the complete set was introduced into the reactor's nucleus, following the normal routine, but performing a special follow-up on their behavior inside the nucleus. This experimental element has only one outside fuel plate, and the remaining (15) structural plates are aluminum. In order to study the burnup, the test element was taken out of the nucleus, in mid- November 1999, and left to decay until June 2000, when it was moved to the laboratory (High Activity Cell), to start the burnup measurements, with a gamma spectroscopy system. This work aims to show the results of these measurements and in addition to meet the following objectives: (a) Visual test of the plate's general condition; (b) Sipping test of fission products; (c) Study of burn-up distribution in the plate; (d) Check and improve the calculus algorithm; (e) Comparison of the results obtained from the spectroscopy with the ones from neutron calculus

Physical activity, menopause, and quality of life: the role of affect and self-worth across time.

Science.gov (United States)

Elavsky, Steriani

2009-01-01

Physical activity has been shown to enhance quality of life (QOL); however, few investigations of these effects exist in women undergoing the menopausal transition. The present study examined the long-term effects of physical activity on menopause-related QOL and tested the mediating effects of physical self-worth and positive affect in this relationship. Middle-aged women previously enrolled in a 4-month randomized controlled trial involving walking and yoga, and a control group completed a follow-up mail-in survey 2 years after the end of the trial. The survey included a battery of psychological and physical activity measures, including measures of menopausal symptoms and menopause-related QOL. Longitudinal linear panel analysis was conducted within a covariance modeling framework to test whether physical self-worth and positive affect mediated the physical activity-QOL relationship over time. At the end of the trial, physical activity and menopausal symptoms were related to physical self-worth and positive affect, and in turn, greater levels of physical self-worth and positive affect were associated with higher levels of menopause-related QOL. Analyses indicated that increases in physical activity and decreases in menopausal symptoms over the 2-year period were related to increases in physical self-worth (betas = 0.23 and -0.52, physical activity and menopausal symptoms, respectively) and, for symptoms, also to decreased positive affect (beta = -0.47), and both physical self-worth (beta = 0.34) and affect (beta = 0.43) directly influenced enhancements in QOL (R = 0.775). The findings support the position that the effects of physical activity on QOL are mediated, in part, by intermediate psychological outcomes and that physical activity can have long-term benefits for women undergoing the menopausal transition.

1st Fire Behavior and Fuels Conference: Fuels Management-How to Measure Success

Science.gov (United States)

Patricia L. Andrews

2006-01-01

The 1st Fire Behavior and Fuels Conference: Fuels Management -- How to Measure Success was held in Portland, Oregon, March 28-30, 2006. The International Association of Wildland Fire (IAWF) initiated a conference on this timely topic primarily in response to the needs of the U.S. National Interagency Fuels Coordinating Group (http://www.nifc.gov/).

Pyrometric fuel particle measurements in pressurised reactors

Energy Technology Data Exchange (ETDEWEB)

Hernberg, R.; Joutsenoja, T. [Tampere Univ. of Technology (Finland)

1996-12-01

A fiberoptic two-colour pyrometric technique for fuel particle temperature and size measurement is modified and applied to three pressurized reactors of different type in Finland, Germany and France. A modification of the pyrometric method for simultaneous in situ measurement of the temperature and size of individual pulverized coal particles at the pressurized entrained flow reactor in Jyvaeskylae was developed and several series of measurements were made. In Orleans a fiberoptic pyrometric device was installed to a pressurised thermogravimetric reactor and the two-colour temperatures of fuel samples were measured. Some results of these measurements are presented. The project belongs to EU`s Joule 2 extension research programme. (author)

Preliminary investigation of the use of monolithic U-Mo fuel in the MIT reactor

International Nuclear Information System (INIS)

Newton, Thomas H. Jr.; Kazimi, Mujid S.; Pilat, Edward E.; Xu Zhiwen

2003-01-01

Studies have begun on the use of monolithic LEU U-Mo fuel in the MIT Nuclear Research Reactor (MITR-II) using the Monte Carlo Transport code MCNP. These studies have included model benchmarking, LEU fuel optimization, burnup evaluation, in-core facility design, and determination of safety attributes. Benchmarking studies on the initial core have shown favorable agreement between the calculated and measured reactivity worths of the six control blades. In addition, optimization studies on LEU U7Mo MITR-II fuel have shown that an arrangement of ten to twelve plates per fuel element would have initial reactivity values and thermal neutron fluxes comparable to the current HEU core. Burnup studies which have been made using the MCODE depletion program will be presented. Safety attributes such as temperature coefficients, shutdown margins, and coolant subcooled margin are under evaluation. (author)

The Consolidated Net Worth of Private Colleges. Recommendation of a Model.

Science.gov (United States)

Jenny, Hans H.

One of several essential tools for assessing how the financial health of educational institutions is evolving is the Consolidated Net Worth Statement. This essay explores various aspects of conventional "funds" balance sheets and compares them with the Consolidated Net Worth. Emphasis is placed on how the Consolidated Net Worth Statement…

Measurements of the Reactivity Properties of the Aagesta Nuclear Power Reactor at Zero Power

Energy Technology Data Exchange (ETDEWEB)

Bernander, G

1967-07-15

The moderator level and temperature coefficients of reactivity and control rod differential reactivity worths have been determined at zero power by means of period measurements. The moderator level coefficient and the corresponding critical level have been measured for the 32, 68 and 136 fuel assembly cores at room temperature for cores with and without control rods. From these results the worths of control rods have been derived. HETERO calculations give up to 15 % lower values than the experimental results. The cold fresh core has an excess reactivity of 9.0 {+-} 0.2 %. The temperature coefficient and differential control rod worths were measured for the fully loaded core with filled tank in the temperature range between 30 and 210 deg C. Critical positions as a function of temperature were obtained for the corresponding control rod groups. No relevant calculations of the temperature coefficient for comparison with the experimental values have yet been made, but the experimental results together with measured critical control rod positions give good opportunities to check calculational programs. HETERO has been shown in these cases to reproduce differential control rod worths and critical positions fairly well. However, a certain underestimation of the rod effectiveness is quite noticeable. The relative increase in control rod effectiveness with a temperature change from 20 to 220 deg C has been estimated to be 0.29 {+-} 0.06.

It's All About the Money (For Some): Consequences of Financially Contingent Self-Worth.

Science.gov (United States)

Park, Lora E; Ward, Deborah E; Naragon-Gainey, Kristin

2017-05-01

Financial success is an important goal, yet striving for it is often associated with negative outcomes. One reason for this paradox is that financial pressures may be tied to basing self-worth on financial success. Studies 1a to 1c developed a measure of Financial Contingency of Self-Worth (Financial CSW), and found that it predicted more financial social comparisons, financial hassles, stress, anxiety, and less autonomy. In response to a financial (vs. academic) threat, higher Financial CSW participants experienced less autonomy, perceived financial problems more negatively, and disengaged from their financial problems (Study 2). When given an opportunity to self-affirm, however, Financial CSW participants did not show diminished autonomy in response to a financial (vs. academic) threat (Study 3). Finally, participants with higher Financial CSW were less likely to make extravagant spending decisions following a financial (vs. health) threat (Study 4). Together, these studies demonstrate the many consequences of staking self-worth on financial success.

Evaluation of Tehran research reactor (TRR) control rod worth using MCNP4C computer code

International Nuclear Information System (INIS)

Hosseini, Mohammad; Vosoughi, Naser; Hosseini, Seyed Abolfazl

2010-01-01

The main objective of reactor control system is to provide a safe reactor starting up, operation and shutting down. Calculation or measurement of precise values of control rod worth is of great importance in Tehran Research Reactor (TRR), considering the fact that they are the only controlling tools in the reactor. In present paper, simulation of TRR in First Operation Cycle (FOC) and in cold and clean core for the calculation of total and integral worth of control nods is reported. MCNP4C computer code has been used for all simulation process. Two method have been used for control rods worth calculation in this paper, namely the direct approach and perturbation method. It is shown that while the direct approach is appropriate for worth calculation of both the shim and the regulating control rods, the perturbation method is just suitable for tiny reactivity changes, i.e. for small initial part of regulating rods. Results of simulation are compared with the reported data in Safety Analysis Report (SAR) of Tehran research reactor and showed satisfactory agreement. (author)

Results of modeling advanced BWR fuel designs using CASMO-4

International Nuclear Information System (INIS)

Knott, D.; Edenius, M.

1996-01-01

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

29 CFR 4062.4 - Determinations of net worth and collective net worth.

Science.gov (United States)

2010-07-01

... financial condition, and business history. (6) The economic outlook for the person's industry and the market... do not produce income for the business being valued or are not used in the business. (c) Factors for... to sell, or offer to purchase or sell the business of the person made on or about the net worth...

Fuel assemblies for use in nuclear reactors

International Nuclear Information System (INIS)

Mochida, Takaaki.

1987-01-01

Purpose: To increase the plutonium utilization amount and improve the uranium-saving effect in the fuel assemblies of PWR type reactor using mixed uranium-plutonium oxides. Constitution: MOX fuel rods comprising mixed plutonium-uranium oxides are disposed to the outer circumference of a fuel assembly and uranium fuel rods only composed of uranium oxides are disposed to the central portion thereof. In such a fuel assembly, since the uranium fuel rods are present at the periphery of the control rod, the control rod worth is the same as that of the uranium fuel assembly in the prior art. Further, since about 25 % of the entire fuel rods is composed of the MOX fuel rods, the plutonium utilization amount is increased. Further, since the MOX fuel rods at low enrichment degree are present at the outer circumferential portion, mismatching at the boundary to the adjacent MOX fuel assembly is reduced and the problem of local power peaking increase in the MOX fuel assembly is neither present. (Kamimura, M.)

Self-worth therapy for depressive symptoms in older nursing home residents.

Science.gov (United States)

Tsai, Yun-Fang; Wong, Thomas K S; Tsai, Hsiu-Hsin; Ku, Yan-Chiou

2008-12-01

The aim of this study is to report the effects of self-worth therapy on depressive symptoms of older nursing home residents. Depression in older people has become a serious healthcare issue worldwide. Pharmacological and non-pharmacological therapies have been shown to have inconsistent effects, and drug treatment can have important side-effects. A quasi-experimental design was used. Older people were sampled by convenience from residents of a nursing home in northern Taiwan between 2005 and 2006. To be included in the study participants had to: (i) have no severe cognitive deficits; (ii) test positive for depressive status and (iii) take the same anti-depressant medication in the previous 3 months and throughout the study. Participants in the experimental group (n = 31) received 30 minutes of one-to-one self-worth therapy on 1 day a week for 4 weeks. Control group participants (n = 32) received no therapy, but were individually visited by the same research assistant, who chatted with them for 30 minutes on 1 day/week for 4 weeks. Depressive status, cognitive status and functional status were measured at baseline, immediately after the intervention and 2 months later. Data were analysed by mean, standard deviations, t-test, chi-squared test and univariate anova. Self-worth therapy immediately decreased depressive symptoms relative to baseline, but not relative to control treatment. However, 2 months later, depressive symptoms were statistically significantly reduced relative to control. Self-worth therapy is an easily-administered, effective, non-pharmacological treatment with potential for decreasing depressive symptoms in older nursing home residents.

gamma-ray spectra measurements for long cooled MOX spent fuels

International Nuclear Information System (INIS)

Murakami, Kiyonobu; Kobayashi, Iwao

1993-09-01

Gamma-ray spectra of spent fuels have important informations in the estimation of burnup rate, concentration of fission products, cooling time and etc. which are required in the fuel loading control of reactors and special nuclear materials accountancy from the view point of safe guard. Although, some available data are given about uranium dioxide fuels, few data are given about uranium and plutonium dioxide mixtures (MOX fuels). Especially, there is few data about MOX fuels which are irradiated in thermal reactors and cooled more than ten years. Gamma-ray spectra are measured for PuO 2 -UO 2 fuel rods (IFA-159, IFA-160) which are irradiated at HBWR in Norway up to 9,420 and 5,340MWd/t respectively. Gamma-ray spectra had been measured about the two fuels ten years ago at the spent fuel pond of Japan Demonstration Reactor (JPDR). The objectives of this measurement is to know how decayed the gamma-ray spectra in these ten years and some fission products are there which are effective to estimate burnup rate of spent MOX fuels. (author)

Simulation of TRIGA Mark II Benchmark Experiment using WIMSD4 and CITATION codes

International Nuclear Information System (INIS)

Dalle, Hugo Moura; Pereira, Claubia

2000-01-01

This paper presents a simulation of the TRIGA Mark II Benchmark Experiment, Part I: Steady-State Operation and is part of the calculation methodology validation developed to the neutronic calculation of the CDTN's TRIGA IPR - R1 reactor. A version of the WIMSD4, obtained in the Centro de Tecnologia Nuclear, in Cuba, was used in the cells calculation. In the core calculations was adopted the diffusion code CITATION. Was adopted a 3D representation of the core and the calculations were carried out at two energy groups. Many of the experiments were simulated, including, K eff , control rods reactivity worth, fuel elements reactivity worth distribution and the fuel temperature reactivity coefficient. The comparison of the obtained results, with the experimental results, shows differences in the range of the accuracy of the measurements, to the control rods worth and fuel temperature reactivity coefficient, or on an acceptable range, following the literature, to the K eff and fuel elements reactivity worth distribution and the fuel temperature reactivity coefficient. The comparison of the obtained results, with the experimental. results, shows differences in the range of the accuracy of the measurements, to the control rods worth and fuel temperature reactivity coefficient, or in an acceptable range, following the literature, to the K eff and fuel elements reactivity worth distribution. (author)

Self-worth, perceived competence, and behaviour problems in children with cerebral palsy

NARCIS (Netherlands)

Schuengel, C.; Voorman, J.; Stolk, J.; Dallmeijer, A.J.; Vermeer, A; Becher, J.

2006-01-01

Purpose. To examine the relevance of physical disabilities for self-worth and perceived competence in children with cerebral palsy (CP), and to examine associations between behaviour problems and self-worth and perceived competence. Methods. The Harter scales for self-worth and perceived competence

Oxide thickness measurement for monitoring fuel performance at high burnup

International Nuclear Information System (INIS)

Jaeger, M.A.; Van Swam, L.F.P.; Brueck-Neufeld, K.

1991-01-01

For on-site monitoring of the fuel performance at high burnup, Advanced Nuclear Fuels uses the linear scan eddy current method to determine the oxide thickness of irradiated Zircaloy fuel cans. Direct digital data acquisition methods are employed to collect the data on magnetic storage media. This field-proven methodology allows oxide thickness measurements and rapid interpretation of the data during the reactor outages and makes it possible to immediately reinsert the assemblies for the next operating cycle. The accuracy of the poolside measurements and data acquisition/interpretation techniques have been verified through hot cell metallographic measurements of rods previously measured in the fuel pool. The accumulated data provide a valuable database against which oxide growth models have been benchmarked and allow for effective monitoring of fuel performance. (orig.) [de

Manufacturing method for fuel assembly

International Nuclear Information System (INIS)

Yamaguchi, Takashi.

1997-01-01

In an FBR type reactor, uranium/plutonium mixed oxide fuels (MOX fuels) are used. Nuclear fuel materials containing uranium and plutonium are filled to a portion or all of a plurality of fuel rods. In this case, an equivalent fissile coefficient (B) based on a combustion guarantee method defined by the formula: (B) = (M) · (F) is determined. (M) is a combustion matrix constituted based on the solution of equation of combustion which is a differential equation representing change with time of each of nuclear fuel materials during combustion. (F) is an equivalent fissile coefficient based on a reactivity keeping method which is a coefficient representing a reactivity worth equivalent with plutonium-239. The content of each of the nuclear fuel materials is determined so that the effective multiplication factor at the final stage of the operation cycle is substantially constant by using the equivalent fissile coefficient (B) based on the combustion guarantee method. (I.N.)

Social Support and Neighborhood Stressors Among African American Youth: Networks and Relations to Self-Worth.

Science.gov (United States)

McMahon, Susan D; Felix, Erika D; Nagarajan, Thara

2011-06-01

Although neighborhood stressors have a negative impact on youth, and social support can play a protective role, it is unclear what types and sources of social support may contribute to positive outcomes among at-risk youth. We examined the influences of neighborhood disadvantage and social support on global self-worth among low-income, urban African American youth, both concurrently and longitudinally. We examined social support from both a structural and functional perspective, and tested the main-effects and the stress-buffering models of social support. Participants included 82-130 youth, in 6th-8th grade, who completed self-report measures. Network support results suggest participants received emotional, tangible, and informational support most often from mothers and other female relatives, with friends, fathers, and teachers also playing important roles. Model testing accounted for neighborhood stressors and support from various sources, revealing support from close friends was associated with concurrent self-worth; whereas, parent support predicted self-worth longitudinally, above and beyond initial levels of self-worth. The findings provide evidence for the main-effects model of social support and not the stress-buffering model. Our findings illustrate the importance of extended family networks and the types of support that youth rely upon in African American impoverished communities, as well as how support contributes to global self-worth. Implications and suggestions for future research and intervention are discussed.

In-core fuel element temperature and flow measurment of HFETR

International Nuclear Information System (INIS)

Chen Daolong; Jiang Pei

1988-02-01

The HFETR in-core fuel element temperature-flow measurement facility and its measurement system are expounded. The applications of the instrumented fuel element to stationary and transient states measurements during the lift of power, the operation test of all lifetime at first load, and the deepening burn-up test at second load are described. The method of determination of the hot point temperature under the fin is discussed. The error analysis is made. The fuel element out-of-pile water deprivation test is described. The development of this measurement facility and succesful application have made important contribution to high power and deep burn-up safe operation at two load, in-core fuel element irradiation, and varied investigation of HFETR. After operation at two loads, the integrated power of this instrumented fuel element arrives at 90.88 MWd, its maximum point burn-up is about 64.9%, so that the economy of fuel use of HFETR is raised very much

Mental Health and Self-Worth in Socially Transitioned Transgender Youth.

Science.gov (United States)

Durwood, Lily; McLaughlin, Katie A; Olson, Kristina R

2017-02-01

Social transitions are increasingly common for transgender children. A social transition involves a child presenting to other people as a member of the "opposite" gender in all contexts (e.g., wearing clothes and using pronouns of that gender). Little is known about the well-being of socially transitioned transgender children. This study examined self-reported depression, anxiety, and self-worth in socially transitioned transgender children compared with 2 control groups: age- and gender-matched controls and siblings of transgender children. As part of a longitudinal study (TransYouth Project), children (9-14 years old) and their parents completed measurements of depression and anxiety (n = 63 transgender children, n = 63 controls, n = 38 siblings). Children (6-14 years old; n = 116 transgender children, n = 122 controls, n = 72 siblings) also reported on their self-worth. Mental health and self-worth were compared across groups. Transgender children reported depression and self-worth that did not differ from their matched-control or sibling peers (p = .311), and they reported marginally higher anxiety (p = .076). Compared with national averages, transgender children showed typical rates of depression (p = .290) and marginally higher rates of anxiety (p = .096). Parents similarly reported that their transgender children experienced more anxiety than children in the control groups (p = .002) and rated their transgender children as having equivalent levels of depression (p = .728). These findings are in striking contrast to previous work with gender-nonconforming children who had not socially transitioned, which found very high rates of depression and anxiety. These findings lessen concerns from previous work that parents of socially transitioned children could be systematically underreporting mental health problems. Copyright © 2016 American Academy of Child and Adolescent Psychiatry. Published by Elsevier Inc. All rights reserved.

Contact-type displacement measuring mechanism for fuel assembly in reactor

International Nuclear Information System (INIS)

Yokota, Yoshio; Ko, Kuniaki.

1995-01-01

The measuring mechanism of the present invention, which is used in a lmfbr type reactor, is suspended by a gripper of a fuel handing machine, and it comprises a combination of a displacement amount measuring jig allowed to be inserted into a handling head of a fuel assembly and a displacement amount measuring ring disposed at the lower portion in the handling head. The displacement amount measuring jig has a structure comprising a releasable handle and a columnar or cylindrical measuring portion allowable to be inserted into the handling head formed at the lower portion of the handle, which are connected with each other. When an interference (contact) occurred between the displacement amount measuring jig and the stepwise displacement amount measuring ring during the measurement, change of load and a phenomenon that the fuel handing machine can not be lowered are recognized, so that core displacement amount can be recognized based on the stroke of the gripper portion. Then, remote measurement is possible for displacement and deformation of the fuel assembly in the reactor container, and the measurement can be conducted by the same procedures and in the same period of time as in a case of ordinary fuel exchange operation. A flow channel for coolants passing through the fuel assembly can be ensured, thereby enabling to measure the amount of core displacement which is closer to an actual value in the reactor. (N.H.)

Fuel element burnup measurements for the equilibrium LEU silicide RSG GAS (MPR-30) core under a new fuel management strategy

International Nuclear Information System (INIS)

Pinem, Surian; Liem, Peng Hong; Sembiring, Tagor Malem; Surbakti, Tukiran

2016-01-01

Highlights: • Burnup measurement of fuel elements comprising the new equilibrium LEU silicide core of RSG GAS. • The burnup measurement method is based on a linear relationship between reactivity and burnup. • Burnup verification was conducted using an in-house, in-core fuel management code BATAN-FUEL. • A good agreement between the measured and calculated burnup was confirmed. • The new fuel management strategy was confirmed and validated. - Abstract: After the equilibrium LEU silicide core of RSG GAS was achieved, there was a strong need to validate the new fuel management strategy by measuring burnup of fuel elements comprising the core. Since the regulatory body had a great concern on the safety limit of the silicide fuel element burnup, amongst the 35 burnt fuel elements we selected 22 fuel elements with high burnup classes i.e. from 20 to 53% loss of U-235 (declared values) for the present measurements. The burnup measurement method was based on a linear relationship between reactivity and burnup where the measurements were conducted under subcritical conditions using two fission counters of the reactor startup channel. The measurement results were compared with the declared burnup evaluated by an in-house in-core fuel management code, BATAN-FUEL. A good agreement between the measured burnup values and the calculated ones was found within 8% uncertainties. Possible major sources of differences were identified, i.e. large statistical errors (i.e. low fission counters’ count rates), variation of initial U-235 loading per fuel element and accuracy of control rod indicators. The measured burnup of the 22 fuel elements provided the confirmation of the core burnup distribution planned for the equilibrium LEU silicide core under the new fuel management strategy.

Reducing contingent self-worth: a defensive response to self-threats.

Science.gov (United States)

Buckingham, Justin; Lam, Tiffany A; Andrade, Fernanda C; Boring, Brandon L; Emery, Danielle

2018-04-10

Previous research shows that people with high self-esteem cope with threats to the self by reducing the extent to which their self-worth is contingent on the threatened domain (Buckingham, Weber, & Sypher, 2012). The present studies tested the hypothesis that this is a defensive process. In support of this hypothesis, Study 1 (N = 160), showed that self-affirmation attenuates the tendency for people with high self-esteem to reduce their contingencies of self-worth following self-threat. Furthermore, Study 2 (N = 286), showed that this tendency was more prevalent among people with defensive self-esteem than among those with secure self-esteem. The present studies imply that reducing contingent self-worth after self-threat is a defensive process. We discuss implications for theories of contingent self-worth.

Importance measures

International Nuclear Information System (INIS)

Gomez Cobo, A.

1997-01-01

The presentation discusses the following: general concepts of importance measures; example fault tree, used to illustrate importance measures; Birnbaum's structural importance; criticality importance; Fussel-Vesely importance; upgrading function; risk achievement worth; risk reduction worth

Self-protection of FLIP fuel: Experimental measurements

International Nuclear Information System (INIS)

Johnson, A.G.; Ringle, J.C.; Anderson, T.V.; Dodd, B.; Bennett, S.; Carpenter, W.T.

1980-01-01

During the last year and one-half non-power reactor licensees of the Nuclear Regulatory Commission (NRC) have been repeatedly alerted to the fact that the Commission planned to tighten physical security regulations for their class of reactors. Essentially all of the new NRC physical security requirements for these reactors have been based on the type, quantity and enrichment of the special nuclear materials (e.g. enriched uranium and plutonium) possessed by the licensees. While each licensee's security classification will be evaluated individually, it is now apparent that non-power reactor licensees using or storing formula quantities of special nuclear material not meeting the 100 rem per hour at three feet self-protection exemption (e.g. 5 kilograms or more of non-self-protected fuel enriched to 20% or more with U-235) will be required to meet extremely stringent NRC physical security requirements patterned after the new Physical Protection Upgrade Rule. Implementation of these much stricter security requirements poses many economic and operational difficulties for university research reactors, to the point where facility closure could be the only alternative. TRIGA reactors utilizing a full FLIP fueled core, such as the Oregon State University TRIGA (OSTR), qualify for the highest physical protection category unless fuel can be maintained at self-protecting radiation levels. In order to demonstrate that OSTR fuel could be consistently kept above the 100 rem per hour threshold, a computer program was written which predicts the gamma radiation levels from an irradiated FLIP fuel element at 1, 2 and 3 feet in air and water. Furthermore, in order to verify the accuracy of the computer program, actual measurements of irradiated fuel elements were made at 3 feet in air and at 1, 2 and 3 feet in water, and the results compared very favorably to the predictions. The results of specific measurements, the instrumentation used and its calibration, the personnel doses

An overview of economic and technical issues related to LWR MOX fuel usage

International Nuclear Information System (INIS)

Malone, J.P.; Varley, G.; Goldstein, L.

1999-01-01

This paper will present comparisons of the economics of MOX versus UO 2 fuels. In addition to the economics of the front end, the scope of the comparison will include the back end of the fuel cycle. Management of spent MOX fuel assemblies presents utilities with some technical issues that can complicate spent fuel pool operation. Alternative spent fuel management methods, such as dry storage of spent MOX fuel assemblies, will also be discussed. Differences in decay heat loads versus time for spent MOX and UO 2 fuel assemblies will be presented. This difference is one of the main problems confronting spent fuel managers relative to MOX. The difference in decay heat loads will serve as the basis for a performance overview of the various spent fuel technologies available today. The economics of the front end of MOX will be presented relative to UO 2 fuel. Availability of MOX manufacturing capability will also be discussed, along with a discussion of its impact on future MOX fabrication prices. The in-core performance of MOX will be compared to that of UO 2 fuel with similar performance characteristics. The information will include highlights of nuclear design and related operational considerations such as: Reactivity reduction with burnup is slower for MOX fuel than for UO 2 fuel; Spectral hardening resulting in lower control rod worths and a lower soluble boron worth; and more negative moderator, void and fuel temperature coefficients. A comparison of Westinghouse and ABB-CE core designs for use on disposition of weapons MOX in 12- and 18-month cycles will be presented. (author)

Judgement on the data for fuel assembly outlet temperatures of WWER fuel assemblies in power reactors based on measurements with experimental fuel assemblies

International Nuclear Information System (INIS)

Krause, F.

1986-01-01

In the period from 1980 to 1985, in the Rheinsberg nuclear power plant experimental fuel assemblies were used on lattices at the periphery of the core. These particular fuel assemblies dispose of an extensive in-core instrumentation with different sensors. Besides this, they are fit out with a device to systematically thottle the coolant flow. The large power gradient present at the core position of the experimental fuel assembly causes a temperature profile along the fuel assemblies which is well provable at the measuring points of the outlet temperature. Along the direction of flow this temperature profile in the coolant degrades only slowly. This effect is to be taken into account when measuring the fuel assembly outlet temperature of WWER fuel assemblies. Besides this, the results of the measurements hinted both at a γ-heating of the temperature measuring points and at tolerances in the calculation of the micro power density distribution. (author)

Calculation code of heterogeneity effects for analysis of small sample reactivity worth

International Nuclear Information System (INIS)

Okajima, Shigeaki; Mukaiyama, Takehiko; Maeda, Akio.

1988-03-01

The discrepancy between experimental and calculated central reactivity worths has been one of the most significant interests for the analysis of fast reactor critical experiment. Two effects have been pointed out so as to be taken into account in the calculation as the possible cause of the discrepancy; one is the local heterogeneity effect which is associated with the measurement geometry, the other is the heterogeneity effect on the distribution of the intracell adjoint flux. In order to evaluate these effects in the analysis of FCA actinide sample reactivity worth the calculation code based on the collision probability method was developed. The code can handle the sample size effect which is one of the local heterogeneity effects and also the intracell adjoint heterogeneity effect. (author)

Small-sample-worth perturbation methods

International Nuclear Information System (INIS)

1985-01-01

It has been assumed that the perturbed region, R/sub p/, is large enough so that: (1) even without a great deal of biasing there is a substantial probability that an average source-neutron will enter it; and (2) once having entered, the neutron is likely to make several collisions in R/sub p/ during its lifetime. Unfortunately neither assumption is valid for the typical configurations one encounters in small-sample-worth experiments. In such experiments one measures the reactivity change which is induced when a very small void in a critical assembly is filled with a sample of some test-material. Only a minute fraction of the fission-source neutrons ever gets into the sample and, of those neutrons that do, most emerge uncollided. Monte Carlo small-sample perturbations computations are described

Accuracy of fuel motion measurements using in-core detectors

International Nuclear Information System (INIS)

Dupree, S.A.

1975-01-01

An initial assessment has been made as to how accurately fuel motion can be measured with in-core detectors. A portion of this assessment has involved the calculation of the response of various detectors to fuel motion and the development of a formalism for correlating uncertainties in a neutron flux measurement to uncertainties in the fuel motion. Initially, four idealized configurations were studied in one dimension. These configurations consisted of (1) a single fuel-pin test using ACPR, (2) a seven fuel-pin test using ACPR, (3) a full subassembly (271 pin) test using a Class I ANL-type SAREF, and (4) a full subassembly plus six partial subassemblies (approximately 1000 pin) test using a Class III GE-type SAREF. It was assumed that melt would occur symmetrically at the center of the test fuel and that fuel would therefore disappear from the center of the geometry. For each case of series of calculations was performed in which detector responses were determined at several radial locations for the unperturbed core and for the core with various fractions of the fuel replaced with Na. This fuel loss was assumed to occur essentially instantaneously such that the power level in the remaining portion of the test fuel remained unchanged from that of the initial unperturbed condition

Achieving equal pay for comparable worth through arbitration.

Science.gov (United States)

Wisniewski, S C

1982-01-01

Traditional "women's jobs" often pay relatively low wages because of the effects of institutionalized stereotypes concerning women and their role in the work place. One way of dealing with sex discrimination that results in job segregation is to narrow the existing wage differential between "men's jobs" and "women's jobs." Where the jobs are dissimilar on their face, this narrowing of pay differences involves implementing the concept of "equal pay for jobs of comparable worth." Some time in the future, far-reaching, perhaps even industrywide, reductions in male-female pay differentials may be achieved by pursuing legal remedies based on equal pay for comparable worth. However, as the author demonstrates, immediate, albeit more limited, relief for sex-based pay inequities found in specific work places can be obtained by implementing equal pay for jobs of comparable worth through the collective bargaining and arbitration processes.

Concentration measurements of biodiesel in engine oil and in diesel fuel

International Nuclear Information System (INIS)

Mäder, A; Eskiner, M; Burger, C; Rossner, M; Krahl, J; Ruck, W

2012-01-01

This work comprised a method for concentration measurements of biodiesel in engine oil as well as biodiesel in diesel fuel by a measurement of the permittivity of the mixture at a frequency range from 100 Hz to 20 kHz. For this purpose a special designed measurement cell with high sensitivity was designed. The results for the concentration measurements of biodiesel in the engine oil and diesel fuel shows linearity to the measurement cell signal for the concentration of biodiesel in the engine oil between 0.5% Vol. to 10% Vol. and for biodiesel in the diesel fuel between 0% Vol. to 100% Vol. The method to measure the concentration of biodiesel in the engine oil or the concentration of biodiesel in the diesel fuel is very accurate and low concentration of about 0.5% Vol. biodiesel in engine oil or in diesel fuel can be measured with high accuracy.

Calculation analysis of TRIGA MARK II reactor core composed of two types of fuel elements

International Nuclear Information System (INIS)

Ravnik, M.

1988-11-01

The most important properties of mixed cores are treated for TRIGA MARK II reactor, composed of standard (20% enriched, 8.5w% U content) and FLIP (70% enriched, 8.5w% U content) fuel elements. Large difference in enrichment and presence of burnable poison in FLIP fuel have strong influence on the main core characteristics, such as: fuel temperature coefficient, power defect, Xe and Sm worth, power and flux distributions, etc. They are significantly different for both types of fuel. Optimal loading of mixed cores therefore strongly depends on the loading pattern of both types of fuel elements. Results of systematic calculational analysis of mixed cores are presented. Calculations on the level of fuel element are performed with WIMSD-4 computer code with extended cross-section library. Core calculations are performed with TRIGAP two-group 1-D diffusion code. Results are compared to measurements and physical explanation is provided. Special concern is devoted to realistic mixed cores, for which optimal in-core fuel management is derived. Refs, figs and tabs

Childhood Predictors of Adolescent Competence and Self-Worth in Rural Youth

Science.gov (United States)

Rew, Lynn; Grady, Matthew W.; Spoden, Micajah

2012-01-01

Problem Urban children who become competent adults despite circumstances that place their development and mental health at risk are considered to be resilient. Less is known about the risk and protective factors that characterize resilience among Hispanic/Latinos living in rural areas. Methods Data for regression analyses were collected when children (N = 603; 54% Hispanic/Latino) enrolled in the study in fifth grade, (M=10.4 years of age) and again five years later when they were in high school (M=15 years of age). Findings Statistically significant predictors of competence and self-worth in high schoolers included gender, ethnicity, and mother’s education, as well as stress, temperament (task persistence), and competences measured in grade school. Conclusions Parents’ perceptions of child’s temperament is a significant predictor of future competence and self-worth among rural adolescents. PMID:23121139

Childhood predictors of adolescent competence and self-worth in rural youth.

Science.gov (United States)

Rew, Lynn; Grady, Matthew W; Spoden, Micajah

2012-11-01

Urban children who become competent adults despite circumstances that place their development and mental health at risk are considered to be resilient. Less is known about the risk and protective factors that characterize resilience among Hispanic/Latinos living in rural areas. Data for regression analyses were collected when children (n = 603; 54% Hispanic/Latino) enrolled in the study in fifth grade (M = 10.4 years of age), and again 5 years later when they were in high school (M = 15 years of age). Statistically significant predictors of competence and self-worth in high schoolers included gender, ethnicity, and mother's education, as well as stress, temperament (task persistence), and competences measured in grade school. Parents' perception of child's temperament is a significant predictor of future competence and self-worth among rural adolescents. © 2012 Wiley Periodicals, Inc.

Void worths in subcritical cores cooled by lead-bismuth

International Nuclear Information System (INIS)

Wallenius, Janne; Tucek, Kamil; Gudowski, Waclaw

2001-01-01

The introduction lead-bismuth coolant in accelerator driven transmutation systems (ADS) was: good neutron economy (higher source efficiency); natural circulation possible (decay heat removal); synergy with spallation target (simplified coolant management); high temperature of boiling (larger overpower margin); smaller void worths (operation at higher k-values). This paper deals with different aspects of the void worths in JAERI ADS

Fuel Rod Vibration Measurement Method using a Flap and its Verification

Energy Technology Data Exchange (ETDEWEB)

Yoo, Joo Young; Park, Nam Gyu; Suh, Jung Min; Jeon, Kyeong Lak [KEPCO NF Co., Daejeon (Korea, Republic of)

2011-10-15

Flow-induced vibration is a critical factor for the mechanical integrity of a fuel rod. This vibration can cause leaked fuel through the mechanism, such as grid to rod fretting. To minimize the failures caused by flow-induced vibration, a robust design is needed which takes into account vibrational characteristics. That is, the spacer grid design should be developed to avoid any excessive vibration. On the one hand, if fuel rod vibration can be measured, an estimation of the excitation forces, which are a critical cause of rod failure, should be possible. Therefore, by applying an external force, flow-induced vibration can be roughly estimated when the fuel rod vibration model is used. KEPCO Nuclear Fuel developed the test loop to research flow-induced vibration as shown in Fig.1. The investigation flow-induced vibration (INFINIT) - the test facility - can measure the grid strap vibration and pressure drop of a 5x5 small scale fuel bundle. Basically, using a Laser Doppler Vibrometer (LDV), the vibration of a structure immersed in high speed fluid can be measured. Grid strap vibration is easily measured using an LDV. However, it is quite difficult to measure fuel rod vibration because of the round surface shape of the rods. In addition, measuring current method using the LDV, it was only possible to directly measure fuel rod vibration at the first row of the bundle as the rods behind the first row are obscured. To solve this problem, a thin flap, as shown in Fig. 2(a) can be used as a reflecting target, gaining access to rods within the bundle. The flap is attached to the fuel rod, as in Fig. 2(b). As a result, most of the inner rod vibration can be measured. Before using a flap to measure fuel rod vibration, a verification process was needed to show whether the LDV signal from the flap vibration provided equivalent and reliable signals. Therefore, impact testing was carried out on the fuel rod using a flap. The LDV signals were then compared with accelerometer

Burn-up measurements of spent fuel using gamma spectrometry technique

International Nuclear Information System (INIS)

Pereda, C.; Henriquez, C.; Klein, J.; Medel, J.

2005-01-01

Burn-up results obtained for HEU (45% of 235 U) fuel assemblies of the RECH-1 Research Reactor using gamma spectrometry technique are presented. The spectra were got from an in-pool facility built in the reactor to be mainly used to measure the burnup of irradiated fuel assemblies with short cooling time, where 95 Zr is being evaluated as possible fission monitor. A program to measure all spent fuel assemblies of the RECH-1 reactor was initiated in the frame of the Regional Project RLA/4/018: 'Management of Spent Fuel from Research Reactors'. The results presented here were obtained from HEU spent fuel assemblies with cooling time greater than 100 days and 137 Cs was used as fission monitor. The efficiency of the in-pool system was determined using a slightly burnt experimental fuel assembly, which has one fuel plate (one of the outer plates) and the rest are dummy plates. An average burn-up of 2.8% of 235 U was previously measured for the experimental fuel assembly utilizing a facility installed in a hot cell and 137 Cs was used as monitor. (author)

A shielded measurement system for irradiated nuclear fuel measurements

International Nuclear Information System (INIS)

Mosby, W.R.; Aumeier, S.E.; Klann, R.T.

1999-01-01

The US Department of Energy (DOE) is driving a transition toward dry storage of irradiated nuclear fuel (INF), toward characterization of INF for final disposition, and toward resumption of measurement-based material control and accountability (MC and A) efforts for INF. For these reasons, the ability to efficiently acquire radiological measurements of INF in a dry environment is important. The DOE has recently developed a guidance document proposing MC and A requirements for INF. The intent of this document is to encourage the direct measurement of INF on inventory within DOE. The guidance document reinforces and clarifies existing material safeguards requirements as they pertain to INF. Validation of nuclear material contents of non-self-protecting INF must be accomplished by direct measurement, application of validated burnup codes using qualified initial fissile content, burnup data, and age or by other valid means. The fuel units must remain intact with readable identification numbers. INF may be subject to periodic inventories with visual item accountability checks. Quantitative measurements may provide greater assurance of the integrity of INF inventories at a lower cost and with less personnel exposure than visual item accountability checks. Currently, several different approaches are used to measure the radiological attributes of INF. Although these systems are useful for a wide variety of applications, there is currently no relatively inexpensive measurement system that is readily deployable for INF measurements for materials located in dry storage. The authors present the conceptual design of a shielded measurement system (SMS) that could be used for this purpose. The SMS consists of a shielded enclosure designed to house a collection of measurement systems to allow measurements on spent fuel outside of a hot cell. The phase 1 SMS will contain 3 He detectors and ionization chambers to allow for gross neutron and gamma-ray measurements. The phase 2 SMS

Improved Monte Carlo-perturbation method for estimation of control rod worths in a research reactor

International Nuclear Information System (INIS)

Kalcheva, Silva; Koonen, Edgar

2009-01-01

A hybrid method dedicated to improve the experimental technique for estimation of control rod worths in a research reactor is presented. The method uses a combination of Monte Carlo technique and perturbation theory. Perturbation method is used to obtain the equation for the relative efficiency of control rod insertion. A series of coefficients, describing the axial absorption profile are used to correct the equation for a composite rod, having a complicated burn-up irradiation history. These coefficients have to be determined - by experiment or by using some theoretical/numerical method. In the present paper they are derived from the macroscopic absorption cross-sections, obtained from detailed Monte Carlo calculations by MCNPX 2.6.F of the axial burn-up profile during control rod life. The method is validated on measurements of control rod worths at the BR2 reactor. Comparison with direct MCNPX evaluations of control rod worths is also presented

Fuel management in CANDU reactors: Daniel Rozon's contribution

International Nuclear Information System (INIS)

Rozon, D.; Varin, E.; Chambon, R.

2010-01-01

The CANDU fuel management optimization problem is in many ways different from LWRs fuel management, because of the on-line refueling and the complete 3-D geometry problem. Daniel Rozon was an outstanding leader in the understanding and resolution of this optimization problem and remained during his entire career. Daniel Rozon and his students have used the generalized adjoint formalism implemented in standard mathematical programming methods to solve the optimization of the exit burnup in the reactor as well as the optimization of control rod worth or fuel enrichment. We have summarized here the theoretical basis of fuel management and resolution methods, the latest approaches of optimization and results as obtained using the OPTEX code. (author)

The thorium fuel cycle

International Nuclear Information System (INIS)

Merz, E.R.

1977-01-01

The utilization of the thorium fuel cycle has long since been considered attractive owing to the excellent neutronic characteristics of 233 U, and the widespread and cheap thorium resources. Rapidly increasing uranium prices, public reluctance for widespread Pu recycling and expected delays for the market penetration of fast breeders have led to a reconsideration of the thorium fuel cycle merits. In addition, problems associated with reprocessing and waste handling, particularly with re-fabrication by remote handling of 233 U, are certainly not appreciably more difficult than for Pu recycling. To divert from uranium as a nuclear energy source it seems worth while intensifying future efforts for closing the Th/ 233 U fuel cycle. HTGRs are particularly promising for economic application. However, further research and development activities should not concentrate on this reactor type alone. Light- and heavy-water-moderated reactors, and even future fast breeders, may just as well take advantage of a demonstrated thorium fuel cycle. (author)

Use of zero power plutonium reactor measurements as a support of criticality prediction for the SNR-300

International Nuclear Information System (INIS)

Pilate, S.; de Wouters, R.; Wehmann, U.; Helm, F.; Scholtyssek, W.

1978-01-01

Evaluations of criticality measurements performed in various SNEAK and Zero Power Plutonium Reactor (ZPPR) cores are compared. The best available methods of calculations (including transport theory) are used. The ZPPR results support well the trend indicated by the SNEAK evaluations for clean cores and for cores with followers; for cores with absorbers partially inserted, the agreement is only rough. Evaluations of control rod worth measurements are therefore also compared, using the routine method of calculation for SNR-300 (diffusion theory). The control rod worths are largely underestimated in SNEAK (C/E = 0.89), but only slightly underestimated in the ZPPR (C/E = 0.97). The difference in the nature of core fuel (uranium in SNEAK, plutonium in the ZPPR) could be at the origin of this discrepancy

Extending Data Worth Analyses to Select Multiple Observations Targeting Multiple Forecasts

DEFF Research Database (Denmark)

Vilhelmsen, Troels Norvin; Ferre, Ty Paul

2017-01-01

. In the present study, we extend previous data worth analyses to include: simultaneous selection of multiple new measurements and consideration of multiple forecasts of interest. We show how the suggested approach can be used to optimize data collection. This can be used in a manner that suggests specific...... measurement sets or that produces probability maps indicating areas likely to be informative for specific forecasts. Moreover, we provide examples documenting that sequential measurement election approaches often lead to suboptimal designs and that estimates of data covariance should be included when...

Gamma densitometer for measuring Pu density in fuel tubes

International Nuclear Information System (INIS)

Winn, W.G.

1982-01-01

A fuel-gamma-densitometer (FGD) has been developed to examine nondestructively the uniformity of plutonium in aluminum-clad fuel tubes at the Savannah River Plant (SRP). The monitoring technique is γ-ray spectroscopy with a lead-collimated Ge(Li) detector. Plutonium density is correlated with the measured intensity of the 208 keV γ-ray from 237 U (7d) of the 241 Pu (15y) decay chain. The FGD measures the plutonium density within 0.125- or 0.25-inch-diameter areas of the 0.133- to 0.183-inch-thick tube walls. Each measurement yields a density ratio that relates the plutonium density of the measured area to the plutonium density in normal regions of the tube. The technique was used to appraise a series of fuel tubes to be irradated in an SRP reactor. High-density plutonium areas were initially identified by x-ray methods and then examined quantitatively with the FGD. The FGD reliably tested fuel tubes and yielded density ratios over a range of 0.0 to 2.5. FGD measurements examined (1) nonuniform plutonium densities or hot spots, (2) uniform high-density patches, and (3) plutonium density distribution in thin cladding regions. Measurements for tubes with known plutonium density agreed with predictions to within 2%. Attenuation measurements of the 208-keV γ-ray passage through the tube walls agreed to within 2 to 3% of calculated predictions. Collimator leakage measurements agreed with model calculations that predicted less than a 1.5% effect on plutonium density ratios. Finally, FGD measurements correlated well with x-ray transmission and fluoroscopic measurements. The data analysis for density ratios involved a small correction of about 10% for γ-shielding within the fuel tube. For hot spot examinations, limited information for this correction dictated a density ratio uncertainty of 3 to 5%

Precise measurement of fuel content of irradiated and nonirradiated materials

International Nuclear Information System (INIS)

Harker, Y.D.; Napper, P.R.; Proctor, A.E.

1984-01-01

This paper discusses the application of precise reactivity measurements in the Advanced Reactivity Measurement Facility at Idaho National Engineering Laboratory (INEL) to determine th fuel content in irradiated and nonirradiated materials. Different methods of reactivity measurements and examples of how they have been are presented, which provides an insight in capabilities available to analyze samples with different geometrical sizes from small volumes approx. 100 cc to 12 ft long fuel pins and also samples with different fuel content ranges from approx. 2 mg to approx. 600 g. The overall accuracy of these measurements is approx. 0.5% (1sigma)

Contingent self-worth moderates the relationship between school stressors and psychological stress responses.

Science.gov (United States)

Ishizu, Kenichiro

2017-04-01

This study examined the moderating role of contingent self-worth on the relationships between school stressors and psychological stress responses among Japanese adolescents. A total of 371 Japanese junior high school students (184 boys and 187 girls, M age  = 12.79 years, SD = 0.71) completed the Japanese version of the Self-Worth Contingency Questionnaire and a mental health checklist at two points separated by a two-month interval. Hierarchical multiple regression analyses were then used to determine whether contingent self-worth moderated the relationship between school stressors and psychological stress responses. The results indicated that, when psychological stress responses were controlled for at Time 1, contingent self-worth did not predict the psychological stress responses at Time 2. However, a two-way interaction between contingent self-worth and stressors was found to significantly influence psychological stress responses, thus indicating that stressors had a stronger impact on psychological stress responses among those with high contingent self-worth compared to those with low contingent self-worth. Copyright © 2017 The Foundation for Professionals in Services for Adolescents. Published by Elsevier Ltd. All rights reserved.

Fuel film thickness measurements using refractive index matching in a stratified-charge SI engine operated on E30 and alkylate fuels

Science.gov (United States)

Ding, Carl-Philipp; Sjöberg, Magnus; Vuilleumier, David; Reuss, David L.; He, Xu; Böhm, Benjamin

2018-03-01

This study shows fuel film measurements in a spark-ignited direct injection engine using refractive index matching (RIM). The RIM technique is applied to measure the fuel impingement of a high research octane number gasoline fuel with 30 vol% ethanol content at two intake pressures and coolant temperatures. Measurements are conducted for an alkylate fuel at one operating case, as well. It is shown that the fuel volume on the piston surface increases for lower intake pressure and lower coolant temperature and that the alkylate fuel shows very little spray impingement. The fuel films can be linked to increased soot emissions. A detailed description of the calibration technique is provided and measurement uncertainties are discussed. The dependency of the RIM signal on refractive index changes is measured. The RIM technique provides quantitative film thickness measurements up to 0.9 µm in this engine. For thicker films, semi-quantitative results of film thickness can be utilized to study the distribution of impinged fuel.

Method of performing shutdown reactivity measurements in spent nuclear fuel storage pools

International Nuclear Information System (INIS)

Levine, S.H.; Schultz, M.A.; Chang, D.

1981-01-01

The objective of this paper is to develop a device to measure the k/infinity/ of a spent fuel assembly used in light water reactors. A subcritical assembly having a cross configuration is designed to allow measurement of the k/sub //infinity/ of a spent fuel assembly by comparing the change in its multiplication with that of a fuel assembly of known k/infinity/. Calculations have been performed using nucleonic codes to develop polynomial equations that relate the k/infinity/ of the spent fuel assembly to measured data. The measurements involve taking count rates with the spent fuel assembly in the center position of the subcritical assembly, and the measured data are the count rate ratio of the spent fuel assembly over the count rate taken with a fuel assembly of known k/infinity/. The polynomial equations are easy to program on a microcomputer, which, together with the subcritical assembly, form the k/infinity/ meter. 9 refs

Computerized x-ray radiographic system for fuel pellet measurements

International Nuclear Information System (INIS)

Green, D.R.; Karnesky, R.A.; Bromley, C.

1977-01-01

The development and operation of a computerized system for determination of fuel pellet diameters from x-ray radiography is described. Actual fuel pellet diameter measurements made with the system are compared to micrometer measurements on the same pellets, and statistically evaluated. The advantages and limitations of the system are discussed, and recommendations are made for further development

Alternative Measuring Approaches in Gamma Scanning on Spent Nuclear Fuel

Energy Technology Data Exchange (ETDEWEB)

Sihm Kvenangen, Karen

2007-06-15

In the future, the demand for energy is predicted to grow and more countries plan to utilize nuclear energy as their source of electric energy. This gives rise to many important issues connected to nuclear energy, such as finding methods that can verify that the spent nuclear fuel has been handled safely and used in ordinary power producing cycles as stated by the operators. Gamma ray spectroscopy is one method used for identification and verification of spent nuclear fuel. In the specific gamma ray spectroscopy method called gamma scanning the gamma radiation from the fission products Cs-137, Cs-134 and Eu-154 are measured in a spent fuel assembly. From the results, conclusions can be drawn about the fuels characteristics. This degree project examines the possibilities of using alternative measuring approaches when using the gamma scanning method. The focus is on examining how to increase the quality of the measured data. How to decrease the measuring time as compared with the present measuring strategy, has also been investigated. The main part of the study comprises computer simulations of gamma scanning measurements. The simulations have been validated with actual measurements on spent nuclear fuel at the central interim storage, Clab. The results show that concerning the quality of the measuring data the conventional strategy is preferable, but with other starting positions and with a more optimized equipment. When focusing on the time aspect, the helical measuring strategy can be an option, but this needs further investigation.

Alternative Measuring Approaches in Gamma Scanning on Spent Nuclear Fuel

International Nuclear Information System (INIS)

Sihm Kvenangen, Karen

2007-06-01

In the future, the demand for energy is predicted to grow and more countries plan to utilize nuclear energy as their source of electric energy. This gives rise to many important issues connected to nuclear energy, such as finding methods that can verify that the spent nuclear fuel has been handled safely and used in ordinary power producing cycles as stated by the operators. Gamma ray spectroscopy is one method used for identification and verification of spent nuclear fuel. In the specific gamma ray spectroscopy method called gamma scanning the gamma radiation from the fission products Cs-137, Cs-134 and Eu-154 are measured in a spent fuel assembly. From the results, conclusions can be drawn about the fuels characteristics. This degree project examines the possibilities of using alternative measuring approaches when using the gamma scanning method. The focus is on examining how to increase the quality of the measured data. How to decrease the measuring time as compared with the present measuring strategy, has also been investigated. The main part of the study comprises computer simulations of gamma scanning measurements. The simulations have been validated with actual measurements on spent nuclear fuel at the central interim storage, Clab. The results show that concerning the quality of the measuring data the conventional strategy is preferable, but with other starting positions and with a more optimized equipment. When focusing on the time aspect, the helical measuring strategy can be an option, but this needs further investigation

The performance of 3500 MWth homogeneous and heterogeneous metal fueled core designs

International Nuclear Information System (INIS)

Turski, R.; Yang, Shi-tien.

1987-11-01

Performance parameters are calculated for a representative 3500 MWth homogeneous and a heterogeneous metal fueled reactor design. The equilibrium cycle neutronic characteristics, safety coefficients, control system requirements, and control rod worths are evaluated. The thermal-hydraulic characteristics for both configurations are also compared. The heavy metal fuel loading requirements and neutronic performance characteristics are also evaluated for the uranium startup option. 14 refs., 14 figs., 20 tabs

Recalculation of measured fuel nuclide concentrations

International Nuclear Information System (INIS)

Moeller, W.

1984-01-01

The concentrations and concentration ratios of heavy fuel nuclides determined in the Central Institute for Nuclear Research Rossendorf on the basis of destructive burnup measurements are compared with the results of microburnup calculations. The possibility is discussed to improve the results by taking into account the spectral characteristics at the positions of the measuring samples. (author)

Methods for measuring of fuel can deformation under radiation conditions

International Nuclear Information System (INIS)

Zelenchuk, A.V.; Fetisov, B.V.; Lakin, Yu.G.; Tonkov, V.Yu.

1978-01-01

The possibility for measuring fuel can deformation under radiation conditions by means of the acoustic method and tensoresistors is considered. The construction and operation of the in-pile facility for measuring creep of the fuel can specimen loaded by the internal pressure is described. The data on neutron radiation effect on changes in creep rate for zirconium fuel can are presented. The results obtained with tensoresistors are in a good agreement with those obtained by the acoustic method, which enables to recommend the use of both methods for the irradiation creep investigation of the fuel element cans

Measurement of dynamic interaction between a vibrating fuel element and its support

Energy Technology Data Exchange (ETDEWEB)

Fisher, N.J.; Tromp, J.H.; Smith, B.A.W. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada). Chalk River Labs.

1996-12-01

Flow-induced vibration of CANDU{reg_sign} fuel can result in fretting damage of the fuel and its support. A WOrk-Rate Measuring Station (WORMS) was developed to measure the relative motion and contact forces between a vibrating fuel element and its support. The fixture consists of a small piece of support structure mounted on a micrometer stage. This arrangement permits position of the support relative to the fuel element to be controlled to within {+-} {micro}m. A piezoelectric triaxial load washer is positioned between the support and micrometer stage to measure contact forces, and a pair of miniature eddy-current displacement probes are mounted on the stage to measure fuel element-to-support relative motion. WORMS has been utilized to measure dynamic contact forces, relative displacements and work-rates between a vibrating fuel element and its support. For these tests, the fuel element was excited with broadband random force excitation to simulate flow-induced vibration due to axial flow. The relationship between fuel element-to-support gap or preload (i.e., interference or negative gap) and dynamic interaction (i.e., relative motion, contact forces and work-rates) was derived. These measurements confirmed numerical simulations of in-reactor interaction predicted earlier using the VIBIC code.

Local Measurement of Fuel Energy Deposition and Heat Transfer Environment During Fuel Lifetime Using Controlled Calorimetry

International Nuclear Information System (INIS)

Don W. Miller; Andrew Kauffmann; Eric Kreidler; Dongxu Li; Hanying Liu; Daniel Mills; Thomas D. Radcliff; Joseph Talnagi

2001-01-01

A comprehensive description of the accomplishments of the DOE grant titled, ''Local Measurement of Fuel Energy Deposition and Heat Transfer Environment During Fuel Lifetime using Controlled Calorimetry''

Validation of the MC{sup 2}-3/DIF3D Code System for Control Rod Worth via the BFS-75-1 Reactor Physics Experiment

Energy Technology Data Exchange (ETDEWEB)

Yun, Sunghwan; Kim, Sang Ji [KAERI, Daejeon (Korea, Republic of)

2016-05-15

In this paper, control rod worths of the BFS-75-1 reactor physics experiments were examined using continuous energy MCNP models and deterministic MC2-3/DIF3D models based on the ENDF/B-VII.0 library. We can conclude that the ENDF/B-VII.0 library shows very good agreement in small-size metal uranium fuel loaded core which is surrounded by the depleted uranium blanket. However, the control rod heterogeneity effect reported by the reference is not significant in this problem because the tested control rod models were configured by single rod. Hence comparison with other control rod worth measurements data such as the BFS-109-2A reactor physics experiment is planned as a future study. The BFS-75-1 critical experiment was carried out in the BFS-1 facility of IPPE in Russia within the framework of validating an early phase of KALIMER- 150 design. The Monte-Carlo model of the BFS- 75-1 critical experiment had been developed. However, due to incomplete information for the BFS- 75-1 experiments, Monte-Carlo models had been generated for the reference criticality and sodium void reactivity measurements with disk-wise homogeneous model. Recently, KAERI performed another physics experiment, BFS-109-2A, by collaborating with Russian IPPE. During the review process of the experimental report of the BFS-109-2A critical experiments, valuable information for the BFS-1 facility which can also be used for the BFS-75-1 experiments was discovered.

Reliability Worth Analysis of Distribution Systems Using Cascade Correlation Neural Networks

DEFF Research Database (Denmark)

Heidari, Alireza; Agelidis, Vassilios; Pou, Josep

2018-01-01

Reliability worth analysis is of great importance in the area of distribution network planning and operation. The reliability worth's precision can be affected greatly by the customer interruption cost model used. The choice of the cost models can change system and load point reliability indices....... In this study, a cascade correlation neural network is adopted to further develop two cost models comprising a probabilistic distribution model and an average or aggregate model. A contingency-based analytical technique is adopted to conduct the reliability worth analysis. Furthermore, the possible effects...

Core concepts for 'zero-sodium-void-worth core' in metal fuelled fast reactor

International Nuclear Information System (INIS)

Chang, Y.I.; Hill, R.N.; Fujita, E.K.; Wade, D.C.; Kumaoka, Y.; Suzuki, M.; Kawashima, M.; Nakagawa, H.

1991-01-01

Core design options to reduce the sodium void worth in metal fuelled LMRs are investigated. Two core designs which achieve a zero sodium void worth are analyzed in detail. The first design is a 'pancaked' and annular core with enhanced transuranic burning capabilities; the high leakage in this design yields a low breeding ratio and small void worth. The second design is an axially multilayered annular core which is fissile self-sufficient; in this design, the upper and lower core regions are neutronically decoupled for reduced void worth while fissile self-sufficiency is achieved using internal axial blankets plus external radial and axial blanket-zones. The neutronic performance characteristics of these low void worth designs are assessed here; their passive safety properties are discussed in a companion paper. (author)

Fuel rod-to-support contact pressure and stress measurement for CHASNUPP-1(PWR) fuel

International Nuclear Information System (INIS)

Waseem; Elahi, N.; Siddiqui, A.; Murtaza, G.

2011-01-01

Research highlights: → A detailed finite element model of spacer grid cell with fuel rod-to-support has been developed to determine the contact pressure between the supports of the grid and fuel rod cladding. → The spring hold-down force is calculated using the contact pressure obtained from the FE model. → Experiment has also been conducted in the same environment for the measurement of this force. → The spring hold-down force values obtained from both studies confirm the validation of this analysis. → The stress obtained through this analysis is less than the yield strength of spacer grid material, thus fulfils the structural integrity criteria of grid. - Abstract: This analysis has been made in an attempt to measure the contact pressure of the PWR fuel assembly spacer grid spring and to verify its structural integrity at room temperature in air. A detailed finite element (FE) model of spacer grid cell with fuel rod-to-support has been developed to determine the contact pressure between the supports of the grid and fuel rod cladding. The FE model of a fuel rod-to-support system is produced with shell and contact elements. The spring hold-down force is calculated using the contact pressure obtained from the FE model. Experiment has also been conducted in the same environment for the measurement of this force. The spring hold-down force values obtained from both studies are compared, which show good agreement, and in turn confirm the validation of this analysis. The Stress obtained through this analysis is less than the yield strength of spacer grid material (Inconel-718), thus fulfils the structural integrity criteria of grid.

Fuel rod-to-support contact pressure and stress measurement for CHASNUPP-1(PWR) fuel

Energy Technology Data Exchange (ETDEWEB)

Waseem, E-mail: wazim_me@hotmail.co [Directorate General Nuclear Power Fuel, Pakistan Atomic Energy Commission, P.O. Box No. 1847, Islamabad 44000 (Pakistan); Elahi, N.; Siddiqui, A.; Murtaza, G. [Directorate General Nuclear Power Fuel, Pakistan Atomic Energy Commission, P.O. Box No. 1847, Islamabad 44000 (Pakistan)

2011-01-15

Research highlights: A detailed finite element model of spacer grid cell with fuel rod-to-support has been developed to determine the contact pressure between the supports of the grid and fuel rod cladding. The spring hold-down force is calculated using the contact pressure obtained from the FE model. Experiment has also been conducted in the same environment for the measurement of this force. The spring hold-down force values obtained from both studies confirm the validation of this analysis. The stress obtained through this analysis is less than the yield strength of spacer grid material, thus fulfils the structural integrity criteria of grid. - Abstract: This analysis has been made in an attempt to measure the contact pressure of the PWR fuel assembly spacer grid spring and to verify its structural integrity at room temperature in air. A detailed finite element (FE) model of spacer grid cell with fuel rod-to-support has been developed to determine the contact pressure between the supports of the grid and fuel rod cladding. The FE model of a fuel rod-to-support system is produced with shell and contact elements. The spring hold-down force is calculated using the contact pressure obtained from the FE model. Experiment has also been conducted in the same environment for the measurement of this force. The spring hold-down force values obtained from both studies are compared, which show good agreement, and in turn confirm the validation of this analysis. The Stress obtained through this analysis is less than the yield strength of spacer grid material (Inconel-718), thus fulfils the structural integrity criteria of grid.

Spent fuel disposal: is the underground the sole solution?

International Nuclear Information System (INIS)

Nachmilner, L.

1997-01-01

The following 4 major approaches to spent fuel disposal are discussed: permanent storage in an underground repository, reprocessing, partitioning and transmutation, and accelerator driven transmutation. It is concluded that underground disposal will remain the basic option for the near future, although pursuing the other methods is certainly worth while. (P.A.)

Fission gas release in LWR fuel measured during nuclear operation

International Nuclear Information System (INIS)

Appelhans, A.D.; Skattum, E.; Osetek, D.J.

1980-01-01

A series of fuel behavior experiments are being conducted in the Heavy Boiling Water Reactor in Halden, Norway, to measure the release of Xe, Kr, and I fission products from typical light water reactor design fuel pellets. Helium gas is used to sweep the Xe and Kr fission gases out of two of the Instrumented Fuel Assembly 430 fuel rods and to a gamma spectrometer. The measurements of Xe and Kr are made during nuclear operation at steady state power, and for 135 I following reactor scram. The first experiments were conducted at a burnup of 3000 MWd/t UO 2 , at bulk average fuel temperatures of approx. 850 K and approx. 23 kW/m rod power. The measured release-to-birth ratios (R/B) of Xe and Kr are of the same magnitude as those observed in small UO 2 specimen experiments, when normalized to the estimated fuel surface-to-volume ratio. Preliminary analysis indicates that the release-to-birth ratios can be calculated, using diffusion coefficients determined from small specimen data, to within a factor of approx. 2 for the IFA-430 fuel. The release rate of 135 I is shown to be approximately equal to that of 135 Xe

PWR and BWR spent fuel assembly gamma spectra measurements

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-11

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

Measurement of soluble nuclide dissolution rates from spent fuel

International Nuclear Information System (INIS)

Wilson, C.N.; Gray, W.J.

1990-01-01

Gaining a better understanding of the potential release behavior of water-soluble radionuclides is the focus of new laboratory spent fuel dissolution studies being planned in support of the Yucca Mountain Project. Previous studies have suggested that maximum release rates for actinide nuclides, which account for most of the long-term radioactivity in spent fuel, should be solubility-limited and should not depend on the characteristics or durability of the spent fuel waste form. Maximum actinide concentrations should be sufficiently low to meet the NRC (Nuclear Regulatory Commission) annual release limits. Potential release rates for soluble nuclides such as 99 Tc, 135 Cs, 14 C and 129 I, which account for about 1-2% of the activity in spent fuel at 1,000 years, are less certain and may depend on processes such as oxidation of the fuel in the repository air environment. Dissolution rates for several soluble nuclides have been measured from spent fuel specimens using static and semi-static methods. However, such tests do not provide a direct measurement of fuel matrix dissolution rates that may ultimately control soluble-nuclide release rates. Flow-through tests are being developed as a potential supplemental method for determining the matrix component of soluble-nuclide dissolution. Advantages and disadvantages of both semi-static and flow-through methods are discussed. Tests with fuel specimens representing a range of potential fuel states that may occur in the repository, including oxidized fuel, are proposed. Preliminary results from flow-through tests with unirradiated UO 2 suggesting that matrix dissolution rates are very sensitive to water composition are also presented

Nondestructive nuclear measurement in the fuel cycle. Part 1

International Nuclear Information System (INIS)

Lyoussi, A.

2005-01-01

Nondestructive measurement techniques are today widely used in practically all steps of the fuel cycle. This article is devoted to the presentation of the control and characterization needs and to the main passive nondestructive nuclear methods used: 1 - nondestructive nuclear measurement, needs and motivation: nuclear fuel cycle, nondestructive nuclear measurements (passive and active methods), comments; 2 - main passive nondestructive nuclear measurement methods: gamma spectroscopy (principle, detectors, electronic systems, data acquisition and signal processing, domains of application, main limitations), passive neutronic measurements (needs and motivations, neutron detectors, total neutronic counting, neutronic coincidences counting, neutronic multiplicities counting, comments). (J.S.)

In-pile measurement of the thermal conductivity of irradiated metallic fuel

International Nuclear Information System (INIS)

Bauer, T.H.; Holland, J.W.

1995-01-01

Transient test data and posttest measurements from recent in-pile overpower transient experiments are used for an in situ determination of metallic fuel thermal conductivity. For test pins that undergo melting but remain intact, a technique is described that relates fuel thermal conductivity to peak pin power during the transient and a posttest measured melt radius. Conductivity estimates and their uncertainty are made for a database of four irradiated Integral Fast Reactor-type metal fuel pins of relatively low burnup (<3 at.%). In the assessment of results, averages and trends of measured fuel thermal conductivity are correlated to local burnup. Emphasis is placed on the changes of conductivity that take place with burnup-induced swelling and sodium logging. Measurements are used to validate simple empirically based analytical models that describe thermal conductivity of porous media and that are recommended for general thermal analyses of irradiated metallic fuel

Control Rod Reactivity Measurements in the Aagesta Reactor with the Pulsed Neutron Method

Energy Technology Data Exchange (ETDEWEB)

Bjoereus, K

1969-07-01

An extensive series of control rod measurements was made in the Aagesta reactor during the low power experimental period following the first criticality. This report describes the part of these investigations made with the pulsed neutron method, comprising nearly 300 measurements. The main objective was the determination of control rod reactivity worths for different rods and groups of rods, but some supplementary measurements were also made, e.g. a determination of the prompt neutron decay constant for the delayed critical condition and four different cores. The cores consisted of 20, 32, 68, and 140 fuel elements respectively, and measurements were made at room temperature and with the moderator level close to critical for each core, and for the 140-element core also with full moderator height and at the temperatures 140 deg C and 215 deg C. Both fully and partly inserted control rod groups were investigated. The measurements at critical water level give directly the control rod reactivity worths, whereas those with full water height give the shut-down reactivity. A comparison was made between measured reactivity worths for a number of rod groups and those calculated with the HETERO code. The prompt neutron decay constant at delayed criticality {alpha}{sub 0}={beta}/l, for the full core at 215 deg C was found to be 9.60 {+-} 0.30/sec, corresponding to l = 0.76 {+-} 0.02 msec. The shut-down reactivity with 16 coarse control rods in pos. A-D 22, 40-04, 44, 26 is -5% at 25 deg C and -13% at 215 deg C. The relative error is usually around 8% in the reactivity worths, originating mainly from the higher harmonics content in the measured curves.

The Economics of Comparable Worth.

Science.gov (United States)

Killingsworth, Mark R.

This document concludes that the basic difficulty with comparable worth is that it is an ill-conceived solution to a serious problem and that alternative policies, such as equal employment opportunity legislation or application of antitrust laws, provide means of addressing employment discrimination that are both more effective and less likely to…

Shielding Performance Measurements of Spent Fuel Transportation Container

Directory of Open Access Journals (Sweden)

SUN Hong-chao

2015-11-01

Full Text Available The safety supervision of radioactive material transportation package has been further stressed and implemented. The shielding performance measurements of spent fuel transport container is the important content of supervision. However, some of the problems and difficulties reflected in practice need to be solved, such as the neutron dose rate on the surface of package is too difficult to measure exactly, the monitoring results are not always reliable, etc. The monitoring results using different spectrometers were compared and the simulation results of MCNP runs were considered. An improvement was provided to the shielding performance measurements technique and management of spent fuel transport.

Development of experimental methods for measuring fuel elements burnup

International Nuclear Information System (INIS)

PEREDA, C; HENRIQUEZ, C; NAVARRO, G; TORRES, H; KLEIN, J; CALDERON, D; MEDEL, J; MUTIS, O; DAIE, J; ITURRIETA, L; LONCOMILLA, M; ZAMBRANO, J; KESTELMAN, A

2003-01-01

This paper is a summary of the work carried out during the last two years in fuel burning measurements at RECH-1 for different enrichments, cooling times and burning rates. The measurements were made in two gamma-spectrometric facilities, one is installed in a hot cell and the other inside of the secondary pool of the RECH-1, where the element is under 2 meters of water. The hot cell measurements need at least 100 cooling days because of the problems generated by the transport of highly active fuel elements from the Reactor to the cell. This was the main reason for using the in-pool facility because of its capability to measure the burning of fuel elements without having to wait so long, that is with only 5 cooling days. The accumulated experience in measurements achieved in both facilities and the encouraging results show that this measuring method is reliable. The results agreed well with those obtained using the reactor's physics codes, which was the way they were obtained previously (Cw)

Fuels Management-How to Measure Success: Conference Proceedings

Science.gov (United States)

Patricia L. Andrews; Bret W. Butler

2006-01-01

Fuels management programs are designed to reduce risks to communities and to improve and maintain ecosystem health. The International Association of Wildland Fire initiated the 1st Fire Behavior and Fuels Conference to address development, implementation, and evaluation of these programs. The focus was on how to measure success. Over 500 participants from several...

Passive measurements of mixed-oxide fuel for nuclear nonproliferation

International Nuclear Information System (INIS)

Dolan, Jennifer L.; Flaska, Marek; Pozzi, Sara A.; Chichester, David L.

2013-01-01

We present new results on passive measurements and simulations of mixed-oxide fuel-pin assemblies. Potential tools for mixed-oxide fuel pin characterization are discussed for future nuclear-nonproliferation applications. Four EJ-309 liquid scintillation detectors coupled with an accurate pulse timing and digital, offline and optimized pulse-shape discrimination method were used. Measurement analysis included pulse-height distributions to distinguish between purely fission neutron sources and alpha-n plus fission neutrons sources. Time-dependent cross-correlation functions were analyzed to measure the fission neutron contribution to the measured sample's neutron source. The use of Monte Carlo particle transport code MCNPX-PoliMi is discussed in conjunction with the measurements

Subcritical Noise Analysis Measurements with Fresh and Spent Research Reactor Fuels Elements

International Nuclear Information System (INIS)

Valentine, T.E.; Mihalczo, J.T.; Kryter, R.C.; Miller, V.C.

1999-01-01

The verification of the subcriticality is of utmost importance for the safe transportation and storage of nuclear reactor fuels. Transportation containers and storage facilities are designed such that nuclear fuels remain in a subcritical state. Such designs often involve excess conservatism because of the lack of relevant experimental data to verify the accuracy of Monte Carlo codes used in nuclear criticality safety analyses. A joint experimental research program between Oak Ridge National Laboratory, Westinghouse Safety Management Solutions, Inc., and the University of Missouri was initiated to obtain measured quantities that could be directly related to the subcriticality of simple arrays of Missouri University Research Reactor (MURR) fuel elements. A series of measurement were performed to assess the reactivity of materials such as BORAL, stainless steel, aluminum, and lead that are typically used in the construction of shipping casks. These materials were positioned between the fuel elements. In addition, a limited number of measurements were performed with configurations of fresh and spent (irradiated) fuel elements to ascertain the reactivity of the spent fuel elements. In these experiments, fresh fuel elements were replaced by spent fuel elements such that the subcritical reactivity change could be measured. The results of these measurements were used by Westinghouse Safety Management Solutions to determine the subcriticality of MURR fuel elements isolated by absorbing materials. The measurements were interpreted using the MCNP-DSP Monte Carlo code to obtain the subcritical neutron multiplication factor k(sub eff), and the bias in K(sub eff) that are used in criticality safety analyses

Sodium-cooled fast reactor core designs for transmutation of MHR spent fuel

International Nuclear Information System (INIS)

Hong, S. G.; Kim, Y. H.; Venneri, F.

2010-01-01

In this paper, the core design analyses of sodium cooled fast reactors (SFR) are performed for the effective transmutation of the DB (Deep Burn)-MHR (Modular Helium Reactor). In this concept, the spent fuels of DB-MHR are transmuted in SFRs with a closed fuel cycle after TRUs from LWR are first incinerated in a DB-MHR. We introduced two different type SFR core designs for this purpose, and evaluated their core performance parameters including the safety-related parameters. In particular, the cores are designed to have lower transmutation rate relatively to our previous work so as to make the fuel characteristics more feasible. The first type cores which consist of two enrichment regions are typical homogeneous annular cores and they rate 900 MWt power. On the other hand, the second type cores which consist of a central non-fuel region and a single enrichment fuel region rate relatively higher power of 1500 MWt. For these cores, the moderator rods (YH 1.8 ) are used to achieve less positive sodium void worth and the more negative Doppler coefficient because the loading of DB-MHR spent fuel leads to the degradation of these safety parameters. The analysis results show that these cores have low sodium void worth and negative reactivity coefficients except for the one related with the coolant expansion but the coolant expansion reactivity coefficient is within the typical range of the typical SFR cores. (authors)

Safety characteristics of mid-sized MOX fueled liquid metal reactor core of high converter type in the initiating phase of unprotected loss of flow accident. Effect of low specific fuel power density on ULOF behavior brought by employment of large diameter fuel pins

International Nuclear Information System (INIS)

Ishida, Masayoshi; Kawada, Kenichi; Niwa, Hajime

2003-07-01

Safety characteristics in core disruptive accidents (CDAs) of mid-sized MOX fueled liquid metal reactor core of high converter type have been examined by using the CDA initiating phase analysis code SAS4A. The design concept of high converter type reactor core has been studied as one of options in the category of sodium-cooled reactor in Phase II of Feasibility Study on Commercialized Fast Reactor Cycle System. An unprotected loss-of-flow accident (ULOF) has been selected as a representative CDA initiator for this study. A core concept of high converter type, which employed a large diameter fuel pin of 11.1 mm with 1.2 m core height to get a large fuel volume fraction in the core to achieve high internal conversion ratio was proposed in JFY2001. Each fuel subassembly of the core (abbreviated here as UPL120)was provided with an upper sodium plenum directly above the core to reduce the sodium void reactivity worth. Because of the large fuel pin diameter, average specific fuel power density (31 kW/kg-MOX) of UPL120 is about one half of those of conventional large MOX cores. The reactivity worth of sodium voiding is 6$ in the whole core, and -1$ in the all upper plenums. Initiating phase of ULOF accident in UPL120 under the conditions of nominal design and best estimate analysis resulted in a slightly super-prompt critical power burst. The causes of the super-prompt criticality have been identified twofold: (a) the low specific fuel power density of core reduced the effectiveness of prompt negative reactivity feedback of Doppler and axial fuel expansion effects upon increase in reactor power, and (b) the longer core height compared with conventional 1m cores brought, together with the lower specific power density, a remarkable delay in insertion of negative fuel dispersion reactivity after the onset of fuel disruption in sodium voided subassembly due to the lower linear heat rating in the top portion of the core. During the delay, burst-type fuel failures in sodium un

Measuring the plutonium distribution in fuel elements by the gamma scanning method

International Nuclear Information System (INIS)

Gorobets, A.K.; Leshchenko, Yu.I.; Semenov, A.L.

1982-01-01

An on-line system designed for measuring Pu distribution in the length of fresh fuel elements with vibrocompacted UO 2 -PuO 2 fuel rods by the γ-scanning method is described. An algorithm for measurement result processing and the procedure of determination of calibration parameters necessary for the valid signal separat.ion by means of a two-channel analyzer and for evaluation of the self-absorption effect are considered. The device scanning unit consists of two NaI(Tl) detectors simultaneously detecting γ-radiation from the opposite sides of a measured fuel rod section. The cesium source with Esub(γ)=660 keV is used for fuel scanning. On the base of the analysis of the results obtained when studying the BOR-60 experimental fuel elements with fuel rods of 400 mm long by means of the described device clusion is made that fuel element scanning during 20 min (scanning step is 4 mm, measuring time at each step is 10 s) makes it possible to determine Pu distribution with the error less than +-4% at the confidence probability of 0.68

Nuclear analysis of the experimental VHTR fuel lattice

International Nuclear Information System (INIS)

Doi, Takeshi; Shindo, Ryuiti; Hirano, Mitsumasa; Takano, Makoto

1984-11-01

Nuclear properties of a fuel lattice in the experimental VHTR core were analyzed with DELIGHT-6 and SRAC codes. Analytical results by both codes were compared by using various calculational model. The nuclear parameters were analyzed, such as a multiplication factor of a fuel lattice and it's variation with burnup, a temperature effect on reactivity, an effect of double-heterogeniety in a resonance absorption calculation, a resonance integral of 238 U and a reactivity worth of burnable poison. From these analyses, following results were obtained. Firstly, on calculational models, 1) Effect of double-heterogeniety in the resonance absorption calculation for Mark-III fuel element, causing a decrease of about 5.5 barns in the resonance integral and an increase of about 2.6 %ΔK in the infinite multiplication factor, 2) The heterogeneous calculation with the collision probability method resulted in about 0.6 %ΔK higher the multiplication factor of fuel lattice than that with the point model, 3) The reactivity worth of burnable poison rod by a multi-region model is about 20 % less than that by a 2-region model at an initial state of burnup and it's variation with burnup are fairly different, Secondly, on comparison between the results by DELIGHT-6 and SRAC, 4) The nuclear parameters obtained with both codes agreed well, Lastly, on the improvement in DELIGHT-6, 5) Consideration of the neutron spectrum shielding effect in the resonance effective cross section calculation caused a decrease of about 2.4 %ΔK in the multiplication factor of fuel lattice, 6) The lattice multiplication factor increased about 0.5 %ΔK by introducing lambda-parameters for the non-resonant nuclie. (J.P.N.)

Synergistic Smart Fuel For Microstructure Mediated Measurements

Energy Technology Data Exchange (ETDEWEB)

James A. Smith; Dale K. Kotter; Steven L. Garrett; Randall A. Ali

2013-07-01

Advancing the Nuclear Fuel Cycle and Next Generation Nuclear Power Plants requires enhancing our basic understanding of fuel and materials behavior under irradiation. The two most significant issues limiting the effectiveness and lifespan of the fuel are the loss of thermal conductivity of the fuel and the mechanical strength of both fuel and cladding. The core of a nuclear reactor presents an extremely harsh and challenging environment for both sensors and telemetry due to elevated temperatures and large fluxes of energetic and ionizing particles from radioactive decay processes. The majority of measurements are made in reactors using “radiation hardened” sensors and materials. A different approach has been pursued in this research that exploits high temperatures and materials that are robust with respect to ionizing radiation. This synergistically designed thermoacoustic sensor will be self-powered, wireless, and provide telemetry. The novel sensor will be able to provide reactor process information even if external electrical power and communication are unavailable. In addition, the form-factor for the sensor is identical to the existing fuel rods within reactors and contains no moving parts. Results from initial proof of concept experiments designed to characterize porosity, surface properties and monitor gas composition will be discussed.

Feasibility study of the design of homogeneously mixed thorium-uranium oxide and all-uranium fueled reactor cores for civil nuclear marine propulsion - 15082

International Nuclear Information System (INIS)

Alam, S.B.; Lindley, B.A.; Parks, G.T.

2015-01-01

In this reactor physics study, we attempt to design a civil marine reactor core that can operate over a 10 effective-full-power-years life at 333 MWth using ThUO 2 and all-UO 2 fuel. We use WIMS to develop subassembly designs and PANTHER to examine whole-core arrangements, optimizing: subassembly and core geometry; fuel enrichment; burnable and moveable poison design; and whole-core loading patterns. We compare designs with a 14% fissile loading for ThUO 2 and all-UO 2 fuel in 13*13 assemblies with ZrB 2 integral fuel burnable absorber pins for reactivity control. Taking advantage of self-shielding effects, the ThUO 2 option shows greater promise in the final burnable poison design while maintaining low, stable reactivity with minimal burnup penalty. For the final poisoning design with ZrB 2 , ThUO 2 contributes 2.5% more initial reactivity suppression, although the all-UO 2 design exhibits lower reactivity swing. All the candidate materials show greater rod worth for the ThUO 2 design. For both fuels, B 4 C has the highest reactivity worth, providing 10% higher control rod worth for ThUO 2 fuel than all-UO 2 . Finally, optimized assemblies were loaded into a 3D reactor model in PANTHER. The PANTHER results show that after 10 years, the core is on the border of criticality, confirming the fissile loading is well-designed. (authors)

Improved Monte Carlo - Perturbation Method For Estimation Of Control Rod Worths In A Research Reactor

International Nuclear Information System (INIS)

Kalcheva, Silva; Koonen, Edgar

2008-01-01

A hybrid method dedicated to improve the experimental technique for estimation of control rod worths in a research reactor is presented. The method uses a combination of Monte Carlo technique and perturbation theory. The perturbation theory is used to obtain the relation between the relative rod efficiency and the buckling of the reactor with partially inserted rod. A series of coefficients, describing the axial absorption profile are used to correct the buckling for an arbitrary composite rod, having complicated burn up irradiation history. These coefficients have to be determined - by experiment or by using some theoretical/numerical method. In the present paper they are derived from the macroscopic absorption cross sections, obtained from detailed Monte Carlo calculations by MCNPX 2.6.F of the axial burn up profile during control rod life. The method is validated on measurements of control rod worths at the BR2 reactor. Comparison with direct Monte Carlo evaluations of control rod worths is also presented. The uncertainties, arising from the used approximations in the presented hybrid method are discussed. (authors)

Analysis of alternative light water reactor (LWR) fuel cycles

International Nuclear Information System (INIS)

Heeb, C.M.; Aaberg, R.L.; Boegel, A.J.; Jenquin, U.P.; Kottwitz, D.A.; Lewallen, M.A.; Merrill, E.T.; Nolan, A.M.

1979-12-01

Nine alternative LWR fuel cycles are analyzed in terms of the isotopic content of the fuel material, the relative amounts of primary and recycled material, the uranium and thorium requirements, the fuel cycle costs and the fraction of energy which must be generated at secured sites. The fuel materials include low-enriched uranium (LEU), plutonium-uranium (MOX), highly-enriched uranium-thorium (HEU-Th), denatured uranium-thorium (DU-Th) and plutonium-thorium (Pu-Th). The analysis is based on tracing the material requirements of a generic pressurized water reactor (PWR) for a 30-year period at constant annual energy output. During this time period all the created fissile material is recycled unless its reactivity worth is less than 0.2% uranium enrichment plant tails

Recent advances in hardware and software are to improve spent fuel measurements

International Nuclear Information System (INIS)

Staples, P.; Beddingfield, D.H.; Lestone, J.P.; Pelowitz, D.G.; Bytchkov, M.; Starovich, Z.; Harizanov, I.; Luna-Vellejo, J.; Lavender, C.

2001-01-01

Vast quantities of spent fuel are available for safeguard measurements, primarily in Commonwealth of Independent States (CIS) of the former Soviet Union. This spent fuel, much of which consists of long-cooling-time material, is going to become less unique in the world safeguards arena as reprocessing projects or permanent repositories continue to be delayed or postponed. The long cooling time of many of the spent fuel assemblies being prepared for intermediate term storage in the CIS countries promotes the possibility of increased accuracy in spent fuel assays. This improvement is made possible through the process of decay of the Curium isotopes and of fission products. An important point to consider for the future that could advance safeguards measurements for reverification and inspection would be to determine what safeguards requirements should be imposed upon this 'new' class of spent fuel, Improvements in measurement capability will obviously affect the safeguards requirements. What most significantly enables this progress in spent fuel measurements is the improvement in computer processing power and software enhancements leading to user-friendly Graphical User Interfaces (GUT's). The software used for these projects significantly reduces the IAEA inspector's time expenditure for both learning and operating computer and data acquisition systems, At the same time, by standardizing the spent fuel measurements, it is possible to increase reproducibility and reliability of the measurement data. Hardware systems will be described which take advantage of the increased computer control available to enable more complex measurement scenarios. A specific example of this is the active regulation of a spent fuel neutron coincident counter's 3 He tubes high voltage, and subsequent scaling of measurement results to maintain a calibration for direct assay of the plutonium content of Fast Breeder Reactor spent fuel. The plutonium content has been successfully determined for

Measurements of fuel temperature coefficient of reactivity on a commercial AGR

International Nuclear Information System (INIS)

Telford, A.; Bridge, M.J.

1978-01-01

Tests have been carried out on the commercial AGR at Hikley Point to determine the fuel temperature coefficient of reactivity, an important safety related parameter. Reactor neutron flux was measured during transients induced by movement of a bank of control rods from one steady position to another. An inverse kinetics analysis was applied to the measured flux to determine the change which occured in core reactivity as the fuel temperature changed. The variation of mean fuel temperature was deduced from the flux transient by means of a nine-plane thermal hydraulics representation of the AGR fuel channel. Results so far obtained confirm the predicted variation of fuel temperature coefficient with butn-up. (author)

Spent-fuel composition: a comparison of predicted and measured data

International Nuclear Information System (INIS)

Thomas, C.C. Jr.; Cobb, D.D.; Ostenak, C.A.

1981-03-01

The uncertainty in predictions of the nuclear materials content of spent light-water reactor fuel was investigated to obtain guidelines for nondestructive spent-fuel verification and assay. Values predicted by the reactor operator were compared with measured values from fuel reprocessors for six reactors (three PWR and three BWR). The study indicates that total uranium, total plutonium, fissile uranium, fissile plutonium, and total fissile content can be predicted with biases ranging from 1 to 6% and variabilities (1-sigma) ranging from 2 to 7%. The higher values generally are associated with BWRs. Based on the results of this study, nondestructive assay measurements that are accurate and precise to 5 to 10% (1sigma) or better should be useful for quantitative analyses of typical spent fuel

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

International Nuclear Information System (INIS)

Magnusson, Per

2008-01-01

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 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% 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 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

Mapping of sodium void worth and doppler effect for sodium-cooled fast reactor - 15458

International Nuclear Information System (INIS)

Krepel, J.; Pelloni, S.; Bortot, S.; Panadero, A.L.; Mikityuk, K.

2015-01-01

The sodium-cooled fast reactor (SFR) represents the reference and the most technologically mastered system among the Generation-IV reactors. Nevertheless, the sodium void worth in the fuel regions of SFR is usually positive. To overcome this safety drawback, low-void sodium-cooled fast spectrum core (CFV) was proposed by CEA. Such a CFV core is used in the frame of WP6 'Core safety' of the FP7 Euratom ESNII+ project as a reference SFR design. The overall sodium void effect is negative for the CFV core. Nevertheless, locally it is positive in the fuel region and negative in the sodium plenum. Similarly, also the Doppler effect is spatially dependent and it varies between the inner and outer fuel regions and between the middle and lower blankets. Accordingly, knowledge of the local distributions or actually mappings of the two safety-related parameters will be necessary, before safety assessment and transient analysis can be done. In this study these maps have been produced using the deterministic code ERANOS. The obtained mapping shows strong local dependency of both safety-related effects. A sensitivity of the void effect to the sodium plenum modeling was also demonstrated. The results may serve as an input for the transient analysis of the CFV core or as a cross-check for the Monte Carlo method based maps. (authors)

Development of the method for the dimensional measurement of the HANARO nuclear fuel

International Nuclear Information System (INIS)

Kim, Tae Yeon; Lee, K. S.; Park, D. G.; Choo, Y. S.; Ahn, S. B.

1998-06-01

Dimension of the nuclear fuel is altered in nuclear reactor because of the neutron exposure with high pressure water. If the deformation is overlarge, the severe problem in safety of the nuclear fuel and the reactor come about. Therefore the accurate dimensional data of the nuclear fuel in diameter and length is very important for the design of the nuclear fuel and the estimation of the nuclear safety. Measurement of diameter for the dummy HANARO fuel rod which has not filled with real fuel material was carried out in hot cell. And also the length of the HANARO fuel assembly and the rod are measured. Dimensional measuring method for the HANARO fuel was developed. The test result show our method is good enough to distinguish change in volume with statistical uncertainty of 0.6 %. (author). 2 refs., 7 tabs., 20 figs

Applications of a transportable spent-fuel measurement system

International Nuclear Information System (INIS)

Halbig, J.K.; Bosler, G.E.; Klosterbuer, S.F.; Rinard, P.M.

1985-01-01

A portable tool for making measurements on irradiated fuel has been developed to where in-plant installations having a 1 to 2% measurement uncertainty of relative exposure are feasible. The measurement uses a passive gross neutron signal and data from a gross gamma measurement as a consistency check of the neutron result and the operators declaration of cooling time. The uncertainties are about the same as those obtained using high-resolution gamma-ray techniques without the instrumentation being as obtrusive. The battery-operated microprocessor-based electronics package used with the irradiated fuel measurement system can also be used with single channel pulse counting detectors for other applications. This feature together with the large dynamic range of its current-mode ion chamber channel makes ION-I a good building block to be used in emergencies with an arsenal of detectors at a variety of nuclear plants. 8 figs., 3 tabs

Time-resolved measurements of supersonic fuel sprays using synchrotron X-rays.

Science.gov (United States)

Powell, C F; Yue, Y; Poola, R; Wang, J

2000-11-01

A time-resolved radiographic technique has been developed for probing the fuel distribution close to the nozzle of a high-pressure single-hole diesel injector. The measurement was made using X-ray absorption of monochromatic synchrotron-generated radiation, allowing quantitative determination of the fuel distribution in this optically impenetrable region with a time resolution of better than 1 micros. These quantitative measurements constitute the most detailed near-nozzle study of a fuel spray to date.

Use of TRIGA flip fuel for improved in-core irradiations

Energy Technology Data Exchange (ETDEWEB)

Whittemore, W L [General Atomic Co., San Diego, CA (United States)

1974-07-01

Use of standard TRIGA fuel (20% enriched uranium) in a reactor provides a suitable facility for in-core irradiations. However, large numbers of in-core samples irradiated for long periods (many months) can be handled more economically with a TRIGA loaded with FLIP fuel. As an example, ten or more in-core thermionic devices (each worth 50 to 80 cents with respect to a water-filled position) were irradiated in the Mark III TRIGA at General Atomic Company for 18 months with only a modest change in excess reactivity due to core burnup. A core loading of FLIP fuel has been added to the General Atomic Mark F reactor in order to provide numerous in-core irradiation sites for the production of radioisotopes. Since the worth of a 500-gram sample of a molybdenum compound (used for the production of {sup 99}Mo) is about 25 to 50 cents with respect to a water-filled position, use of a FLIP- TRIGA core will permit the irradiation of more than 5 kilograms of a molybdenum compound. A procedure is under development for the production of {sup 99}Mo with relatively high specific activity. Several techniques to concentrate {sup 99}Mo have been tested experimentally. The results will be reported. (author)

[Contingencies of self-worth in Japanese culture: validation of the Japanese contingencies of self-worth scale].

Science.gov (United States)

Uchida, Yukiko

2008-08-01

The author developed a Japanese version of the Contingencies of Self-Worth Scale (CSWS) that was originally developed in the United States (Crocker, Luhtanen, Cooper, & Bouvrette, 2003). The Japanese version of the scale measures seven contingencies of self-esteem: Defeating others in competition, appearance, relationship harmony, other's approval, academic competence, virtue, and support of family and friends. Scores on the scale had systematic relationships with related variables, and the scale therefore exhibited satisfactory levels of construct validity: Relationship harmony, other's approval, and support of family and friends were positively correlated with sympathy and interdependence, whereas competitiveness was negatively correlated with sympathy. Moreover, competitiveness and academic achievement contingencies predicted competitive motivation, whereas the support of family and friends contingency predicted self-sufficient motivation. The scale has adequate test-retest reliability and a seven-factor structural model was confirmed. The implications for self-esteem and interpersonal relationships in Japanese culture are discussed.

Measurement of reactivity effect caused by nonuniform fuel distribution

International Nuclear Information System (INIS)

Yamane, Yoshihiro; Hirano, Yasushi; Yasui, Hazime; Nishina, Kojiro; Shiroya, Seiji

1991-01-01

A reactivity effect due to a spatial variation of nuclear fuel concentration is an important problem in a reprocessing plant. To estimate this reactivity effect theoretically, the ''Goertzel's necessary condition, and th Fuel Importance'' theory have been proposed. In order to verify these theories, we have performed systematic measurements of reactivity effect due to the nonuniformity in the fuel distribution within the Kyoto University Critical Assembly. Neutron flux distribution and Fuel Importance distribution were also determined. A nonuniform assembly whose fuel concentration in the center region was 40% higher than the uniform one was found to have an excess reactivity of 0.3%Δk/k, with the same total uranium mass for which the uniform assembly was just critical. Moreover, its spatial distribution of thermal neutron flux and of Fuel Importance were more flat than those of the uniform assembly, as expected by the Goertzel's condition and the Fuel Importance theory. (Author)

Optimization analysis of the nuclear fuel cycle transition to the last core

International Nuclear Information System (INIS)

Rebollo, L.; Blanco, J.

2001-01-01

The Zorita NPP was the first Spanish commercial nuclear reactor connected to the grid. It is a 160 MW one loop PWR, Westinghouse design, owned by UFG, in operation since 1968. The configuration of the reactor core is based on 69 fuel elements type 14 x 14, the standard reload of the present equilibrium cycle being based on 16 fuel elements with 3.6% enrichment in 235 U. In order to properly plan the nuclear fuel management of the transition cycles to its end of life, presently foreseen by 2008, an based on the non-reprocessing option required by the policy of the Spanish Administration, a technical-economical optimization analysis has been performed. As a result, a fuel management strategy has been defined looking for getting simultaneously the minimum integral fuel cost of the transition from the present equilibrium cycle to the last core, as well as the minimum residual worth of the fuel remaining in the core after the final outage. Based on the ''lessons learned'' derived from the study, the time margin for the decision making has been determined, and a planning of the nuclear fuel supply for the transition reloads, specifying both the number of fuel elements and their enrichment in 235 U, as been prepared. Finally, based on the calculated economical worth of the partially burned fuel of the last core, after the end of its operation cycle, a financial cover for yearly compensation from now on of the foreseen final lost has been elaborated. Most of the conceptual conclusions obtained are applicable to the other commercial nuclear reactors in operation owned by UFG, so that they are understood to be of general interest and broad application to commercial PWR. (author)

Study on evaluating the reactivity worth of the control rods of the PWR 900 MWe

International Nuclear Information System (INIS)

Phan Quoc Vuong; Tran Vinh Thanh; Tran Viet Phu

2015-01-01

Control rods of a nuclear reactor are divided into two groups: shut down and power control. Reactivity worth of the control rods depends nonlinearly on the rods' compositions and positions where the rods are inserted into the core. Therefore, calculation of control rod worth is of high important. In this study, we calculated the reactivity worth of the power control rod bank of the Mitsubishi PWR 900 MWe. The results are integral and differential worth calibration of the control rods. (author)

Correlates of self-worth and body size dissatisfaction among obese Latino youth.

Science.gov (United States)

Mirza, Nazrat M; Mackey, Eleanor Race; Armstrong, Bridget; Jaramillo, Ana; Palmer, Matilde M

2011-03-01

The current study examined self-worth and body size dissatisfaction, and their association with maternal acculturation among obese Latino youth enrolled in a community-based obesity intervention program. Upon entry to the program, a sample of 113 participants reported global self-worth comparable to general population norms, but lower athletic competence and perception of physical appearance. Interestingly, body size dissatisfaction was more prevalent among younger respondents. Youth body size dissatisfaction was associated with less acculturated mothers and higher maternal dissatisfaction with their child's body size. By contrast, although global self-worth was significantly related to body dissatisfaction, it was not influenced by mothers' acculturation or dissatisfaction with their own or their child's body size. Obesity intervention programs targeted to Latino youth need to address self-worth concerns among the youth as well as addressing maternal dissatisfaction with their children's body size. Copyright © 2010 Elsevier Ltd. All rights reserved.

Laser induced fluorescence measurements of the mixing of fuel oil with air

Energy Technology Data Exchange (ETDEWEB)

Arnold, A; Bombach, R; Hubschmid, W; Kaeppeli, B [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

1999-08-01

We report on measurements of the mixing of fuel oil with air at atmospheric pressure in an industrial premixed gas turbine burner. The concentration of the vaporized fuel oil was measured with laser induced fluorescence. We reason that the fuel oil concentration can be considered with good accuracy as proportional to the fluorescence intensity. (author) 6 fig., 3 refs.

Self-reported "worth it" rating of aesthetic surgery in social media.

Science.gov (United States)

Domanski, Mark C; Cavale, Naveen

2012-12-01

A wide variety of surveys have been used to validate the satisfaction of patients who underwent aesthetic surgery. However, such studies are often limited by patient number and number of surgeons. Social media now allows patients, on a large scale, to discuss and rate their satisfaction with procedures. The views of aesthetic procedures patients expressed in social media provide unique insight into patient satisfaction. The "worth it" percentage, average cost, and number of respondents were recorded on October 16, 2011, for all topics evaluated on the aesthetic procedure social media site www.realself.com . Procedures were divided into categories: surgical, liposuction, nonsurgical, and dental. For each group, procedures with the most respondents were chosen and ordered by "worth it" score. A literature search was performed for the most commonly rated surgical procedures and the satisfaction rates were compared. A total of 16,949 evaluations of 159 aesthetic surgery topics were recorded. A correlation between cost of the procedure and percentage of respondents indicating that the procedure was "worth it" was not found. The highest-rated surgical procedure was abdominoplasty, with 93 % of the 1,589 self-selected respondents expressing that abdominoplasty was "worth it." The average self-reported cost was $8,400. The highest-rated nonsurgical product was Latisse, with 85 % of 231 respondents reporting it was "worth it" for an average cost of $200. The satisfaction scores in the literature for commonly rated surgical procedures ranged from 62 to 97.6 %. No statistically significant correlations between literature satisfaction scores and realself.com "worth it" scores were found. Abdominoplasty had the highest "worth it" rating among aesthetic surgical procedures. Aesthetic surgeons should be wary that satisfaction scores reported in the literature might not correlate with commonly achieved results. Social media has opened a new door into how procedures are

Detailed analysis for a control rod worth of the gas turbine high temperature reactor (GTHTR300)

Energy Technology Data Exchange (ETDEWEB)

Nakata, Tetsuo; Katanishi, Shoji; Takada, Shoji; Yan, Xing; Kunitomi, Kazuhiko [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

2002-11-01

GTHTR300 is composed of a simplified and economical power plant based on an inherent safe 600 MWt reactor and a nearly 50% high efficiency gas turbine power conversion cycle. GTHTR300 core consist of annular fuel region, center and outer side reflectors because of cooling it effectively in depressurized accident conditions, and all control rods are located in both side reflectors of annular core. As a thermal neutron spectrum is strongly distorted in reflector regions, an accurate calculation is especially required for the control rod worth evaluation. In this study, we applied the detailed Monte Carlo calculations of a full core model, and confirmed that our design method has enough accuracy. (author)

Time-resolved measurements of supersonic fuel sprays using synchrotron x-rays

International Nuclear Information System (INIS)

Powell, C.F.; Yue, Y.; Poola, R.; Wang, J.

2000-11-01

A time-resolved radiographic technique has been developed for probing the fuel distribution close to the nozzle of a high-pressure single-hole diesel injector. The measurement was made using X-ray absorption of monochromatic synchrotron-generated radiation, allowing quantitative determination of the fuel distribution in this optically impenetrable region with a time resolution of better than 1 μs. These quantitative measurements constitute the most detailed near-nozzle study of a fuel spray to date

Development method for measuring thickness of nuclei and coating of fuel plates

International Nuclear Information System (INIS)

Borges Junior, Reinaldo

2013-01-01

One of the most important components of a nuclear reactor is the Nuclear Fuel. Currently, the most advanced commercial fuel, whose applicability in Brazilian reactors has been developed by IPEN since 1985, is the silicide U 3 Si 2 . This is formed by fuel plates with nuclei dispersion (where the fissile material (U 3 Si 2 ) is homogeneously dispersed in a matrix of aluminum) coated aluminum. This fuel is produced in Brazil with developed technology, the result of the efforts made by the group of manufacturing nuclear fuel (CCN - Center of Nuclear Fuel) of IPEN. Considering the necessity of increasing the power of the IEA- R1 and Brazilian Multipurpose Reactor Building (RMB), for the production of radioisotopes - mainly for the area of medicine - there will be significant increase in the production of nuclear fuel at IPEN. Given this situation, if necessary, make the development of more modern and automated classification techniques. Aiming at this goal, this work developed a new computational method for measuring thickness of core and cladding of fuel plates, which are able to perform such measurements in less time and with more meaningful statistical data when compared with the current method of measurement. (author)

The prospect for fuel ion ratio measurements in ITER by collective Thomson scattering

DEFF Research Database (Denmark)

Stejner Pedersen, Morten; Korsholm, Søren Bang; Nielsen, Stefan Kragh

2012-01-01

We show that collective Thomson scattering (CTS) holds the potential to become a new diagnostic principle for measurements of the fuel ion ratio, nT/nD, in ITER. Fuel ion ratio measurements will be important for plasma control and machine protection in ITER. Measurements of ion cyclotron structures...... in CTS spectra have been suggested as the basis for a new fuel ion ratio diagnostic which would be well suited for reactor environments and capable of providing spatially resolved measurements in the plasma core. Such measurements were demonstrated in recent experiments in the TEXTOR tokamak. Here we...... conduct a sensitivity study to investigate the potential measurement accuracy of a CTS fuel ion ratio diagnostic on ITER. The study identifies regions of parameter space in which CTS can be expected to provide useful information on plasma composition, and we find that a CTS fuel ion ratio diagnostic could...

Proximal predictors of depressive symptomatology: perceived losses in self-worth and interpersonal domains and introjective and anaclitic mood states.

Science.gov (United States)

Kopala-Sibley, Daniel C; Zuroff, David C

2010-01-01

Although much research has demonstrated a relationship between negative life events and depressive symptoms, relatively little research has examined the mechanisms that may mediate this relationship. The theories of Blatt (1974), Bowlby (1980), and Gilbert (1992) each propose proximal predictors of depression. In accordance with these theories, this study examined the relationships among perceived losses in self-worth and interpersonal relationships, anaclitic (dependent) and introjective (self-critical) mood states, and depressive symptoms following a significant negative life event. A sample of 172 undergraduate students completed measures of depressive symptoms and depressive vulnerability factors and retrospectively described the worst period of their lives. They also rated the extent to which the events surrounding this worst period affected their self-worth and their relationships with close others. Structural equation modeling demonstrated that the effect of a perceived loss of self-worth on depressive symptoms was fully mediated by both introjective and anaclitic mood states, whereas the effect of a perceived loss of interpersonal relationships on depressive symptoms was fully mediated by an anaclitic mood state. Additionally, perceived losses of self-worth showed a stronger effect on introjective mood in highly self-critical individuals. Findings highlight the importance of perceived losses in both self-worth and interpersonal domains in response to adverse life events and suggest pathways through which perceived losses may affect depressive symptoms.

Reliabilty worth: Development of a relationship with outage magnitude, duration and frequency

International Nuclear Information System (INIS)

Turner, F.P.P.; Katrichak, A.M.; Dwyer, A.; Edwards, D.; Ibrahim, A.

1994-01-01

British Columbia Hydro's Worth Project Team was founded to determine values for reliability for reference in evaluation of investment and operating decisions. Work to date has produced key preliminary values for specific outages and concepts for the shape of the relationship between value and these determinates of reliability worth, frequency, magnitude and duration. These values and concepts are described. The values are developed through an iterative, trial and refinement approach. The approach incorporates direct input from customers, common sense and judgement, and micro- and macro-economic concepts. Reliability worth values for reduced or prevented outages are presented for residential, commercial, small industrial and mixed sectors and various outage durations. Reliability worth values were obtained through customer surveys. Limitations of the reliability worth value are numerous and are listed. Study of cost vs magnitude of interruption using microeconomic models has shown that costly system improvements to reduce the possibility of widespread outages may not be justified. The case of exceptionally large area outages (blackouts) is examined. The cost vs frequency relationship was examined in terms of the economic concept of utility or satisfaction. Different loss/frequency characteristics are demonstrated for different customer classes. Customer value for reduced outage duration is expressed in a curve with flatter slope than that for eliminated outages. 2 refs., 6 figs

Measuring method for amount of fissionable gas in spent fuel pellet

International Nuclear Information System (INIS)

Kashibe, Shinji.

1992-01-01

The method of the present invention separately measures the amount of both of a fission product (FP) gas accumulated in bubbles at the crystal grain boundary of spent fuel pellets and an FP gas accumulated in the crystal grains. That is, in a radial position of the spent fuel pellet, a microfine region is mechanically destroyed. The amount of the FP gas released by the destruction from the crystal grains is measured by using a mass analyzer. Then, when the destroyed pieces formed by the destruction are recovered and dissolved, FP gas accumulated in the crystal grains of the pellet is released. The amount released is measured by the mass analyzer. With such procedures, the amount of FP gas accumulated in the bubbles at the crystal grain boundary and in the crystal grains at the radial position of the spent fuel pellet can be measured discriminately. Accordingly, the integrity of the fuel pellet can be recognized appropriately. (I.S.)

What factors mediate the relationship between global self-worth and weight and shape concerns?

Science.gov (United States)

Murphy, Edel; Dooley, Barbara; Menton, Aoife; Dolphin, Louise

2016-04-01

The primary aim of this study was to investigate whether the relationship between global self-worth and weight concerns and global self-worth and shape concerns was mediated by pertinent body image factors, while controlling for gender and estimated BMI. Participants were 775 adolescents (56% male) aged 12-18years (M=14.6; SD=1.50). Mediation analysis revealed a direct and a mediated effect between global self-worth and two body image models: 1) weight concerns and 2) shape concerns. The strongest mediators in both models were physical appearance, restrained eating, and depression. Partial mediation was observed for both models, indicating that body image factors which span cognitive, affective, and behavioral constructs, explain the association between global self-worth and weight and shape concerns. Implications for future research, weight and shape concern prevention and global self-worth enhancement programs are discussed. Copyright © 2016 Elsevier Ltd. All rights reserved.

Measurement of the Velocity and Pressure Drop in a Tubular Type Fuel

International Nuclear Information System (INIS)

Jonghark Park; Heetaek Chae; Cheol Park; Heonil Kim

2006-01-01

We have developed a tubular type fuel assembly design as one of candidates for fuel to be used in the Advanced HANARO Reactor (AHR). The tubular type fuel has several merits over a rod type fuel with respect to the thermal-hydraulic and structural safety; the larger ratio of surface area to volume makes the surface temperature of a fuel element become lower, and curved plate is stronger against longitudinal bending and vibration. In the other side, a disadvantage is expected such that the flow velocity can be distributed unevenly channel by channel because the flow channels are isolated from each other in a tubular type fuel assembly. In addition to the design development, we also investigated the flow characteristics of the tubular fuel experimentally. To examine the flow velocity distribution and pressure drop, we made an experiment facility and a mockup of the tubular fuel assembly. The fuel assembly consists of 6 concentric fuel tubes so that 7 layers are made between fuel tubes. Since each layer is divided into three sections by stiffeners, 21 isolated flow channels are made in total. We employed pitot-tubes to measure the coolant velocity in each channel. The maximum velocity was measured as large as about 28% of the average velocity. It was observed in the innermost channel contrarily to the expectation from the hydraulic diameter. A change in the total flow rate did not affect the flow distribution. Meanwhile, the pressure drop was measured as about 70% of the drop in the rod type fuel assembly in use in HANARO. (authors)

Instrument for measuring fuel cladding strain

International Nuclear Information System (INIS)

Billeter, T.R.

1976-01-01

Development work to provide instrumentation for the continuous measurement of strain of material specimens such as nuclear fuel cladding has shown that a microwave sensor and associated instrumentation hold promise. The cylindrical sensor body enclosing the specimen results in a coaxial resonator absorbing microwave energy at frequencies dependent upon the diameter of the specimen. Diametral changes of a microinch can be resolved with use of the instrumentation. Very reasonable values of elastic strain were measured at 75 0 F and 1000 0 F for an internally pressurized 20 percent C.W. 316 stainless steel specimen simulating nuclear fuel cladding. The instrument also indicated the creep strain of the same specimen pressurized at 6500 psi and at a temperature of 1000 0 F for a period of 700 hours. Although the indicated strain appears greater than actual, the sensor/specimen unit experienced considerable oxidation even though an inert gas purge persisted throughout the test duration. By monitoring at least two modes of resonance, the measured strain was shown to be nearly independent of sensor temperature. To prevent oxidation, a second test was performed in which the specimen/sensor units were contained in an evacuated enclosure. The strain of the two prepressurized specimens as indicated by the microwave instrumentation agreed very closely with pre- and post-test measurements obtained with use of a laser interferometer

Outlet temperature measurement correction of Gd fuel assemblies at Dukovany NPP

International Nuclear Information System (INIS)

Jurickova, M.

2008-01-01

In year 2006 we started data processing from the Dukovany NPP operating history database that contained data from the old measurement system VK3 and the new Scorpio-VVER. The work has been done in cooperation with the reactor physicists at Dukovany NPP. Obtained data from database were compared with calculated parameters from 3D diffusion macrocode Mobydick. During the data processing it was found that the Gd fuel assemblies have different time plot of measured assembly outlet temperature compared to the non-Gd fuel assemblies. Experimental studies in RRC KI found that there is insufficient coolant mixing in the region from the fuel bundle to the fuel assembly thermocouple. Due to this fact the thermocouple measure temperature is systematically higher than real temperature. There are two methods to solve this problem. The first method analyses the flow and heat transfer in the region from the fuel bundle to the fuel assembly thermocouple - this method is developed in Skoda JS. The second method statistically studies differences between the measured and calculated temperature by the Mobydick code using the operational history database. Our study is focused on the second method. Several calculation methods for the correction of measured assembly outlet temperature were developed. All correction methods were applied to the measured temperatures from the Dukovany NPP operating history database and the methods were mutually compared. In near future it is planned to compare results of our chosen correction method with modeling method, which is developing in Skoda JS and it is planned to validate both of them. Consequently, the one of these correction methods will be implemented in the modernized Scorpio-VVER for Dukovany NPP. (author)

Subcritical Measurements Research Program for Fresh and Spent Materials Test Reactor Fuels

International Nuclear Information System (INIS)

Blanchard, A.

1999-01-01

'A series of subcritical noise measurements were performed on fresh and spent University of Missouri Research Reactor fuel assemblies. These experimental measurements were performed for the purposes of providing benchmark quality data for validating transport theory computer codes and nuclear cross-section data used to perform criticality safety analyses for highly enriched, uranium-aluminum Material Test Reactor fuel assemblies. A mechanical test rig was designed and built to hold up to four fuel assemblies and neutron detectors in a subcritical array. The rig provided researchers with the ability to evaluate the reactivity effects of variable fuel/detector spacing, fuel rotation, and insertion of metal reflector plates into the lattice.'

Cooling flow measurement in fuel elements of the RA-6

International Nuclear Information System (INIS)

Brollo, F; Silin, N

2009-01-01

Under the UBERA6 project for the core change and power increase of the RA-6 reactor, the total coolant flow was increased to meet the requirements imposed by the new operating conditions. The flow through the fuel elements is an important parameter and is difficult to determine due to the geometric complexity of the core. To ensure safe operation of the reactor, adequate safety margins must be kept for all operating conditions. In the present work we performed the direct measurement of the cooling flow rate of a fuel in the reactor core, for which we used a turbine flowmeter built specifically for this use. This helped to confirm previous results obtained during the launch, made by an indirect method based on measuring the pressure difference of the core. The turbine flowmeter was chosen due to its robustness, ease of operation and low disturbance of the input stream to the fuel. We describe the calibration of this instrument and the results of flow measurements made on some of the RA6 reactor fuel elements under conditions of zero power. [es

Thermal Conductivity Measurement and Analysis of Fully Ceramic Microencapsulated fuel

International Nuclear Information System (INIS)

Lee, H. G.; Kim, D. J.; Park, J. Y.; Kim, W. J.; Lee, S. J.

2015-01-01

FCM nuclear fuel is composed of tristructural isotropic(TRISO) fuel particle and SiC ceramic matrix. SiC ceramic matrix play an essential part in protecting fission product. In the FCM fuel concept, fission product is doubly protected by TRISO coating layer and SiC ceramic matrix in comparison with the current commercial UO2 fuel system of LWR. In addition to a safety enhancement of FCM fuel, thermal conductivity of SiC ceramic matrix is better than that of UO2 fuel. Because the centerline temperature of FCM fuel is lower than that of the current UO2 fuel due to the difference of thermal conductivity of fuel, an operational release of fission products from the fuel can be reduced. SiC ceramic has attracted for nuclear fuel application due to its high thermal conductivity properties with good radiation tolerant properties, a low neutron absorption cross-section and a high corrosion resistance. Thermal conductivity of ceramic matrix composite depends on the thermal conductivity of each component and the morphology of reinforcement materials such as fibers and particles. There are many results about thermal conductivity of fiber-reinforced composite like as SiCf/SiC composite. Thermal conductivity of SiC ceramics and FCM pellets with the volume fraction of TRISO particles were measured and analyzed by analytical models. Polycrystalline SiC ceramics and FCM pellets with TRISO particles were fabricated by hot press sintering with sintering additives. Thermal conductivity of the FCM pellets with TRISO particles of 0 vol.%, 10 vol.%, 20 vol.%, 30 vol.% and 40 vol.% show 68.4, 52.3, 46.8, 43.0 and 34.5 W/mK, respectively. As the volume fraction of TRISO particles increased, the measured thermal conductivity values closely followed the prediction of Maxwell's equation

Flow measurement by Laser Doppler Anemometry in a nuclear fuel assembly

International Nuclear Information System (INIS)

Kehoe, A.

1984-12-01

Development of a Laser Doppler Anemometer measurement system and its operation are examined in this research. The system is designed for flow measurement in laboratory models of nuclear fuel assemblies. Use of the system is demonstrated by measuring turbulent velocity profiles in the laboratory model at full scale reactor flow rates. The reactors at the Savanah River Plant (SRP) are heavy water moderated and operate at low temperatures and pressures. Reactor power is currently limited by the temperature of the water in the nuclear fuel assembly. These temperature limits are conservatively calculated without allowing for any turbulent mixing. This research incorporates the design, fabriction and operation of a plexiglas model fuel assembly for the purpose of making turbulent velocity measurement via a Laser Doppler Anemometer System

Using fine-scale fuel measurements to assess wildland fuels, potential fire behavior and hazard mitigation treatments in the southeastern USA

International Nuclear Information System (INIS)

Ottmar, Roger D.; Blake, John I.; Crolly, William T.

2012-01-01

The inherent spatial and temporal heterogeneity of fuelbeds in forests of the southeastern United States may require fine scale fuel measurements for providing reliable fire hazard and fuel treatment effectiveness estimates. In a series of five papers, an intensive, fine scale fuel inventory from the Savanna River Site in the southeastern United States is used for building fuelbeds and mapping fire behavior potential, evaluating fuel treatment options for effectiveness, and providing a comparative analysis of landscape modeled fire behavior using three different data sources including the Fuel Characteristic Classification System, LANDFIRE, and the Southern Wildfire Risk Assessment. The research demonstrates that fine scale fuel measurements associated with fuel inventories repeated over time can be used to assess broad scale wildland fire potential and hazard mitigation treatment effectiveness in the southeastern USA and similar fire prone regions. Additional investigations will be needed to modify and improve these processes and capture the true potential of these fine scale data sets for fire and fuel management planning.

Poor Performance in Mathematics: Is There a Basis for a Self-Worth Explanation for Women?

Science.gov (United States)

Thompson, Ted; Dinnel, Dale L.

2007-01-01

The self-worth theory of achievement motivation holds that in situations in which poor performance is likely to reveal low ability, certain students (known as self-worth protective students) intentionally withdraw effort in order to avoid the negative implications of poor performance in terms of damage to self-worth. In this study, evidence of…

Development of examination technique for oxide layer thickness measurement of irradiated fuel rods

International Nuclear Information System (INIS)

Koo, D. S.; Park, S. W.; Kim, J. H.; Seo, H. S.; Min, D. K.; Kim, E. K.; Chun, Y. B.; Bang, K. S.

1999-06-01

Technique for oxide layer thickness measurement of irradiated fuel rods was developed to measure oxide layer thickness and study characteristic of fuel rods. Oxide layer thickness of irradiated fuels were measured, analyzed. Outer oxide layer thickness of 3 cycle-irradiated fuel rods were 20 - 30 μm, inner oxide layer thickness 0 - 10 μm and inner oxide layer thickness on cracked cladding about 30 μm. Oxide layer thickness of 4 cycle-irradiated fuel rods were about 2 times as thick as those of 1 cycle-irradiated fuel rods. Oxide layer on lower region of irradiated fuel rods was thin and oxide layer from lower region to upper region indicated gradual increase in thickness. Oxide layer thickness from 2500 to 3000 mm showed maximum and oxide layer thickness from 3000 to top region of irradiated fuel rods showed decreasing trend. Inner oxide layer thicknesses of 4 cycle-irradiated fuel rod were about 8 μm at 750 - 3500 mm from the bottom end of fuel rod. Outer oxide layer thickness were about 8 μm at 750 - 1000 mm from the bottom end of fuel rod. These indicated gradual increase up to upper region from the bottom end of fuel rod. These indicated gradual increase up to upper region from the bottom end of fuel. Oxide layer thickness technique will apply safety evaluation and study of reactor fuels. (author). 6 refs., 14 figs

Ground measurements of fuel and fuel consumption from experimental and operational prescribed fires at Eglin Air Force Base, Florida

Science.gov (United States)

Roger D. Ottmar; Robert E. Vihnanek; Clinton S. Wright; Andrew T. Hudak

2014-01-01

Ground-level measurements of fuel loading, fuel consumption, and fuel moisture content were collected on nine research burns conducted at Eglin Air Force Base, Florida in November, 2012. A grass or grass-shrub fuelbed dominated eight of the research blocks; the ninth was a managed longleaf pine (Pinus palustrus) forest. Fuel loading ranged from 1.7 Mg ha-1 on a...

Development of a reactivity worth correction scheme for the one-dimensional transient analysis

International Nuclear Information System (INIS)

Cho, J. Y.; Song, J. S.; Joo, H. G.; Kim, H. Y.; Kim, K. S.; Lee, C. C.; Zee, S. Q.

2003-11-01

This work is to develop a reactivity worth correction scheme for the MASTER one-dimensional (1-D) calculation model. The 1-D cross section variations according to the core state in the MASTER input file, which are produced for 1-D calculation performed by the MASTER code, are incorrect in most of all the core states except for exactly the same core state where the variations are produced. Therefore this scheme performs the reactivity worth correction factor calculations before the main 1-D transient calculation, and generates correction factors for boron worth, Doppler and moderator temperature coefficients, and control rod worth, respectively. These correction factors force the one dimensional calculation to generate the same reactivity worths with the 3-dimensional calculation. This scheme is applied to the control bank withdrawal accident of Yonggwang unit 1 cycle 14, and the performance is examined by comparing the 1-D results with the 3-D results. This problem is analyzed by the RETRAN-MASTER consolidated code system. Most of all results of 1-D calculation including the transient power behavior, the peak power and time are very similar with the 3-D results. In the MASTER neutronics computing time, the 1-D calculation including the correction factor calculation requires the negligible time comparing with the 3-D case. Therefore, the reactivity worth correction scheme is concluded to be very good in that it enables the 1-D calculation to produce the very accurate results in a few computing time

LWR spent-fuel radiochemical measurements and comparison with ORIGEN2 predictions

International Nuclear Information System (INIS)

Blahnik, D.E.; Jenquin, U.P.; Guenther, R.J.

1988-01-01

The Materials Characterization Center (MCC) at Pacific Northwest Laboratory is responsible for providing characterized spent fuel, designated approved testing material (ATMs) for subsequent use in the investigation of nuclear waste disposal forms by the US Department of Energy geologic repository project. The ATMs are selected to assure that test material is available that has a representative range of characteristics important to spent-fuel behavior in a geologic repository. Burnup and fission gas release were the primary criteria for selecting the ATMs. The five spent-fuel ATMs (ATM-101, -103, -104, -105, and -106) currently being characterized by the MCC have rod average burnups ranging from 20 to 43 MWd/kg M, fission gas releases ranging from 0.2 to 11.2%, and are from both boiling water reactors and pressurized water reactors. Radiochemical analyses of the fuel included measurements of 148 Nd (for fuel burnup), the isotopes of uranium and plutonium, and nuclides of importance to repository performance. Cladding samples were analyzed for 14 C. The measured values of selected nuclides were compared with values obtained from calculations with the ORIGEN2 code that was used to predict isotopic quantities for all of the ATMS. Ratios of the ORIGEN2 calculated values to the measured values for ATM-103 and ATM-106 fuel are given

Enhancing Student Self-Worth in the Primary School Learning Environment: Teachers' Views and Students' Views

Science.gov (United States)

Cushman, Penni; Cowan, Jackie

2010-01-01

This paper reports the findings from a study of teachers and students' views regarding self-worth in the primary school learning environment. The revised New Zealand curriculum recognises the importance of self-worth in students' motivation and ability to learn. While the need to enhance self-worth in the classroom has been well established in the…

Exposure to unwanted intrusions, neutralizing and their effects on self-worth and obsessive-compulsive phenomena.

Science.gov (United States)

Ahern, Claire; Kyrios, Michael; Meyer, Denny

2015-12-01

Although there is a growing body of literature to support the importance of understanding self processes in the experience of obsessive-compulsive disorder (OCD), no experimental research has directly examined the relationship between self-construals and phenomena central to OCD. The current study examined the effect that unwanted intrusions and neutralizing responses have on self-worth, distress and urge to neutralize. After listening to repeated audio recordings of idiosyncratic unwanted intrusions, a combined nonclinical and clinical OCD sample were asked to respond with either their chosen neutralizing strategy (experimental) or a refocus counting strategy (control). Each condition comprised of a 12-min responding period (respond) followed by an equivalent non-response period (listen). Participants completed each condition, and were randomly allocated into the condition completed first. Ratings of discomfort, urge to neutralize, and self-worth were measured throughout. Neutralizing and refocussing responses were both associated with decreases in discomfort and higher self-worth. The expected rebound effect for discomfort and urge to neutralize for the listen period after neutralizing was found. Methodological problems lead to missing data, although this was corrected with the use of Multi Level Modelling (MLM) analysis on a combined sample. The small clinical sample meant that comparison between the two populations was not possible. Findings support cognitive accounts that neutralizing is involved in the development and maintenance of OCD, and suggest that neutralizing is a purposeful response aimed to help reinstate self-worth. Implications and directions for future research are discussed. Copyright © 2015 Elsevier Ltd. All rights reserved.

The influence of social identity on self-worth, commitment, and effort in school-based youth sport.

Science.gov (United States)

Martin, Luc J; Balderson, Danny; Hawkins, Michael; Wilson, Kathleen; Bruner, Mark W

2018-02-01

​​​The current study examined the influence of social identity for individual perceptions of self-worth, commitment, and effort in school-based youth athletes. Using a prospective research design, 303 athletes (M age  = 14.89, SD = 1.77; 133 female) from 27 sport teams completed questionnaires at 2 time points (T1 - demographics, social identity; T2 - self-worth, commitment, effort) during an athletic season. Multilevel analyses indicated that at the individual level, the social identity dimension of in-group ties (IGT) predicted commitment (b = 0.12, P = .006) and perceived effort (b = 0.14, P = .008), whereas in-group affect (IGA) predicted commitment (b = 0.25, P = .001) and self-worth (b = 2.62, P = .006). At the team level, means for IGT predicted commitment (b = 0.31, P < .001) and self-worth (b = 4.76, P = .024). Overall, social identity accounted for variance at both levels, ranging from 4% (self-worth) to 15% (commitment). Identifying with a group to a greater extent was found to predict athlete perceptions of self-worth, commitment, and effort. More specifically, at the individual level, IGT predicted commitment and effort, and IGA predicted commitment and self-worth. At the team level, IGT predicted commitment and self-worth.

Measurement of burnup in FBR MOX fuel irradiated to high burnup

International Nuclear Information System (INIS)

Koyama, Shin-ichi; Osaka, Masahiko; Sekine, Takashi; Morozumi, Katsufumi; Namekawa, Takashi; Itoh, Masahiko

2003-01-01

The burnup of fuel pins in the subassemblies irradiated at the range from 0.003 to 13.28% FIMA in the JOYO MK-II core were measured by the isotope dilution analysis. For the measurement, 75 and 51 specimens were taken from the fuel pins of driver fuel and irradiation test subassemblies, respectively. The data of burnup could be obtained within an experimental error of 4%, and were compared with the ones calculated by 3-dimensional neutron diffusion codes MAGI and ESPRIT-J, which are used for JOYO core management system. Both data of burnup almost agree with each other within an error of 5%. For the fuel pins loaded at the outer region of the subassembly in the 4th row, which was adjacent to reflectors, however, some of the calculation results were 15% less at most than the measured values. It is suggested from the calculation by a Monte Carlo code MCNP-4A that this difference between the calculated and the measured data attribute from the softening of neutron flux in the region adjacent to the reflector. (author)

Validating criticality calculations for spent fuel with 252Cf-source-driven noise measurements

International Nuclear Information System (INIS)

Mihalczo, J.T.; Krass, A.W.; Valentine, T.E.

1992-01-01

The 252 Cf-Source-driven noise analysis method can be used for measuring the subcritical neutron multiplication factor k of arrays of spent light water reactor (LWR) fuel. This type of measurement provides a parameter that is directly related to the criticality state of arrays of LWR fuel. Measurements of this parameter can verify the criticality safety margins of spent LWR fuel configurations and thus could be a means of obtaining the information to justify burnup credit for spent LWR transportation/storage casks. The practicality of a measurement depends on the ability to install the hardware required to perform the measurement. Source chambers containing the 252 Cf at the required source intensity for this application have been constructed and have operated successfully for ∼10 years and can be fabricated to fit into control rod guide tubes of PWR fuel elements. Fission counters especially developed for spent-fuel measurements are available that would allow measurements of a special 3 x 3 spent fuel array and a typical burnup credit rail cask with spent fuel in unborated water. Adding a moderator around these fission counters would allow measurements with the typical burnup credit rail cask with borated water and the special 3 x 3 array with borated water. The recent work of Ficaro on modifying the KENO Va code to calculate by the Monte Carlo method the time sequences of pulses at two detectors near a fissile assembly from the fission chain multiplication process, initiated by a 252 Cf source in the assembly allows a direct computer calculation of the noise analysis data from this measurement method

The interpretation of fuel centre temperature measurements on a suspected leaking fuel pin

International Nuclear Information System (INIS)

Ainscough, J.B.; Lang, C.; Clough, D.J.

1983-01-01

In order to study fuel densification a series of single instrumented pin irradiations has been carried out in the High Pressure Water Loop of DIDO at Harwell. The behaviour of two of these pins was different from that expected. In the fifth test, where the fuel was 95% dense pellet UO 2 and expected to densify readily in-reactor, the fuel centre temperature increased from its starting value of approx. 1300 deg. C at a rate somewhat higher than expected on the basis of predicted densification rates. After about six days, the temperature increased rapidly and unexpectedly to 2100-2200 deg. C and remained steady at this level for a further eight days until a reactor trip occurred and the pin was unloaded. Predictions made using the HOTROD code imply a maximum fuel temperature of less than 1500 deg. C after densification. Post-irradiation examination confirmed that fission gas release had occurred, that the measured temperatures were consistent with the fuel microstructure and that the pin had a high internal gas pressure. The fourth pin in the series contained 97% dense UO 2 which was also expected to be dimensionally unstable. Qualitatively its behaviour was similar to that of the fifth pin though the temperatures throughout were lower. This pin experienced a number of major power cycles and failed after about 30 days in-reactor. It is probable that coolant ingress occurred in both pins via the thermocouple Hoke seal, degrading the filling gas conductivity and allowing the fuel to densify rapidly with consequent increase in the fuel/clad gap and hence in fuel temperature. These irradiations show that, for a short time at least, an apparently unfailed pin could operate undetected with temperatures significantly higher than those predicted for normal operation. (author)

Measurement of plutonium in spent nuclear fuel by self-induced x-ray fluorescence

Energy Technology Data Exchange (ETDEWEB)

Hoover, Andrew S [Los Alamos National Laboratory; Rudy, Cliff R [Los Alamos National Laboratory; Tobin, Steve J [Los Alamos National Laboratory; Charlton, William S [Los Alamos National Laboratory; Stafford, A [TEXAS A& M; Strohmeyer, D [TEXAS A& M; Saavadra, S [ORNL

2009-01-01

Direct measurement of the plutonium content in spent nuclear fuel is a challenging problem in non-destructive assay. The very high gamma-ray flux from fission product isotopes overwhelms the weaker gamma-ray emissions from plutonium and uranium, making passive gamma-ray measurements impossible. However, the intense fission product radiation is effective at exciting plutonium and uranium atoms, resulting in subsequent fluorescence X-ray emission. K-shell X-rays in the 100 keV energy range can escape the fuel and cladding, providing a direct signal from uranium and plutonium that can be measured with a standard germanium detector. The measured plutonium to uranium elemental ratio can be used to compute the plutonium content of the fuel. The technique can potentially provide a passive, non-destructive assay tool for determining plutonium content in spent fuel. In this paper, we discuss recent non-destructive measurements of plutonium X-ray fluorescence (XRF) signatures from pressurized water reactor spent fuel rods. We also discuss how emerging new technologies, like very high energy resolution microcalorimeter detectors, might be applied to XRF measurements.

Site-Based Budgeting in Fort Worth, Texas.

Science.gov (United States)

Peternick, Lauri; Sherman, Joel

1998-01-01

Examines the Fort Worth Independent School District's decentralized decision-making system through three lenses: a review of site-based decision-making procedures at several schools; an examination of who participates; and stakeholders' perceptions. Some schools operated democratically, significantly including teachers, parents, and community…

To seek work and worth

International Nuclear Information System (INIS)

Im, Yong Gyu

2010-07-01

It describes the documentary which shows US writers effect and process to seek worth though the work related nuclear power for half a century such as international nuclear school start of use of nuclear energy industry, establishment of nuclear society, by becoming a member of a standing committee and introduction of KINS, KANS and NSSC. It also describes his personal history about family and work and a brief summary of his career.

Case Study: A Picture Worth a Thousand Words? Making a Case for Video Case Studies

Science.gov (United States)

Pai, Aditi

2014-01-01

A picture, they say, is worth a thousand words. If a mere picture is worth a thousand words, how much more are "moving pictures" or videos worth? The author poses this not merely as a rhetorical question, but because she wishes to make a case for using videos in the traditional case study method. She recommends four main approaches of…

Self-Efficacy, Self-Worth and Stress

Science.gov (United States)

Flynn, Deborah M.; Chow, Peter

2017-01-01

One of the most stressful periods of life has been reported to be the time spent in the post secondary education system (Hales, 2009). As a result, researchers are interested in determining the various correlates associated with the successful coping during this time. It has been well established that self-esteem and self-worth are both factors…

Temperature and neutron dose rate measurements at a spent fuel shipping cask

International Nuclear Information System (INIS)

Krause, F.

1982-01-01

Apart from some other requirements, spent fuel shipping casks have to ensure sufficient heat removal and radiation shielding. Results of temperature and neutron dose rate measurements at a spent fuel shipping cask are presented for different loading and heat removal by air. The measurements show that in shipping higher burnup fuel assemblies neutron radiation has to be taken into account when estimating the shielding of the shipping cask. On the other hand, unallowable high temperatures have been observed neither at the fuel assemblies nor at the shipping cask for a maximum heat output of Q <= 12 kW. (author)

Heavy truck modeling for fuel consumption. Simulations and measurements

Energy Technology Data Exchange (ETDEWEB)

Sandberg, T.

2001-12-01

Fuel consumption for heavy trucks depends on many factors like roads, weather, and driver behavior that are hard for a manufacturer to influence. However, one design possibility is the power train configuration. Here a new simulation program for heavy trucks is created to find the configuration of the power train that gives the lowest fuel consumption for each transport task. For efficient simulations the model uses production code for speed and gear control, and it uses exchangeable data sets to allow simulation of the whole production range of engine types, on recorded road profiles from all over the world. Combined with a graphical user interface this application is called STARS (Scania Truck And Road Simulation). The forces of rolling resistance and air resistance in the model are validated through an experiment where the propeller shaft torque of a heavy truck is measured. It is found that the coefficient of rolling resistance is strongly dependent on tire temperature, not only on vehicle speed as expected. This led to the development of a new model for rolling resistance. The model includes the dynamic behavior of the tires and relates rolling resistance to tire temperature and vehicle speed. In another experiment the fuel consumption of a test truck in highway driving is measured. The altitude of the road is recorded with a barometer and used in the corresponding simulations. Despite of the limited accuracy of this equipment the simulation program manage to predict a level of fuel consumption only 2% lower than the real measurements. It is concluded that STARS is a good tool for predicting fuel consumption for trucks in highway driving and for comparing different power train configurations.

An Advanced Sodium-Cooled Fast Reactor Core Concept Using Uranium-Free Metallic Fuels for Maximizing TRU Burning Rate

Directory of Open Access Journals (Sweden)

Wuseong You

2017-12-01

Full Text Available In this paper, we designed and analyzed advanced sodium-cooled fast reactor cores using uranium-free metallic fuels for maximizing burning rate of transuranics (TRU nuclides from PWR spent fuels. It is well known that the removal of fertile nuclides such as 238U from fuels in liquid metal cooled fast reactor leads to the degradation of important safety parameters such as the Doppler coefficient, coolant void worth, and delayed neutron fraction. To resolve the degradation of the Doppler coefficient, we considered adding resonant nuclides to the uranium-free metallic fuels. The analysis results showed that the cores using uranium-free fuels loaded with tungsten instead of uranium have a significantly lower burnup reactivity swing and more negative Doppler coefficients than the core using uranium-free fuels without resonant nuclides. In addition, we considered the use of axially central B4C absorber region and moderator rods to further improve safety parameters such as sodium void worth, burnup reactivity swing, and the Doppler coefficient. The results of the analysis showed that the final design core can consume ~353 kg per cycle and satisfies self-controllability under unprotected accidents. The fuel cycle analysis showed that the PWR–SFR coupling fuel cycle option drastically reduces the amount of waste going to repository and the SFR burner can consume the amount of TRUs discharged from 3.72 PWRs generating the same electricity.

Development of an instantaneous local fuel-concentration measurement probe: an engine application

Science.gov (United States)

Guibert, P.; Boutar, Z.; Lemoyne, L.

2003-11-01

This work presents a new tool which can deliver instantaneous local measurements of fuel concentration in an engine cylinder with a high temporal resolution, particularly during compression strokes. Fuel concentration is represented by means of equivalence fuel-air ratio, i.e. the real engine mass ratio of fuel to air divided by the same ratio in ideal stoichiometry conditions. Controlling the mixture configuration for any strategy in a spark ignition engine and for auto-ignition combustion has a dominant effect on the subsequent processes of ignition, flame propagation and auto-ignition combustion progression, pollutant formation under lean or even stoichiometric operating conditions. It is extremely difficult, under a transient operation, to control the equivalence air/fuel ratio precisely at a required value and at the right time. This requires the development of a highly accurate equivalence air/fuel ratio control system and a tool to measure using crank angle (CA) resolution. Although non-intrusive laser techniques have considerable advantages, they are most of the time inappropriate due to their optical inaccessibility or the complex experimental set-up involved. Therefore, as a response to the demand for a relatively simple fuel-concentration measurement system a probe is presented that replaces a spark plug and allows the engine to run completely normally. The probe is based on hot-wire like apparatus, but involves catalytic oxidation at the wire surface. The development, characteristics and calibration of the probe are presented followed by applications to in-cylinder engine measurements.

Pressure drop ana velocity measurements in KMRR fuel rod bundles

International Nuclear Information System (INIS)

Yagn, Sun Kyu; Chung, Heung June; Chung, Chang Whan; Chun, Se Young; Song, Chul Wha; Won, Soon Yeun; Chung, Moon Ki

1990-01-01

The detailed hydraulic characteristic measurements in subchannels of longitudinally finned rod bundles using one-component LDV(Laser Doppler Velocimeter) were performed. Time mean axial velocity, turbulent intensity, and turbulent micro scales, such as time auto-correlation, Eulerian integral and micro scale, Kolmogorov length and time scale, and Taylor micro length scale were measured. The signals from LDV are inherently more or less discontinuous. The spectra of signals having such intermittent defects can be obtained by the fast Fourier transformation (FFT) of the auto-correlation function. The turbulent crossflow mixing rate between neighboring subchannels and dominant frequencies were evaluated from the measured data. Pressure drop data were obtained for the typical 36-element and 18-element fuel rod bundles fabricated by the design requirement of KMRR fuel and for other type of fuels assembled with 6-fin rods to investigate the fin effects on the pressure drop characteristics

Ten Ideas Worth Stealing from New Zealand.

Science.gov (United States)

Jarchow, Elaine

1992-01-01

New Zealand educators have some ideas worth stealing, including morning tea-time, the lie-flat manifold duplicate book for recording classroom observation comments, school uniforms, collegial planning and grading of college assignments, good meeting etiquette, a whole-child orientation, portable primary architecture, group employment interviews…

Using fine-scale fuel measurements to assess wildland fuels, potential fire behavior and hazard mitigation treatments in the southeastern USA

Science.gov (United States)

Roger D. Ottmar; John I. Blake; William T. Crolly

2012-01-01

The inherent spatial and temporal heterogeneity of fuel beds in forests of the southeastern United States may require fine scale fuel measurements for providing reliable fire hazard and fuel treatment effectiveness estimates. In a series of five papers, an intensive, fine scale fuel inventory from the Savanna River Site in the southeastern United States is used for...

Results from Mobile Lab Measurements Obtained in the Barnett Shale with Emphasis on Methane and Gaseous Mercury Emissions (Fort Worth, TX)

Science.gov (United States)

Laine, P. L.; Lan, X.; Anderson, D.; Talbot, R. W.

2013-12-01

Our work is part of a comprehensive analysis conducted through a collaboration of ground based measurements and airborne measurements with several research groups in order to gain a better understanding of methane and mercury emissions in the Barnett Shale. It's a vast rock formation that sits in the 5,000 square miles surrounding the Fort Worth area. To get the gas to market requires an underground highway of pipelines and compression stations. Texas state records show that since 2000 the number of gas compressors in the Barnett Shale has tripled (from a few hundred to 1,300), and they're ever infringing on populated areas. Recent preliminary data reported by Pétron et al. and Tollefson et al. (from the natural-gas operations in Denver-Julesburg Basin) point to CH4 loss from the process of 4-8%, not including additional losses in the pipeline and distribution system. Additionally, Howarth et al. have conducted a comprehensive analysis of greenhouse gases (methane, in particular) emitted from shale gas as a result of hydraulic fracturing and they estimate up to 8% of all natural gas mined from shale formations leaks to the atmosphere. Not only is this cause for alarm due to the global warming potential of methane, but we would expect similar losses of additional (potentially harmful) gases, i.e., atmospheric Hg, from the extraction systems. These preliminary findings are higher than the current U.S. Environment Protection Agency (EPA) leakage estimate of 2.3 percent. Our strategy employs the use of our mobile laboratory, a four door Chevrolet Silverado pickup truck with a camper shell, outfitted with trace gas instrumentation including a Picarro G2132i and a Tekran 2537 mercury analyzer. The Picarro cavity ring down instrument has high precision and accuracy H2O, CO2, CH4, and 13δC in CH4 and CO2 with very little drift due to precise temperature and pressure controls. The Tekran mercury analyzer allows for accurate total gaseous mercury measurements via

The underwater coincidence counter for plutonium measurements in mixed-oxide fuel assemblies manual

International Nuclear Information System (INIS)

Eccleston, G.W.; Menlove, H.O.; Abhold, M.; Baker, M.; Pecos, J.

1999-01-01

This manual describes the Underwater Coincidence Counter (UWCC) that has been designed for the measurement of plutonium in mixed-oxide (MOX) fuel assemblies prior to irradiation. The UWCC uses high-efficiency 3 He neutron detectors to measure the spontaneous-fission and induced-fission rates in the fuel assembly. Measurements can be made on MOX fuel assemblies in air or underwater. The neutron counting rate is analyzed for singles, doubles, and triples time correlations to determine the 240 Pu effective mass per unit length of the fuel assembly. The system can verify the plutonium loading per unit length to a precision of less than 1% in a measurement time of 2 to 3 minutes. System design, components, performance tests, and operational characteristics are described in this manual

An empirical evaluation of two theoretically-based hypotheses on the directional association between self-worth and hope.

Science.gov (United States)

McDavid, Lindley; McDonough, Meghan H; Smith, Alan L

2015-06-01

Fostering self-worth and hope are important goals of positive youth development (PYD) efforts, yet intervention design is complicated by contrasting theoretical hypotheses regarding the directional association between these constructs. Therefore, within a longitudinal design we tested: (1) that self-worth predicts changes in hope (self theory; Harter, 1999), and (2) that hope predicts changes in self-worth (hope theory; Snyder, 2002) over time. Youth (N = 321; Mage = 10.33 years) in a physical activity-based PYD program completed surveys 37-45 days prior to and on the second day and third-to-last day of the program. A latent variable panel model that included autoregressive and cross-lagged paths indicated that self-worth was a significant predictor of change in hope, but hope did not predict change in self-worth. Therefore, the directional association between self-worth and hope is better explained by self-theory and PYD programs should aim to enhance perceptions of self-worth to build perceptions of hope. Copyright © 2015 The Foundation for Professionals in Services for Adolescents. Published by Elsevier Ltd. All rights reserved.

The Measurement of Fuel-Air Ratio by Analysis for the Oxidized Exhaust Gas

Science.gov (United States)

Gerrish, Harold C.; Meem, J. Lawrence, Jr.

1943-01-01

An investigation was made to determine a method of measuring fuel-air ratio that could be used for test purposes in flight and for checking conventional equipment in the laboratory. Two single-cylinder test engines equipped with typical commercial engine cylinders were used. The fuel-air ratio of the mixture delivered to the engines was determined by direct measurement of the quantity of air and of fuel supplied and also by analysis of the oxidized exhaust gas and of the normal exhaust gas. Five fuels were used: gasoline that complied with Army-Navy fuel Specification No. AN-VV-F-781 and four mixtures of this gasoline with toluene, benzene, and xylene. The method of determining the fuel-air ratio described in this report involves the measurement of the carbon-dioxide content of the oxidized exhaust gas and the use of graphs for the presented equation. This method is considered useful in aircraft, in the field, or in the laboratory for a range of fuel-air ratios from 0.047 to 0.124.

The Measurement of Fuel-air Ratio by Analysis of the Oxidized Exhaust Gas

Science.gov (United States)

Memm, J. Lawrence, Jr.

1943-01-01

An investigation was made to determine a method of measuring fuel-air ratio that could be used for test purposes in flight and for checking conventional equipment in the laboratory. Two single-cylinder test engines equipped with typical commercial engine cylinders were used. The fuel-air ratio of the mixture delivered to the engines was determined by direct measurement of the quantity of air and of fuel supplied and also by analysis of the oxidized exhaust gas and of the normal exhaust gas. Five fuels were used: gasoline that complied with Army-Navy Fuel Specification, No. AN-VV-F-781 and four mixtures of this gasoline with toluene, benzene, and xylene. The method of determining the fuel-air ratio described in this report involves the measurement of the carbon-dioxide content of the oxidized exhaust gas and the use of graphs or the presented equation. This method is considered useful in aircraft, in the field, or in the laboratory for a range of fuel-air ratios from 0.047 to 0.124

Uncertainty of Doppler reactivity worth due to uncertainties of JENDL-3.2 resonance parameters

Energy Technology Data Exchange (ETDEWEB)

Zukeran, Atsushi [Hitachi Ltd., Hitachi, Ibaraki (Japan). Power and Industrial System R and D Div.; Hanaki, Hiroshi; Nakagawa, Tuneo; Shibata, Keiichi; Ishikawa, Makoto

1998-03-01

Analytical formula of Resonance Self-shielding Factor (f-factor) is derived from the resonance integral (J-function) based on NR approximation and the analytical expression for Doppler reactivity worth ({rho}) is also obtained by using the result. Uncertainties of the f-factor and Doppler reactivity worth are evaluated on the basis of sensitivity coefficients to the resonance parameters. The uncertainty of the Doppler reactivity worth at 487{sup 0}K is about 4 % for the PNC Large Fast Breeder Reactor. (author)

Assessment of DUPIC fuel compatibility with CANDU-6

Energy Technology Data Exchange (ETDEWEB)

Choi, H B; Roh, G H; Jeong, C J; Rhee, B W; Choi, J W [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1997-07-01

The compatibility of DUPIC fuel with the existing CANDU reactor was assessed. The technical issues of DUPIC fuel compatibility were chosen based on the CANDU physics design requirements and inherent characteristics of DUPIC fuel. The compatibility was assessed for the reference DUPIC fuel composition which was determined to reduce the composition heterogeneity and improve the spent PWR fuel utilization. Preliminary studies on a CANDU core loaded with DUPIC fuel have shown that the nominal power distribution is flatter than that of a natural uranium core when a 2-bundle shift refueling scheme is used, which reduces the reactivity worths of devices in the core and, therefore, the performance of reactivity devices was assessed. The safety of the core was assessed by a LOCA simulation and it was found that the power pulse upon LOCA can be maintained below that in the natural uranium core when a poison material is used in the DUPIC fuel. For the feasibility of handling DUPIC fuel in the plant, it will be necessary to introduce new equipment to load the DUPIC fuel in the refueling magazine. The radiation effect of DUPIC fuel on both the reactor hardware and the environment will require a quantitative analysis later. (author).

Measurement of gamma ray from fuel of high temperature engineering test reactor. Method of measurement and results

Energy Technology Data Exchange (ETDEWEB)

Fujimoto, Nozomu; Nojiri, Naoki; Takada, Eiji [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment] [and others

2001-02-01

To obtain information in the HTTR core directly, gamma ray from fuel blocks was measured when fuel blocks were discharged from the core and reloaded to the core. Gamma ray was measured using GM detector, CZT semiconductor detector installed in a door valve and area monitors installed in a stand pipe compartment. The measurement was carried out for 20 fuel blocks in 4 columns considering the symmetry of uranium enrichment distribution in the core. Relative axial distribution in each column obtained by the GM detector and CZT detector agreed with calculated results. However, calculation values showed higher values than measured values in upper region of the core, lower those in lower region of the core. The axial distributions were also evaluated by the area monitors. The measured values agreed with calculated values. It became clear that it was possible to obtain the data inside the core by this method. (author)

Exercise effects on depressive symptoms and self-worth in overweight children: a randomized controlled trial.

Science.gov (United States)

Petty, Karen H; Davis, Catherine L; Tkacz, Joseph; Young-Hyman, Deborah; Waller, Jennifer L

2009-10-01

To test the dose-response effects of an exercise program on depressive symptoms and self-worth in children. Overweight, sedentary children (N = 207, 7-11 years, 58% male, 59% Black) were randomly assigned to low or high dose (20 or 40 min/day) aerobic exercise programs (13 +/- 1.6 weeks), or control group. Children completed the Reynolds Child Depression Scale and Self-Perception Profile for Children at baseline and posttest. A dose-response benefit of exercise was detected for depressive symptoms. A race x group interaction showed only White children's global self-worth (GSW) improved. There was some evidence that increased self-worth mediated the effect on depressive symptoms. This study shows dose-response benefits of exercise on depressive symptoms and self-worth in children. However, Blacks did not show increased GSW in response to the intervention. Results provide some support for mediation of the effect of exercise on depressive symptoms via self-worth.

Exercise Effects on Depressive Symptoms and Self-Worth in Overweight Children: A Randomized Controlled Trial*

Science.gov (United States)

Petty, Karen H.; Tkacz, Joseph; Young-Hyman, Deborah; Waller, Jennifer L.

2009-01-01

Objective To test the dose–response effects of an exercise program on depressive symptoms and self-worth in children. Method Overweight, sedentary children (N = 207, 7–11 years, 58% male, 59% Black) were randomly assigned to low or high dose (20 or 40 min/day) aerobic exercise programs (13 ± 1.6 weeks), or control group. Children completed the Reynolds Child Depression Scale and Self-Perception Profile for Children at baseline and posttest. Results A dose–response benefit of exercise was detected for depressive symptoms. A race × group interaction showed only White children's global self-worth (GSW) improved. There was some evidence that increased self-worth mediated the effect on depressive symptoms. Conclusions This study shows dose–response benefits of exercise on depressive symptoms and self-worth in children. However, Blacks did not show increased GSW in response to the intervention. Results provide some support for mediation of the effect of exercise on depressive symptoms via self-worth. PMID:19223278

An improved characterization method for international accountancy measurements of fresh and irradiated mixed oxide (MOX) fuel: helping achieve continual monitoring and safeguards through the fuel cycle

Energy Technology Data Exchange (ETDEWEB)

Evans, Louise G [Los Alamos National Laboratory; Croft, Stephen [Los Alamos National Laboratory; Swinhoe, Martyn T [Los Alamos National Laboratory; Tobin, S. J. [Los Alamos National Laboratory; Menlove, H. O. [Los Alamos National Laboratory; Schear, M. A. [Los Alamos National Laboratory; Worrall, Andrew [U.K. NNL

2011-01-13

Nuclear fuel accountancy measurements are conducted at several points through the nuclear fuel cycle to ensure continuity of knowledge (CofK) of special nuclear material (SNM). Non-destructive assay (NDA) measurements are performed on fresh fuel (prior to irradiation in a reactor) and spent nuclear fuel (SNF) post-irradiation. We have developed a fuel assembly characterization system, based on the novel concept of 'neutron fingerprinting' with multiplicity signatures to ensure detailed CofK of nuclear fuel through the entire fuel cycle. The neutron fingerprint in this case is determined by the measurement of the various correlated neutron signatures, specific to fuel isotopic composition, and therefore offers greater sensitivity to variations in fissile content among fuel assemblies than other techniques such as gross neutron counting. This neutron fingerprint could be measured at the point of fuel dispatch (e.g. from a fuel fabrication plant prior to irradiation, or from a reactor site post-irradiation), monitored during transportation of the fuel assembly, and measured at a subsequent receiving site (e.g. at the reactor site prior to irradiation, or reprocessing facility post-irradiation); this would confirm that no unexpected changes to the fuel composition or amount have taken place during transportation and/ or reactor operations. Changes may indicate an attempt to divert material for example. Here, we present the current state of the practice of fuel measurements for both fresh mixed oxide (MOX) fuel and SNF (both MOX and uranium dioxide). This is presented in the framework of international safeguards perspectives from the US and UK. We also postulate as to how the neutron fingerprinting concept could lead to improved fuel characterization (both fresh MOX and SNF) resulting in: (a) assured CofK of fuel across the nuclear fuel cycle, (b) improved detection of SNM diversion, and (c) greater confidence in safeguards of SNF transportation.

Ignition delay time measurements of primary reference fuel blends

KAUST Repository

Alabbad, Mohammed

2017-02-07

Ignition delay times of four different primary reference fuels (PRF), mixtures of n-heptane and iso-octane, were measured behind reflected shock waves in a high-pressure shock tube facility. The PRFs were formulated to match the RON of two high-octane gasolines (RON 95 and 91) and two prospective low-octane naphtha fuels (RON 80 and 70). Experiments were carried out over a wide range of temperatures (700–1200K), pressures (10, 20, and 40bar) and equivalence ratios (0.5 and 1). Kinetic modeling predictions from four chemical kinetic mechanisms are compared with the experimental data. Ignition delay correlations are developed to reproduce the measured ignition delay times. Brute force sensitivity analyses are carried out to identify reactions that affect ignition delay times at specific temperature, pressure and equivalence ratio. The large experimental data set provided in the current work will serve as a benchmark for the validation of chemical kinetic mechanisms of primary reference fuel blends.

Ignition delay time measurements of primary reference fuel blends

KAUST Repository

Alabbad, Mohammed; Javed, Tamour; Khaled, Fathi; Badra, Jihad; Farooq, Aamir

2017-01-01

Ignition delay times of four different primary reference fuels (PRF), mixtures of n-heptane and iso-octane, were measured behind reflected shock waves in a high-pressure shock tube facility. The PRFs were formulated to match the RON of two high-octane gasolines (RON 95 and 91) and two prospective low-octane naphtha fuels (RON 80 and 70). Experiments were carried out over a wide range of temperatures (700–1200K), pressures (10, 20, and 40bar) and equivalence ratios (0.5 and 1). Kinetic modeling predictions from four chemical kinetic mechanisms are compared with the experimental data. Ignition delay correlations are developed to reproduce the measured ignition delay times. Brute force sensitivity analyses are carried out to identify reactions that affect ignition delay times at specific temperature, pressure and equivalence ratio. The large experimental data set provided in the current work will serve as a benchmark for the validation of chemical kinetic mechanisms of primary reference fuel blends.

Description of the PIE facility for research reactors irradiated fuels in CNEA

International Nuclear Information System (INIS)

Bisca, A.; Coronel, R.; Homberger, V.; Quinteros, A.; Ratner, M.

2002-01-01

The PIE Facility (LAPEP), located at the Ezeiza Atomic Center (CAE), was designed to carry out destructive and non-destructive post-irradiation examinations (PIE) on research and power reactor spent fuels, reactor internals and other irradiated materials, and to perform studies related with: Station lifetime extension; Fuel performance; Development of new fuels; and Failures and determination of their causes. LAPEP is a relevant facility where research and development can be carried out. It is worth mentioning that in this facility the PIE corresponding to the Surveillance Program for the Atucha I Nuclear Power Plant (CNA-1) were successfully performed. Materials testing during the CNA-1 repair and the study of failures in fuel element plugs of the Embalse Nuclear Power Plant (CNE) were also performed. (author)

An environmental impact measure for nuclear fuel cycle evaluation

International Nuclear Information System (INIS)

Ahn, Joonhong

2004-01-01

Review of the models and measures for repository performance assessment has revealed that dedicated measures for environmental impacts need to be developed for the purpose of nuclear-fuel-cycle evaluation from the viewpoint of environmental impact minimization. The present study proposes the total toxicity index of released radionuclides that have accumulated in the region exterior to the repository as an environmental impact measure. The measure is quantitatively evaluated by a radionuclide transport model that incorporates the effects of canister-array configuration and the initial mass loading in the waste canister. With the measure, it is demonstrated that the environmental impact of the repository can be effectively reduced by reduction of the initial mass loading and change in the canister-array configuration in the repository. Environmental impacts of the mill tailings and the depleted uranium are as important as those from the high-level radioactive wastes repository. For a fair comparison of various fuel cycles, the sum of these impacts should be compared. (author)

Cancer-risk benefits of clean fuel technology and policy. A statistical analysis

Energy Technology Data Exchange (ETDEWEB)

Gallagher, Paul [Economics Department, 481 Heady Hall, Iowa State University, Ames, IA 50011 (United States); Lazarus, William [Applied Economics Department, 253 COB, University of Minnesota, St. Paul, MN 55455 (United States); Shapouri, Hosein; Conway, Roger; Duffield, James [Office of Energy Policy and New Uses, 400 Independence Avenue, SW Rm. 4059 So. Bldg, United States Department of Agriculture, WA, DC 20250 (United States)

2009-12-15

The hypothesis of this study is that there is a statistical relationship between the lung cancer mortality rate and the intensity of fuel consumption (measured in gallons/square mile) at a particular location. We estimate cross-section regressions of the mortality rate due to lung cancer against the intensity of fuel consumption using local data for the entire US, before the US Clean Air Act (CAA) in 1974 and after the most recent policy revisions in 2004. The cancer rate improvement estimate suggests that up to 10 lung cancer deaths per 100,000 residents are avoided in the largest urban areas with highest fuel consumption per square mile. In New York City, for instance, the mortality reduction may be worth about USD 5.7 billion annually. Across the US, the estimated value of statistical life (VSL) benefit is USD 27.2 billion annually. There are likely three inseparable reasons that contributed importantly to this welfare improvement. First, the CAA regulations mandated reduction in specific carcinogenic chemicals or smog components. Second, technologies such as the catalytic converter (CC) and low-particulate diesel engine were adopted. Third, biofuels have had important roles, making the adoption of clean air technology possible and substituting for high emission fuels. (author)

Cancer-risk benefits of clean fuel technology and policy: A statistical analysis

Energy Technology Data Exchange (ETDEWEB)

Gallagher, Paul, E-mail: paulg@iastate.ed [Economics Department, 481 Heady Hall, Iowa State University, Ames, IA 50011 (United States); Lazarus, William [Applied Economics Department, 253 COB, University of Minnesota, St. Paul, MN 55455 (United States); Shapouri, Hosein; Conway, Roger; Duffield, James [Office of Energy Policy and New Uses, 400 Independence Avenue, SW (Rm. 4059 So. Bldg), United States Department of Agriculture, WA, DC 20250 (United States)

2009-12-15

The hypothesis of this study is that there is a statistical relationship between the lung cancer mortality rate and the intensity of fuel consumption (measured in gallons/square mile) at a particular location. We estimate cross-section regressions of the mortality rate due to lung cancer against the intensity of fuel consumption using local data for the entire US, before the US Clean Air Act (CAA) in 1974 and after the most recent policy revisions in 2004. The cancer rate improvement estimate suggests that up to 10 lung cancer deaths per 100,000 residents are avoided in the largest urban areas with highest fuel consumption per square mile. In New York City, for instance, the mortality reduction may be worth about $5.7 billion annually. Across the US, the estimated value of statistical life (VSL) benefit is $27.2 billion annually. There are likely three inseparable reasons that contributed importantly to this welfare improvement. First, the CAA regulations mandated reduction in specific carcinogenic chemicals or smog components. Second, technologies such as the catalytic converter (CC) and low-particulate diesel engine were adopted. Third, biofuels have had important roles, making the adoption of clean air technology possible and substituting for high emission fuels.

Cancer-risk benefits of clean fuel technology and policy: A statistical analysis

International Nuclear Information System (INIS)

Gallagher, Paul; Lazarus, William; Shapouri, Hosein; Conway, Roger; Duffield, James

2009-01-01

The hypothesis of this study is that there is a statistical relationship between the lung cancer mortality rate and the intensity of fuel consumption (measured in gallons/square mile) at a particular location. We estimate cross-section regressions of the mortality rate due to lung cancer against the intensity of fuel consumption using local data for the entire US, before the US Clean Air Act (CAA) in 1974 and after the most recent policy revisions in 2004. The cancer rate improvement estimate suggests that up to 10 lung cancer deaths per 100,000 residents are avoided in the largest urban areas with highest fuel consumption per square mile. In New York City, for instance, the mortality reduction may be worth about $5.7 billion annually. Across the US, the estimated value of statistical life (VSL) benefit is $27.2 billion annually. There are likely three inseparable reasons that contributed importantly to this welfare improvement. First, the CAA regulations mandated reduction in specific carcinogenic chemicals or smog components. Second, technologies such as the catalytic converter (CC) and low-particulate diesel engine were adopted. Third, biofuels have had important roles, making the adoption of clean air technology possible and substituting for high emission fuels.

Configurations of actual and perceived motor competence among children: Associations with motivation for sports and global self-worth.

Science.gov (United States)

Bardid, Farid; De Meester, An; Tallir, Isabel; Cardon, Greet; Lenoir, Matthieu; Haerens, Leen

2016-12-01

The present study used a person-centred approach to examine whether different profiles based on actual and perceived motor competence exist in elementary school children. Multilevel regression analyses were conducted to explore how children with different motor competence-based profiles might differ in their autonomous motivation for sports and global self-worth. Validated questionnaires were administered to 161 children (40% boys; age=8.82±0.66years) to assess their perceived motor competence, global self-worth, and motivation for sports. Actual motor competence was measured with the Körperkoordinationstest für Kinder. Cluster analyses identified four motor competence-based profiles: two groups were characterized by corresponding levels of actual and perceived motor competence (i.e., low-low and high-high) and two groups were characterized by divergent levels of actual and perceived motor competence (i.e., high-low and low-high). Children in the low-low and high-low group displayed significantly lower levels of autonomous motivation for sports and lower levels of global self-worth than children in the low-high and high-high group. These findings emphasize that fostering children's perceived motor competence might be crucial to improve their motivation for sports and their global self-worth. Teachers and instructors involved in physical education and youth sports should thus focus on both actual and perceived motor competence. Copyright © 2016 Elsevier B.V. All rights reserved.

How Does Students’ Sense of Self-Worth Influence Their Goal Orientation in Mathematics Achievement?

Directory of Open Access Journals (Sweden)

Gulseren Sekreter

2017-12-01

Full Text Available In learning mathematics, students are naturally motivated to protect their self-worth by maintaining a belief that they are competent in this area. However, there is an important question which educators have to answer: “Why do students often confuse ability with worth?” The most important reason is that in our society students are widely considered to be worthy according to their ability to achieve in the given tasks in mathematics. Irrespective the contributions of the Multiple Intelligence Theory of intelligence in education, unfortunately mathematics is still regarded as predicting students’ overall ability to learn. Educators should realize that the need in order to protect self-worth arises primarily from fear of failure. Therefore, if this fear of failure is strong, some students will not try and gradually they will produce failure- avoiding strategies to avoid certain tasks in order not to look bad or receive negative assessments from others to protect his/her self-worth. It is important to make sure that the performance goals do not promote failure-avoidance (performance-avoidance-oriented behavior, such as avoiding unfavorable judgments of capabilities and looking incompetent when the student encounters greater challenges. The main purpose of this qualitative study, therefore, is to explore students’ achievement goal motivation, their self-worth and how these motivational factors impact their learning goals in mathematics. This study hypothesizes that self- worth protection in math has also been considered from a performance-avoidance goal viewpoint.This study emphasizes that educators, who consider true self-worth as the student’s inherent value, should avoid comparing their students’ ability, capability relative to others as well as students’ academic performance and outcomes with others in class context.

Experimental Measurement and Numerical Modeling of the Effective Thermal Conductivity of TRISO Fuel Compacts

International Nuclear Information System (INIS)

Folsom, Charles

2015-01-01

Accurate modeling capability of thermal conductivity of tristructural-isotropic (TRISO) fuel compacts is important to fuel performance modeling and safety of Generation IV reactors. To date, the effective thermal conductivity (ETC) of tristructural-isotropic (TRISO) fuel compacts has not been measured directly. The composite fuel is a complicated structure comprised of layered particles in a graphite matrix. In this work, finite element modeling is used to validate an analytic ETC model for application to the composite fuel material for particle-volume fractions up to 40%. The effect of each individual layer of a TRISO particle is analyzed showing that the overall ETC of the compact is most sensitive to the outer layer constituent. In conjunction with the modeling results, the thermal conductivity of matrix-graphite compacts and the ETC of surrogate TRISO fuel compacts have been successfully measured using a previously developed measurement system. The ETC of the surrogate fuel compacts varies between 50-30 W m -1 K -1 over a temperature range of 50-600°C. As a result of the numerical modeling and experimental measurements of the fuel compacts, a new model and approach for analyzing the effect of compact constituent materials on ETC is proposed that can estimate the fuel compact ETC with approximately 15-20% more accuracy than the old method. Using the ETC model with measured thermal conductivity of the graphite matrix-only material indicate that, in the composite form, the matrix material has a much greater thermal conductivity, which is attributed to the high anisotropy of graphite thermal conductivity. Therefore, simpler measurements of individual TRISO compact constituents combined with an analytic ETC model, will not provide accurate predictions of overall ETC of the compacts emphasizing the need for measurements of composite, surrogate compacts.

Changes in academic adjustment and relational self-worth across the transition to middle school.

Science.gov (United States)

Ryan, Allison M; Shim, Sungok Serena; Makara, Kara A

2013-09-01

Moving from elementary to middle school is a time of great transition for many early adolescents. The present study examined students' academic adjustment and relational self-worth at 6-month intervals for four time points spanning the transition from elementary school to middle school (N = 738 at time 1; 53 % girls; 54 % African American, 46 % European American). Grade point average (G.P.A.), intrinsic value for schoolwork, self-worth around teachers, and self-worth around friends were examined at every time point. The overall developmental trajectory indicated that G.P.A. and intrinsic value for schoolwork declined. The overall decline in G.P.A. was due to changes at the transition and across the first year in middle school. Intrinsic value declined across all time points. Self-worth around teachers was stable. The developmental trends were the same regardless of gender or ethnicity except for self-worth around friends, which was stable for European American students and increased for African American students due to an ascent at the transition into middle school. Implications for the education of early adolescents in middle schools are discussed.

Validity of the Worth 4 Dot Test in Patients with Red-Green Color Vision Defect.

Science.gov (United States)

Bak, Eunoo; Yang, Hee Kyung; Hwang, Jeong-Min

2017-05-01

The Worth four dot test uses red and green glasses for binocular dissociation, and although it has been believed that patients with red-green color vision defects cannot accurately perform the Worth four dot test, this has not been validated. Therefore, the purpose of this study was to demonstrate the validity of the Worth four dot test in patients with congenital red-green color vision defects who have normal or abnormal binocular vision. A retrospective review of medical records was performed on 30 consecutive congenital red-green color vision defect patients who underwent the Worth four dot test. The type of color vision anomaly was determined by the Hardy Rand and Rittler (HRR) pseudoisochromatic plate test, Ishihara color test, anomaloscope, and/or the 100 hue test. All patients underwent a complete ophthalmologic examination. Binocular sensory status was evaluated with the Worth four dot test and Randot stereotest. The results were interpreted according to the presence of strabismus or amblyopia. Among the 30 patients, 24 had normal visual acuity without strabismus nor amblyopia and 6 patients had strabismus and/or amblyopia. The 24 patients without strabismus nor amblyopia all showed binocular fusional responses by seeing four dots of the Worth four dot test. Meanwhile, the six patients with strabismus or amblyopia showed various results of fusion, suppression, and diplopia. Congenital red-green color vision defect patients of different types and variable degree of binocularity could successfully perform the Worth four dot test. They showed reliable results that were in accordance with their estimated binocular sensory status.

Measurements of decay heat and gamma-ray intensity of spent LWR fuel assemblies

International Nuclear Information System (INIS)

Vogt, J.; Agrenius, L.; Jansson, P.; Baecklin, A.; Haakansson, A.; Jacobsson, S.

1999-01-01

Calorimetric measurements of the decay heat of a number of BWR and PWR fuel assemblies have been performed in the pools at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel, CLAB. Gamma-ray measurements, using high-resolution gamma-ray spectroscopy (HRGS), have been carried out on the same fuel assemblies in order to test if it is possible to find a simple and accurate correlation between the 137 CS -intensity and the decay heat for fuel with a cooling time longer than 10-12 years. The results up to now are very promising and may ultimately lead to a qualified method for quick and accurate determination of the decay heat of old fuel by gamma-ray measurements. By means of the gamma spectrum the operator declared data on burnup, cooling time and initial enrichment can be verified as well. CLAB provides a unique opportunity in the world to follow up the decay heat of individual fuel assemblies during several decades to come. The results will be applicable for design and operation of facilities for wet and dry interim storage and subsequent encapsulation for final disposal of the fuel. (author)

Accuracy of dimension measurements from neutron radiographs of nuclear fuel pins

International Nuclear Information System (INIS)

Domanus, J.C.

1976-01-01

A comparison is given of accuracies obtained with measuring the dimensions (pellet diameter and fuel-clad gap) from neutron and X-ray radiographs of a calibrated nuclear fuel pin performed with a projection microscope, microdensitometers and a video micrometer

75 FR 39621 - Proposed Information Collection (Income-Net Worth and Employment Statement) Activity: Comment...

Science.gov (United States)

2010-07-09

... DEPARTMENT OF VETERANS AFFAIRS [OMB Control No. 2900-0002] Proposed Information Collection (Income-Net Worth and Employment Statement) Activity: Comment Request AGENCY: Veterans Benefits Administration... techniques or the use of other forms of information technology. Title: Income-Net Worth and Employment...

77 FR 20888 - Proposed Information Collection (Income, Net Worth, and Employment Statement) Activity: Comment...

Science.gov (United States)

2012-04-06

... DEPARTMENT OF VETERANS AFFAIRS [OMB Control No. 2900-0002] Proposed Information Collection (Income, Net Worth, and Employment Statement) Activity: Comment Request AGENCY: Veterans Benefits... techniques or the use of other forms of information technology. Title: Income, Net Worth, and Employment...

Improved measurement of aluminum in irradiated fuel reprocessed at the Savannah River Site

International Nuclear Information System (INIS)

Maxwell, S.L. III.

1991-01-01

At the Savannah River Site (SRS), irradiated fuel from research reactor operators or their contract fuel service companies is reprocessed in the H-Canyon Separations Facility. Final processing costs are based on analytical measurements of the amount of total metal dissolved. Shipper estimates for uranium and uranium-235 and measured values at SRS have historically agreed very well. There have occasionally been significant differences between shipper estimates for aluminum and the aluminum content determined at SRS. To minimize analytical error that might contribute to poor shipper-receiver agreement for the reprocessing of off-site fuel, a new analytical method to measure aluminum was developed by SRS Analytical Laboratories at the Central Laboratory Facilities. An EDTA (ethylenediaminetetraacetic acid) titration method, subject to dissolver matrix interferences, was previously used at SRS to measure aluminum in H-Canyon dissolver during the reprocessing of offsite fuel. The new method combines rapid ion exchange technology with direct current argon plasma spectrometry to enhance the reliability of aluminum measurements for off-site fuel. The technique rapidly removes spectral interferences such as uranium and significantly lowers gamma levels due to fission products. Aluminium is separated quantitatively by using an anion exchange technique that employs oxalate complexing, small particle size resin and rapid flow rates. The new method, which has eliminated matrix interference problems with these analyses and improved the quality of aluminum measurements, has improved the overall agreement between shipper-receiver values for offsite fuel processed SRS

Principles of fuel ion ratio measurements in fusion plasmas by collective Thomson scattering

DEFF Research Database (Denmark)

Stejner Pedersen, Morten; Nielsen, Stefan Kragh; Bindslev, Henrik

2011-01-01

ratio. Measurements of the fuel ion ratio will be important for plasma control and machine protection in future experiments with burning fusion plasmas. Here we examine the theoretical basis for fuel ion ratio measurements by CTS. We show that the sensitivity to plasma composition is enhanced......For certain scattering geometries collective Thomson scattering (CTS) measurements are sensitive to the composition of magnetically confined fusion plasmas. CTS therefore holds the potential to become a new diagnostic for measurements of the fuel ion ratio—i.e. the tritium to deuterium density...... by the signatures of ion cyclotron motion and ion Bernstein waves which appear for scattering geometries with resolved wave vectors near perpendicular to the magnetic field. We investigate the origin and properties of these features in CTS spectra and give estimates of their relative importance for fuel ion ratio...

Fuel-Coolant Interactions: Visualization and Mixing Measurements

International Nuclear Information System (INIS)

Loewen, Eric P.; Bonazza, Riccardo; Corradini, Michael L.; Johannesen, Robert E.

2002-01-01

Dynamic X-ray imaging of fuel-coolant interactions (FCI), including quantitative measurement of fuel-coolant volume fractions and length scales, has been accomplished with a novel imaging system at the Nuclear Safety Research Center at the University of Wisconsin, Madison. The imaging system consists of visible-light high-speed digital video, low-energy X-ray digital imaging, and high-energy X-ray digital imaging subsystems. The data provide information concerning the melt jet velocity, melt jet configuration, melt volume fractions, void fractions, and spatial and temporal quantification of premixing length scales for a model fuel-coolant system of molten lead poured into a water pool (fuel temperatures 500 to 1000 K; jet diameters 10 to 30 mm; coolant temperatures 20 to 90 deg. C). Overall results indicate that the FCI has three general regions of behavior, with the high fuel-coolant temperature region similar to what might be expected under severe accident conditions. It was observed that the melt jet leading edge has the highest void fraction and readily fragments into discrete masses, which then subsequently subdivide into smaller masses of length scales <10 mm. The intact jet penetrates <3 to 5 jet length/jet diameter before this breakup occurs into discrete masses, which continue to subdivide. Hydrodynamic instabilities can be visually identified at the leading edge and along the jet column with an interfacial region that consists of melt, vapor, and water. This interface region was observed to grow in size as the water pool temperature was increased, indicating mixing enhancement by boiling processes

Model for the determination of the nuclear fuel

International Nuclear Information System (INIS)

Azevedo, J.B.L. de.

1979-09-01

The Nuclear Fuel Cost Determination Model, MDCN, is a computer program written in FORTRAN IV, meant to calculate the nuclear fuel cost employed in nuclear power plants for heat or electrical energy generation. The economic principles employed are: capital recovery proportional to the energy generation, present worth method for the equivalence of costs and levelized fuel cost calculation. This model presents some inovations in comparasion with other models already in use, since it takes into account refueling and maintenance outages and it does not fix the fuel cycle steps (industrial processes and services). The first inovation leads to a more realistic cost determination and permits the model to be employed together with hydrothermal power system simulators; the second permits a more flexible use of the model, like economical comparison of fuel cycles. Complementing the main body of the work, where the theoretical fundamentals and methodology necessary to the calculation developments are discussed, annexes are included treating in greater detail some specific itens; the more important ones refer to the FORTRAN program, input data preparation and example. (Author) [pt

Interpersonal rejection results in increased appearance satisfaction for women who rely on body weight for self-worth.

Science.gov (United States)

O'Driscoll, Lauren M; Jarry, Josée L

2015-01-01

This study examined whether body weight contingent self-worth (BWCSW), the tendency to base self-worth on body weight, moderates the effects of interpersonal rejection on self-esteem and body satisfaction. In an online survey, female undergraduates (N=148) completed measures of trait self-esteem, depression, and BWCSW. In a subsequent lab session, participants were assigned to either an interpersonal rejection or to a neutral control condition, after which they completed measures of state self-esteem and body satisfaction. Compared to women with lower BWCSW, women with higher BWCSW reported lower appearance self-esteem (p=.001) and body satisfaction (p=.004) across conditions. However, they reacted to rejection by reporting greater appearance self-esteem (p=.034) and body satisfaction (p=.021). Rejection had no effect on women with lower BWCSW. The reaction of women with higher BWCSW is interpreted as a compensatory self-enhancement response to interpersonal rejection within a self-important domain. Copyright © 2014 Elsevier Ltd. All rights reserved.

Comparison of Global Sizing Velocimetry and Phase Doppler Anemometry measurements of alternative jet fuel sprays

Science.gov (United States)

Sadr, Reza; Kannaiyan, Kumaran

2013-11-01

Atomization plays a crucial precursor role in liquid fuel combustion that directly affects the evaporation, mixing, and emission levels. Laser diagnostic techniques are often used to study the spray characteristics of liquid fuels. The objective of this work is to compare the spray measurements of Gas-to Liquid (GTL) jet fuels obtained using Global Sizing Velocimetry (GSV) and Phase Doppler Anemometry (PDA) techniques at global and local levels, respectively. The chemical and physical properties of GTL fuels are different from conventional jet fuels, owing to the difference in their production methodology. In this work, the experimental facility, the measurement techniques, and spray characteristics of two different GTL fuels are discussed and compared with those of Jet A-1 fuel. Results clearly demonstrate that although the global measurement gives an overall picture of the spray, fine details are obtained only through local measurements and complement in gaining more inferences into the spray characteristics. The results also show a close similarity in spray characteristics between GTL and Jet A-1 fuels. Funded by Qatar Science and Technology Park.

Education without Moral Worth? Kantian Moral Theory and the Obligation to Educate Others

Science.gov (United States)

Martin, Christopher

2011-01-01

This article examines the possibility of a Kantian justification of the intrinsic moral worth of education. The author critiques a recent attempt to secure such justification via Kant's notion of the Kingdom of Ends. He gives four reasons why such an account would deny any intrinsic moral worth to education. He concludes with a tentative…

Techniques for computing reactivity changes caused by fuel axial expansion in LMR's

International Nuclear Information System (INIS)

Khalil, H.

1988-01-01

An evaluation is made of the accuracy of methods used to compute reactivity changes caused by axial fuel relocation in fast reactors. Results are presented to demonstrate the validity of assumptions commonly made such as linearity of reactivity with fuel elongation, additivity of local reactivity contributions, and the adequacy of standard perturbation techniques. Accurate prediction of the reactivity loss caused by axial swelling of metallic fuel is shown to require proper representation of the burnup dependence of the expansion reactivity. Some accuracy limitations in the methods used in transient analyses, which are based on the use of fuel worth tables, are identified, and efficient ways to improve accuracy are described. Implementation of these corrections produced expansion reactivity estimates within 5% of higher-order method for a metal-fueled FFTF core representation. 18 refs., 3 figs., 3 tabs

Development of Vision System for Dimensional Measurement for Irradiated Fuel Assembly

International Nuclear Information System (INIS)

Shin, Jungcheol; Kwon, Yongbock; Park, Jongyoul; Woo, Sangkyun; Kim, Yonghwan; Jang, Youngki; Choi, Joonhyung; Lee, Kyuseog

2006-01-01

In order to develop an advanced nuclear fuel, a series of pool side examination (PSE) is performed to confirm in-pile behavior of the fuel for commercial production. For this purpose, a vision system was developed to measure for mechanical integrity, such as assembly bowing, twist and growth, of the loaded lead test assembly. Using this vision system, three(3) times of PSE were carried out at Uljin Unit 3 and Kori Unit 2 for the advanced fuels, PLUS7 TM and 16ACE7 TM , developed by KNFC. Among the main characteristics of the vision system is very simple structure and measuring principal. This feature enables the equipment installation and inspection time to reduce largely, and leads the PSE can be finished without disturbance on the fuel loading and unloading activities during utility overhaul periods. And another feature is high accuracy and repeatability achieved by this vision system

Cardiovascular disease - risk benefits of clean fuel technology and policy. A statistical analysis

Energy Technology Data Exchange (ETDEWEB)

Gallagher, Paul; Fischer, Amelia [Economics Department, 481 Heady Hall, Iowa State University, Ames Iowa 50011 (United States); Lazarus, William; Bachewe, Fantu [Applied Economics Department, 253 COB, University of Minnesota, St. Paul, MN 55455 (United States); Shapouri, Hosein; Conway, Roger [Office of Energy Policy and New Uses, 400 Independence Avenue, SW (Rm.4059 So. Bldg), United States Department of Agriculture, Washington, DC 20250 (United States)

2010-02-15

The hypothesis of this study is that there is a statistical relationship between the cardiovascular disease mortality rate and the intensity of fuel consumption (measured in gallons/square mile) at a particular location. We estimate cross-sectional regressions of the mortality rate due to cardiovascular disease against the intensity of fuel consumption using local data for the entire US, before the US Clean Air Act (CAA) in 1974 and after the most recent policy revisions in 2004. The cardiovascular disease rate improvement estimate suggests that up to 60 cardiovascular disease deaths per 100,000 residents are avoided in the largest urban areas with highest fuel consumption per square mile. In New York City, for instance, the mortality reduction may be worth about $30.3 billion annually. Across the US, the estimated Value of Statistical Life (VSL) benefit is $202.7 billion annually. There are likely three inseparable reasons that contributed importantly to this welfare improvement. First, the CAA regulations banned leaded gasoline, and mandated reduction in specific chemicals and smog components. Second, technologies such as the Catalytic Converter (CC) for the automobile and the low particulate diesel engine were adopted. Third, biofuels have had important roles, making the adoption of clean air technology possible and substituting for high emission fuels. (author)

Cardiovascular disease-risk benefits of clean fuel technology and policy: A statistical analysis

Energy Technology Data Exchange (ETDEWEB)

Gallagher, Paul, E-mail: paulg@iastate.ed [Economics Department, 481 Heady Hall, Iowa State University, Ames Iowa 50011 (United States); Lazarus, William [Applied Economics Department, 253 COB, University of Minnesota, St. Paul, MN 55455 (United States); Shapouri, Hosein; Conway, Roger [Office of Energy Policy and New Uses, 400 Independence Avenue, SW (Rm.4059 So. Bldg), United States Department of Agriculture, Washington, DC 20250 (United States); Bachewe, Fantu [Applied Economics Department, 253 COB, University of Minnesota, St. Paul, MN 55455 (United States); Fischer, Amelia [Economics Department, 481 Heady Hall, Iowa State University, Ames Iowa 50011 (United States)

2010-02-15

The hypothesis of this study is that there is a statistical relationship between the cardiovascular disease mortality rate and the intensity of fuel consumption (measured in gallons/square mile) at a particular location. We estimate cross-sectional regressions of the mortality rate due to cardiovascular disease against the intensity of fuel consumption using local data for the entire US, before the US Clean Air Act (CAA) in 1974 and after the most recent policy revisions in 2004. The cardiovascular disease rate improvement estimate suggests that up to 60 cardiovascular disease deaths per 100,000 residents are avoided in the largest urban areas with highest fuel consumption per square mile. In New York City, for instance, the mortality reduction may be worth about $30.3 billion annually. Across the US, the estimated Value of Statistical Life (VSL) benefit is $202.7 billion annually. There are likely three inseparable reasons that contributed importantly to this welfare improvement. First, the CAA regulations banned leaded gasoline, and mandated reduction in specific chemicals and smog components. Second, technologies such as the Catalytic Converter (CC) for the automobile and the low particulate diesel engine were adopted. Third, biofuels have had important roles, making the adoption of clean air technology possible and substituting for high emission fuels.

Cardiovascular disease-risk benefits of clean fuel technology and policy: A statistical analysis

International Nuclear Information System (INIS)

Gallagher, Paul; Lazarus, William; Shapouri, Hosein; Conway, Roger; Bachewe, Fantu; Fischer, Amelia

2010-01-01

The hypothesis of this study is that there is a statistical relationship between the cardiovascular disease mortality rate and the intensity of fuel consumption (measured in gallons/square mile) at a particular location. We estimate cross-sectional regressions of the mortality rate due to cardiovascular disease against the intensity of fuel consumption using local data for the entire US, before the US Clean Air Act (CAA) in 1974 and after the most recent policy revisions in 2004. The cardiovascular disease rate improvement estimate suggests that up to 60 cardiovascular disease deaths per 100,000 residents are avoided in the largest urban areas with highest fuel consumption per square mile. In New York City, for instance, the mortality reduction may be worth about $30.3 billion annually. Across the US, the estimated Value of Statistical Life (VSL) benefit is $202.7 billion annually. There are likely three inseparable reasons that contributed importantly to this welfare improvement. First, the CAA regulations banned leaded gasoline, and mandated reduction in specific chemicals and smog components. Second, technologies such as the Catalytic Converter (CC) for the automobile and the low particulate diesel engine were adopted. Third, biofuels have had important roles, making the adoption of clean air technology possible and substituting for high emission fuels.

Reactivity and neutron emission measurements of highly burnt PWR fuel rod samples

International Nuclear Information System (INIS)

Murphy, M.F.; Jatuff, F.; Grimm, P.; Seiler, R.; Brogli, R.; Meier, G.; Berger, H.-D.; Chawla, R.

2006-01-01

Fuel rods with burnup values beyond 50 GWd/t are characterised by relatively large amounts of fission products and a high abundance of major and minor actinides. Of particular interest is the change in the reactivity of the fuel as a function of burnup and the capability of modern codes to predict this change. In addition, the neutron emission from burnt fuel has important implications for the design of transport and storage facilities. Measurements have been made of the reactivity effects and the neutron emission rates of highly burnt uranium oxide and mixed oxide fuel rod samples coming from a pressurised water reactor (PWR). The reactivity measurements have been made in a PWR lattice in the PROTEUS zero-energy reactor moderated in turn with: water, a water and heavy water mixture and water containing boron. A combined transport flask and sample changer was used to insert the 400 mm long burnt fuel rod segments into the reactor. Both control rod compensation and reactor period methods were used to determine the reactivities of the samples. For the range of burnup values investigated, an interesting exponential relationship has been found between the neutron emission rate and the measured reactivity

Growth rates of breeder reactor fuel. Final report

International Nuclear Information System (INIS)

Ott, K.O.

1979-01-01

During the contract period, a consistent formalism for the definition of the growth rates (and thus the doubling time) of breeder reactor fuel has been developed. This formalism was then extended to symbiotic operation of breeder and converter reactors. Further, an estimation prescription for the growth rate has been developed which is based upon the breeding worth factors. The characteristics of this definition have been investigated, which led to an additional integral concept, the breeding bonus

Data on Occurrence of Selected Trace Metals, Organochlorines, and Semivolatile Organic Compounds in Edible Fish Tissues From Lake Worth, Fort Worth, Texas 1999

National Research Council Canada - National Science Library

Moring, J. B

2002-01-01

.... Air Force and in collaboration with the Texas Department of Health, collected samples of edible fish tissues from Lake Worth for analysis of selected trace metals, organochlorines, and semivolatile...

An investigation of the associations between contingent self-worth and aspirations among Iranian university students.

Science.gov (United States)

Sabzehara, Milad; Ferguson, Yuna Lee; Sarafraz, Mehdi Reza; Mohammadi, Mostafa

2014-01-01

This study investigated the novel associations between intrinsic and extrinsic aspirations and internal and external domains of contingent self-worth among a sample of 502 Iranian university students. We found a meaningful pattern showing that intrinsic aspirations were positively associated with internal domains, whereas extrinsic aspirations were positively associated with external domains. Our survey data also suggested that the factor structure of the Aspiration Index, as well as the factor structure of the Contingencies of Self-Worth Scale in our Iranian sample were consistent with factor structures of foreign samples. Finally, the types of aspirations and domains of contingencies of self-worth meaningfully predicted variables related to well-being, confirming previous research. We discuss the nature of the associations between the aspirations and the domains of contingent self-worth.

Hydrocarbon Fuel Thermal Performance Modeling based on Systematic Measurement and Comprehensive Chromatographic Analysis

Science.gov (United States)

2016-07-31

distribution unlimited Hydrocarbon Fuel Thermal Performance Modeling based on Systematic Measurement and Comprehensive Chromatographic Analysis Matthew...vital importance for hydrocarbon -fueled propulsion systems: fuel thermal performance as indicated by physical and chemical effects of cooling passage... analysis . The selection and acquisition of a set of chemically diverse fuels is pivotal for a successful outcome since test method validation and

Pragmatic sociology and competing orders of worth in organizations

DEFF Research Database (Denmark)

Jagd, Søren

2011-01-01

primarily has been related to three main themes in organizational research: non-profit and co-operative organizations, inter-organizational co-operation, and organizational change. Third, I discuss how the pragmatic, process-oriented aspect of the research program, focusing on the intertwining of values......Different notions of multiple rationalities have recently been applied to describe the phenomena of co-existence of competing rationalities in organizations. These include institutional pluralism, institutional logics, competing rationalities and pluralistic contexts. The French pragmatic...... studies of organizations. First, I summarize the basic ideas of the framework, stressing the aspects of special relevance for studies of organizations. Second, I review the empirical studies focusing on the coexistence of competing orders of worth in organizations showing that the order of worth framework...

PIV measurement of internal structure of diesel fuel spray

Energy Technology Data Exchange (ETDEWEB)

Cao, Z M [Ecotechnology System Lab., Yokohama National Univ. (Japan); Nishino, K [Div. of Artificial Environment and Systems, Yokohama National Univ. (Japan); Mizuno, S [Yokohama National Univ. (Japan); Torii, K [Dept. of Mechanical Engineering and Materials Science, Yokohama National Univ. (Japan)

2000-12-01

This paper reports particle image velocimetry (PIV) measurements of diesel fuel spray injected from a single-hole nozzle at injection pressures ranging from 30 to 70 MPa, which are comparable to partial-load operating conditions of commercial diesel engines. The fuel is injected into a non-combusting environment pressurized up to 2.0 MPa. A laser-induced fluorescent (LIF) technique is utilized to visualize internal structures of fuel sprays formed by densely-distributing droplets. A specially designed synchronization system is developed to acquire double-frame spray images at an arbitrary time delay after injection. A direct cross-correlation PIV technique is applied to measure instantaneous droplet velocity distribution. Unique large-scale structures in droplet concentration, called 'branch-like structures' by Azetsu et al. (1990), are observed and shown to be associated with active vortical motions, which appear to be responsible for the mixing between droplets and the surrounding gas. It is found that the droplets tend to move out of the vortical structures and accumulate in the regions of low vorticity. Some other interesting features concerning droplet velocity fields are also presented. (orig.)

Development of a standard database for FBR core nuclear design (XI). Analysis of the Experimental Fast Reactor 'JOYO' MK-I start-up test and operation data

International Nuclear Information System (INIS)

Yokoyama, Kenji; Numata, Kazuyuki

2000-03-01

As a recent research, Japan Nuclear Cycle Development Institute (JNC) develops a database of integral data in addition to the JUPITER experiments, aiming at further improvement for accuracy and reliability. In this report, the authors describe the evaluation of the C/E values and the sensitivity analysis for the Experimental Fast Reactor 'JOYO' MK-I core. The minimal criticality, sodium void reactivity worth, fuel assembly worth and burn-up coefficient were analyzed. The results of both the minimal criticality and the fuel assembly worth, which were calculated by the standard analytical method for JUPITER experiments, agreed well with the measured values. On the other hand, the results of the sodium void reactivity worth have a tendency to overestimate. As for the burn-up coefficient, it was seen that the C/E values had a dispersion among the operation cycles. The authors judged that further investigation for the estimation of the experimental error will increase the applicability of the integral data to the adjusted library. Furthermore, sensitivity analyses for the minimal criticality, sodium void reactivity worth and fuel assembly worth showed the characteristics of 'JOYO' MK-I core in comparison with ZPPR-9 core of JUPITER experiments. (J.P.N.)

Analysis of systematic error deviation of water temperature measurement at the fuel channel outlet of the reactor Maria

International Nuclear Information System (INIS)

Bykowski, W.

2000-01-01

The reactor Maria has two primary cooling circuits; fuel channels cooling circuit and reactor pool cooling circuit. Fuel elements are placed inside the fuel channels which are parallely linked in parallel, between the collectors. In the course of reactor operation the following measurements are performed: continuous measurement of water temperature at the fuel channels inlet, continuous measurement of water temperature at the outlet of each fuel channel and continuous measurement of water flow rate through each fuel channel. Based on those thermal-hydraulic parameters the instantaneous thermal power generated in each fuel channel is determined and by use of that value the thermal balance and the degree of fuel burnup is assessed. The work contains an analysis concerning estimate of the systematic error of temperature measurement at outlet of each fuel channel and so the erroneous assessment of thermal power extracted in each fuel channel and the burnup degree for the individual fuel element. The results of measurements of separate factors of deviations for the fuel channels are enclosed. (author)

Recycling of MOX fuel for LWRs

International Nuclear Information System (INIS)

Joo, Hyung Kook; Oh, Soo Youl

1992-01-01

The status and issues related to the thermal recycling of reprocessed nuclear fuels have been reviewed. It is focused on the use of reprecessed plutonium in the form of mixed oxide (MOX) for a light water reactor and the review on reprocessing and fabrication processes is beyond the scope. In spite of the difference in the nuclear characteristics between plutonium and uranium isotopes, the neutronics behavior in a core with MOX fuels is similar to that with normal uranium fuels. However, since the neutron spectrum is hardened in a core with MOX, the Doppler, viod, and moderator temperature coefficients become more negative and the control rod and boron worths are slightly reduced. Therefore, the safety will be evaluated carefully in addition to the core neutronics analysis. The MOX fuel rod behavior related to the rod performance such as the pellet to clad interaction and fission gas release is also similar to that of uranium rods, and no specific problem arises. Substituting MOX fuels for a portion of uranium fuels, it is estimated that the savings be about 25% in uranium ore and 10% in uranium enrichment service requirements. The use of MOX fuel in LWRs has been commercialized in European countries including Germany, France, Belgium, etc., and a demonstration program has been pursued in Japan for the commercial utilization in the late 1990s. Such a worldwide trend indicates that the utilization of MOX fuel in LWRs is a proven technology and meets economics criteria. (Author)

Measuring element for determining the internal pressure in fuel rods

International Nuclear Information System (INIS)

Deckers, H.; Drexler, H.; Reiser, H.

1983-01-01

A pressure cell is situated inside the fuel rod, which contains a magnetic core or a core influenced by magnetism, whose position relative to an outer front surface of an end stopper of the fuel rod can vary. The fuel rod contains a pressure cell directly above the lower end stopper or connected to it. This can consist of closed bellows, where if the internal pressure in the fuel rod rises, a ferrite core moves axially. When the pressure drops, this returns to the initial position, which is precisely defined by a stop. To detect a rod defect, the position of the soft iron core relative to the lower edge of the end stopper is scanned by a special measuring device. (orig./HP) [de

Fuel fragmentation data review and separate effects testing

International Nuclear Information System (INIS)

Yueh, Ken. H.; Snis, N.; Mitchell, D.; Munoz-Reja, C.

2014-01-01

A simple alternative test has been developed to study the fuel fragmentation process at loss of coolant accident (LOCA) temperatures. The new test heats a short section of fuel, approximately two pellets worth of material, in a tube furnace open to air. An axial slit is cut in the test sample cladding to reduce radial restraint and to simulate ballooned condition. The tube furnace allows the fuel fragmentation process be observed during the experiment. The test was developed as a simple alternative so large number of tests could be conducted quickly and efficiently to identify key variables that influence fuel fragmentation and to zeroing on the fuel fragmentation burn-up threshold. Several tests were conducted, using fuel materials from fuel rods that were used in earlier integral tests to benchmark and validate the test technique. High burn-up fuel materials known to be above the fragmentation threshold was used to evaluate the fragmentation process as a function of temperature. Even with an axial slit and both ends open, no significant fuel detachment/release was detected until above 750°C. Additional tests were conducted with fuel materials at burn-ups closer to the fuel fragmentation burn-up threshold. Results from these tests indicate a minor power history effect on the fuel fragmentation burn-up threshold. An evaluation of available literature and data generated from this work suggest a fuel fragmentation burn-up threshold between 70 and 75 GWd/MTU. (author)

Does self-threat promote social connection? The role of self-esteem and contingencies of self-worth.

Science.gov (United States)

Park, Lora E; Maner, Jon K

2009-01-01

Six studies examined the social motivations of people with high self-esteem (HSE) and low self-esteem (LSE) following a threat to a domain of contingent self-worth. Whether people desired social contact following self-threat depended on an interaction between an individual's trait self-esteem and contingencies of self-worth. HSE participants who strongly based self-worth on appearance sought to connect with close others following a threat to their physical attractiveness. LSE participants who staked self-worth on appearance wanted to avoid social contact and, instead, preferred a less interpersonally risky way of coping with self-threat (wanting to enhance their physical attractiveness). Implications for theories of self-esteem, motivation, and interpersonal processes are discussed.

Technical measurement of small fission gas inventory in fuel rod with laser puncturing system

International Nuclear Information System (INIS)

Kim, Hee Moon; Kim, Sung Ryul; Lee, Byoung Oon; Yang, Yong Sik; Baek, Sang Ryul; Song, Ung Sup

2012-01-01

The fission gas release cause degradation of fuel rod. It influences fuel temperature and internal pressure due to low thermal conductivity. Therefore, fission gas released to internal void of fuel rod must be measured with burnup. To measure amount of fission gas, fuel rod must be punctured by a steel needle in a closed chamber. Ideal gas law(PV=nRT) is applied to obtain atomic concentration(mole). Steel needle type is good for large amount of fission gas such as commercial spent fuel rod. But, some cases with small fuel rig in research reactor for R/D program are not available to use needle type because of large chamber volume. The laser puncturing technique was developed to solve measurement of small amount of fission gas. This system was very rare equipment in other countries. Fine pressure gage and strong vacuum system were installed, and the chamber volume was reduced at least. Fiber laser was used for easy operation

The effects of functional group counseling on inspiring low-achieving students' self-worth and self-efficacy in Taiwan.

Science.gov (United States)

Hong, Zuway-R; Lin, Huann-shyang; Wang, Hsin-Hui; Chen, Hsiang-Ting; Yu, Tien-chi

2012-01-01

In this study, we investigated the effects of functional group counseling on inspiring low achievers' self-worth and self-efficacy in Taiwan. Forty-three 10th grade low-achieving students volunteered as the Experimental Group to join a 24-week intervention, which integrated and utilized functional group counseling; another 51 10th grade low-achieving students volunteered to be Comparison Group I. In addition, 43 10th grade moderate or high academic achieving students volunteered to be Comparison Group II. All participants completed the Vocational School Student Questionnaire at the beginning and end of this study to measure their self-worth and self-efficacy. In addition, six target students (two boys and four girls) with the lowest total scores on self-worth or self-efficacy in the pretest were selected from the Experimental Group to be interviewed at the end of the intervention and observed weekly. Analyses of variance, analyses of covariance, and paired t-tests assessed the similarity and differences among groups. The initial findings were as follows: Experimental group students had significantly higher scores on self-efficacy and self-worth than both Comparison Group I and Group II students and functional group counseling was shown to significantly affect the low-achieving students. Qualitative results from interviews and observations were used for triangulation and consolidation of quantitative results. Implications of the study included the recommended use of functional group counseling with low-achieving students.

Measurement of enriched uranium and uranium-aluminum fuel materials with the AWCC

International Nuclear Information System (INIS)

Krick, M.S.; Menlove, H.O.; Zick, J.; Ikonomou, P.

1985-05-01

The active well coincidence counter (AWCC) was calibrated at the Chalk River Nuclear Laboratories (CRNL) for the assay of 93%-enriched fuel materials in three categories: (1) uranium-aluminum billets, (2) uranium-aluminum fuel elements, and (3) uranium metal pieces. The AWCC was a standard instrument supplied to the International Atomic Energy Agency under the International Safeguards Project Office Task A.51. Excellent agreement was obtained between the CRNL measurements and previous Los Alamos National Laboratory measurements on similar mockup fuel material. Calibration curves were obtained for each sample category. 2 refs., 8 figs., 15 tabs

Dry, portable calorimeter for nondestructive measurement of the activity of nuclear fuel

Science.gov (United States)

Beyer, Norman S.; Lewis, Robert N.; Perry, Ronald B.

1976-01-01

The activity of a quantity of heat-producing nuclear fuel is measured rapidly, accurately and nondestructively by a portable dry calorimeter comprising a preheater, an array of temperature-controlled structures comprising a thermally guarded temperature-controlled oven, and a calculation and control unit. The difference between the amounts of electric power required to maintain the oven temperature with and without nuclear fuel in the oven is measured to determine the power produced by radioactive disintegration and hence the activity of the fuel. A portion of the electronic control system is designed to terminate a continuing sequence of measurements when the standard deviation of the variations of the amount of electric power required to maintain oven temperature is within a predetermined value.

A Short Is Worth a Thousand Films!

Science.gov (United States)

Massi, Maria Palmira; Blázquez, Bettiana Andrea

2012-01-01

The importance of visual input in the contemporary ELT classroom is such that it is commonplace to use audiovisual elements provided by pictures, films, clips and the like. The power of images is unquestionable, and as the old saying goes, an image is worth a thousand words. Following this line of reasoning, the objective of this article is to…

Determination of the control rod worth for research reactors

International Nuclear Information System (INIS)

Aldama, D.L.; Gual, M.R.

2000-01-01

Nowadays there is a big interest in developing neutronic analysis methods for research reactor and particularly for the determination of the control rods worth under different operation conditions and core configurations. The reactivity associated with the control rods is of interest in the shutdown margin and in calculations of possible abnormal conditions related to reactivity accidents. For theses studies several computer codes have been developed. The present work is aimed at the validation of the calculation methods of the Nuclear Technology Center of Cuba. For this purpose, in order to evaluate the safety of this type of installations, the reactivity worth of the control rods of the cylindrical configuration of the Brazilian critical assembly IPEN/MB-01 is determined. These calculations, however, are a relatively complex task that requires the use of three-dimensional models. Because of this, the validation of the calculation methods used for this purpose is of great importance. In fact, it is one of the requirements called upon by the quality assurance programs for the development, maintenance and utilization of the calculation codes used in safety analysis. For the calculation of control rod worth the lattice code WIMS-D/4 [8] and the diffusion code SNAP-3D [9] were used. This work presents the obtained results and gives a comparison with the experimental values

Zircaloy sheathed thermocouples for PWR fuel rod temperature measurements

International Nuclear Information System (INIS)

Anderson, J.V.; Wesley, R.D.; Wilkins, S.C.

1979-01-01

Small diameter zircaloy sheathed thermocouples have been developed by EG and G Idaho, Inc., at the Idaho National Engineering Laboratory. Surface mounted thermocouples were developed to measure the temperature of zircaloy clad fuel rods used in the Thermal Fuels Behavior Program (TFBP), and embedded thermocouples were developed for use by the Loss-of-Fluid Test (LOFT) Program for support tests using zircaloy clad electrically heated nuclear fuel rod simulators. The first objective of this developmental effort was to produce zircaloy sheathed thermocouples to replace titanium sheathed thermocouples and thereby eliminate the long-term corrosion of the titanium-to-zircaloy attachment weld. The second objective was to reduce the sheath diameter to obtain faster thermal response and minimize cladding temperature disturbance due to thermocouple attachment

Recent improvements to TRIGLAV code

International Nuclear Information System (INIS)

Zagar, T.; Ravnik, M.; Persic, A.

1998-01-01

TRIGLAV code was developed for TRIGA research reactor calculations and is based on two-dimensional diffusion equation. The main purpose of the program is calculation of the fuel elements burn-up. Calculated core burn-up and excess reactivity results are compared with experimental values. New control rod model is introduced and tested in this paper. Calculated integral control rod worth and calculated integral reactivity curves are presented and compared with measured values. Comparison with measured fuel element worth values is presented as a test for two-dimensional flux distribution calculations.(author)

Fission rate measurements in fuel plate type assembly reactor cores

International Nuclear Information System (INIS)

Rogers, J.W.

1988-01-01

The methods, materials and equipment have been developed to allow extensive and precise measurement of fission rate distributions in water moderated, U-Al fuel plate assembly type reactor cores. Fission rate monitors are accurately positioned in the reactor core, the reactor is operated at a low power for a short time, the fission rate monitors are counted with detectors incorporating automated sample changers and the measurements are converted to fission rate distributions. These measured fission rate distributions have been successfully used as baseline information related to the operation of test and experimental reactors with respect to fission power and distribution, fuel loading and fission experiments for approximately twenty years at the Idaho National Engineering Laboratory (INEL). 7 refs., 8 figs

Accuracy of dimension measurements from neutron radiographs of nuclear fuel pins

International Nuclear Information System (INIS)

Domanus, J. C.

1976-03-01

A review of different methods used for dimension measurements from neutron radiographs. The results are presented of an investigation performed using unirradiated fuel pins with calibrated UO 2 pellet-diameters and fuel-to-clad gaps. A projection microscope, three types of travelling microdensitometers and an electronic image analyzer were used to measure diameters and gaps from neutron radiographs produced at Risoe and Studsvik (Sweden) using different brands of X-ray films and transfer technique with 0.1 mm Dy foil. (author)

Material control in nuclear fuel fabrication facilities. Part II. Accountability, instrumentation and measurement techniques in fuel fabrication facilities

International Nuclear Information System (INIS)

Borgonovi, G.M.; McCartin, T.J.; McDaniel, T.; Miller, C.L.; Nguyen, T.

1978-01-01

This report describes the measurement techniques, the instrumentation, and the procedures used in accountability and control of nuclear materials, as they apply to fuel fabrication facilities. A general discussion is given of instrumentation and measurement techniques which are presently used being considered for fuel fabrication facilities. Those aspects which are most significant from the point of view of satisfying regulatory constraints have been emphasized. Sensors and measurement devices have been discussed, together with their interfacing into a computerized system designed to permit real-time data collection and analysis. Estimates of accuracy and precision of measurement techniques have been given, and, where applicable, estimates of associated costs have been presented. A general description of material control and accounting is also included. In this section, the general principles of nuclear material accounting have been reviewed first (closure of material balance). After a discussion of the most current techniques used to calculate the limit of error on inventory difference, a number of advanced statistical techniques are reviewed. The rest of the section deals with some regulatory aspects of data collection and analysis, for accountability purposes, and with the overall effectiveness of accountability in detecting diversion attempts in fuel fabrication facilities. A specific example of application of the accountability methods to a model fuel fabrication facility is given. The effect of random and systematic errors on the total material uncertainty has been discussed, together with the effect on uncertainty of the length of the accounting period

Material control in nuclear fuel fabrication facilities. Part II. Accountability, instrumentation and measurement techniques in fuel fabrication facilities

Energy Technology Data Exchange (ETDEWEB)

Borgonovi, G.M.; McCartin, T.J.; McDaniel, T.; Miller, C.L.; Nguyen, T.

1978-01-01

This report describes the measurement techniques, the instrumentation, and the procedures used in accountability and control of nuclear materials, as they apply to fuel fabrication facilities. A general discussion is given of instrumentation and measurement techniques which are presently used being considered for fuel fabrication facilities. Those aspects which are most significant from the point of view of satisfying regulatory constraints have been emphasized. Sensors and measurement devices have been discussed, together with their interfacing into a computerized system designed to permit real-time data collection and analysis. Estimates of accuracy and precision of measurement techniques have been given, and, where applicable, estimates of associated costs have been presented. A general description of material control and accounting is also included. In this section, the general principles of nuclear material accounting have been reviewed first (closure of material balance). After a discussion of the most current techniques used to calculate the limit of error on inventory difference, a number of advanced statistical techniques are reviewed. The rest of the section deals with some regulatory aspects of data collection and analysis, for accountability purposes, and with the overall effectiveness of accountability in detecting diversion attempts in fuel fabrication facilities. A specific example of application of the accountability methods to a model fuel fabrication facility is given. The effect of random and systematic errors on the total material uncertainty has been discussed, together with the effect on uncertainty of the length of the accounting period.

Externally Contingent and Unstable Self-Worth as Predictors of Depression in College Women: A 3-Month Study

Science.gov (United States)

Lopez, Frederick G.; Thorne, Brittany; Schoenecker, Sydney; Siffert, Kevin; Chaliman, Rebecca; Castleberry, Erica

2014-01-01

Undergraduate women completed measures of externally contingent self-worth, self-esteem instability, and depression at the beginning (Time 1) and again 3 months later near the end (Time 2) of an academic semester. Findings indicated that when Time 1 depression scores were controlled for, each Time 1 facet of self-esteem made significant…

Self-perceptions, self-worth and sport participation in adolescents.

Science.gov (United States)

Balaguer, Isabel; Atienza, Francisco L; Duda, Joan L

2012-07-01

The purpose of this study was to study the associations between specific self-perceptions and global self-worth with different frequency levels of sport participation among Spanish boys and girls adolescents. Students (457 boys and 460 girls) completed the Self Perception Profile for Children (Harter, 1985) and items assessing sport engagement from The Health Behavior in School Children Questionnaire (Wold, 1995). Results showed that some specific dimensions of self-perception were related to different frequency of sport participation whereas overall judgments of self-worth did not. Specifically, for boys and girls, higher levels of sport participation were positively associated to Athletic Competence, and for boys were also associated with Physical Appearance and Social Acceptance. The potential implications of domain specific socialisation processes on the configuration of self-perceptions are highlighted.

The KALIMER-600 Reactor Core Design Concept with Varying Fuel Cladding Thickness

International Nuclear Information System (INIS)

Hong, Ser Gi; Jang, Jin Wook; Kim, Yeong Il

2006-01-01

Recently, Korea Atomic Energy Research Institute (KAERI) has developed a 600MWe sodium cooled fast reactor, the KALIMER-600 reactor core concept using single enrichment fuel. This reactor core concept is characterized by the following design targets : 1) Breakeven breeding (or fissile-self-sufficient) without any blanket, 2) Small burnup reactivity swing ( 23 n/cm 2 ). In the previous design, the single enrichment fuel concept was achieved by using the special fuel assembly designs where non-fuel rods (i.e., ZrH 1.8 , B 4 C, and dummy rods) were used. In particular, the moderator rods (ZrH 1.8 ) were used to reduce the sodium void worth and the fuel Doppler coefficient. But it has been known that this hydride moderator possesses relatively poor irradiation behavior at high temperature. In this paper, a new core design concept for use of single enrichment fuel is described. In this concept, the power flattening is achieved by using the core region wise cladding thicknesses but all non-fuel rods are removed to simplify the fuel assembly design

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

International Nuclear Information System (INIS)

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

1980-06-01

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

Measuring method for effective neutron multiplication factor upon containing irradiated fuel assembly

International Nuclear Information System (INIS)

Ueda, Makoto; Mitsuhashi, Ishi; Sasaki, Tomoharu.

1993-01-01

A portion of irradiated fuel assemblies at a place where a reactivity effect is high, that is, at a place where neutron importance is high is replaced with standard fuel assemblies having a known composition to measure neutron fluxes at each of the places. An effective composition at the periphery of the standard fuel assemblies is determined by utilizing a calibration curve determined separately based on the composition and neutron flux values of the standard assemblies. By using the calibration curve determined separately based on this composition and the known composition of the standard fuel assemblies, an effective neutron multiplication factor for the fuel containing portion containing the irradiated fuel assemblies is recognized. Then, subcriticality is ensured and critical safety upon containing the fuel assemblies can be secured quantitatively. (N.H.)

76 FR 60364 - Net Worth and Equity Ratio

Science.gov (United States)

2011-09-29

... to ``follow the new Financial Accounting Standards Board (FASB) rule while still allowing the capital... accepted accounting principles and as further defined in Sec. 702.2(f) of this chapter. * * * * * [[Page... also proposed technical changes to the term ``net worth'' to ensure consistency and accurate accounting...

Preliminary assessment on compatibility of DUPIC fuel with CANDU-6

International Nuclear Information System (INIS)

Choi, Hang-Bok; Roh, G.H.; Jeong, C.J.; Rhee, B.W.; Choi, J.W.; Boss, C.R.

1997-01-01

The compatibility of DUPIC fuel with the existing CANDU-6 reactor was assessed. The technical issues of DUPIC fuel compatibility were chosen based on the CANDU physics design requirements and inherent characteristics of DUPIC fuel. The compatibility was assessed for the reference DUPIC fuel composition which was determined to reduce the composition heterogeneity and improve the spent PWR fuel utilization. Preliminary studies on a CANDU core loaded with DUPIC fuel have shown that the nominal power distribution is flatter than that of a natural uranium core when a 2-bundle shift refueling scheme is used, which reduces the reactivity worths of devices in the core and, therefore, the performance of reactivity devices was assessed. The safety of the core was assessed by a LOCA simulation and it was found that the power pulse upon LOCA can be maintained below that in the natural uranium core when a poison material is used in the DUPIC fuel. For the feasibility of handling DUPIC fuel in the plant, it will be necessary to introduce new equipment to load the DUPIC fuel in the refueling magazine. The radiation effect of DUPIC fuel on both the reactor hardware and the environment will be qualitatively analyzed later. (author)

The mediating effect of global self-worth on physical activity in African-American adolescent females.

Science.gov (United States)

Powell-Young, Yolanda M

2009-07-01

Having an excess of body fat has been identified as a predictor for participatory frequency in physical activity, a behavior that influences the development and persistence of obesity. However, the psychological factors that contribute to this pathway have not been as easily identified. This is particularly significant for population subgroups that are not only uniquely impacted by obesity-related morbidities but who are underrepresented in research as well. This study sample consisted of African-American adolescent females (N = 310), from 14 to 18 years of age, who were recruited from the urban South. Data obtained from self-reported and demographic questionnaires, as well as from anthropometric measurements, were analyzed to explore the mediating effect of global self-worth between BMI and physical activity. Mediation analysis revealed that 2% of the influence that BMI exerts on how frequently African-American adolescent females engaged in physical activity can be attributed to global self-worth.

Linear variable differential transformer and its uses for in-core fuel rod behavior measurements

International Nuclear Information System (INIS)

Wolf, J.R.

1979-01-01

The linear variable differential transformer (LVDT) is an electromechanical transducer which produces an ac voltage proportional to the displacement of a movable ferromagnetic core. When the core is connected to the cladding of a nuclear fuel rod, it is capable of producing extremely accurate measurements of fuel rod elongation caused by thermal expansion. The LVDT is used in the Thermal Fuels Behavior Program at the U.S. Idaho National Engineering Laboratory (INEL) for measurements of nuclear fuel rod elongation and as an indication of critical heat flux and the occurrence of departure from nucleate boiling. These types of measurements provide important information about the behavior of nuclear fuel rods under normal and abnormal operating conditions. The objective of the paper is to provide a complete account of recent advances made in LVDT design and experimental data from in-core nuclear reactor tests which use the LVDT

Analysis of Atmospheric Mercury and Associated Trace Gases in Dallas Fort Worth, TX (Barnett Shale area)

Science.gov (United States)

Laine, P. L.; Talbot, R. W.; Lefer, B. L.; Flynn, J. H.

2012-12-01

Throughout the month of June 2011, a variety of air quality measurements were obtained in the Dallas Fort Worth (Barnett Shale) field campaign. Species such as Hg0, O3, CO, NO, NO2, SO2 were monitored continuously along with a variety of volatile organic carbon (VOC) species ranging in size from C2 (ethane) to C9 aromatics to sesquiterpines. Mixed layer boundary heights were also monitored by Ceilometer measurements. At first glance, the mercury data has peaks that reach as high as 750 ppqv (parts per quadrillion by volume) which is approximately a 5 fold increase over the typical background values observed (~ 150 ppqv). The Fort Worth area has underlying Barnett Shale with thousands of natural gas compressor stations scattered throughout the surrounding landscape. We believe that a potential source of the elevated Hg0 is the result of leakage from these stations under the nocturnal boundary layer. A closer look at diurnal variations and backward wind trajectories will yield information pertaining to the types of air masses spanning the area. We will utilize the suite of chemical and meteorological measurements conducted during the campaign to facilitate source identification for specific time periods. Analysis of these data should provide new information on as yet unexplored sources of atmospheric mercury.

Notification determining technical details concerning measures for transportation of nuclear fuel materials

International Nuclear Information System (INIS)

1977-01-01

These provisions are established on the basis of and to enforce ''The regulation for installation and operation of reactor'', ''The regulation concerning the fabricating business of nuclear fuel'' and ''The regulations concerning the reprocessing business of spent fuel''. The terms used hereinafter are according to those used in such regulations. The limit of radioactivity concentration of things contaminated by the nuclear fuel materials which are not required to be enclosed in vessels is defined in the lists attached. In the applications for the approval of the measures concerning the transport of things remarkably difficult to be enclosed in vessels, the name and the address of the applicant, the kind, quantity, form and constitution of the thing contaminated by the nuclear fuel materials to be transported, the date and route of the transport and the measures for the prevention of injuries during the transport must be written. The limit of quantity of nuclear fuel materials classifying the performance of vessels is defined respectively in the lists attached. The radiation dose rates provided for by the Director General of the Science and Technology Agency concerning transported things and transporting apparatuses are 200 millirem per hour on the surfaces of such things and containers. The nuclear fission materials specified, for which the measures for the prevention of criticality are especially required, include uranium 233, uranium 235, plutonium 238, plutonium 239, plutonium 241, and the chemical compounds of such substances, and the nuclear fuel materials containing one or two and more of such substances, excluding the nuclear fuel materials with less than 15 grams of such uranium and plutonium. (Okada, K.)

Sodium-cooled Fast Reactor Cores using Uranium-Free Metallic Fuels for Maximizing TRU Support Ratio

International Nuclear Information System (INIS)

You, WuSeung; Hong, Ser Gi

2014-01-01

The depleted uranium plays important roles in the SFR burner cores because it substantially contributes to the inherent safety of the core through the negative Doppler coefficient and large delayed neutron. However, the use of depleted uranium as a diluent nuclide leads to a limited value of TRU support ratio due to the generation of TRUs through the breeding. In this paper, we designed sodium cooled fast reactor (SFR) cores having uranium-free fuels 3,4 for maximization of TRU consumption rate. However, the uranium-free fuelled burner cores can be penalized by unacceptably small values of the Doppler coefficient and small delayed neutron fraction. In this work, metallic fuels of TRU-(W or Ni)-Zr are considered to improve the performances of the uranium-free cores. The objective of this work is to consistently compare the neutronic performances of uranium-free sodium cooled fast reactor cores having TRU-Zr metallic fuels added with Ni or W and also to clarify what are the problematic features to be resolved. In this paper, a consistent comparative study of 400MWe sodium cooled burner cores having uranium-based fuels and uranium-free fuels was done to analyze the relative core neutronic features. Also, we proposed a uranium-free metallic fuel based on Nickel. From the results, it is found that tungsten-based uranium-free metallic fuel gives large negative Doppler coefficient due to high resonance of tungsten isotopes but this core has large sodium void worth and small effective delayed neutron fraction while the nickel-based uranium-free metallic fuelled core has less negative Doppler coefficient but smaller sodium void worth and larger effective delayed neutron fraction than the tungsten-based one. On the other hand, the core having TRU-Zr has very high burnup reactivity swing which may be problematic in compensating it using control rods and the least negative Doppler coefficient

Beyond Rational Autonomy: Levinas and the Incomparable Worth of the Student as Singular Other

Science.gov (United States)

Joldersma, Clarence W.

2008-01-01

This article explores the question: Why are students of worth? Educationally, an answer often involves a Kantian response: They are of worth because they are always ends and never means. This response is usually connected to a notion of autonomy interpreted as individual, rational self-determination. The article argues for a different answer. The…

Evaluation of accuracy of Monte Carlo code MVP with VHTRC experiments. Multiplication factor at criticality, burnable poison worth and void worth

International Nuclear Information System (INIS)

Nojiri, Naoki; Yamashita, Kiyonobu; Fiujimoto, Nozomu; Nakano, Masaaki , Yamane, Tsuyoshi; Akino, Fujiyoshi.

1997-11-01

Experimental data of VHTRC (Very High Temperature Reactor Critical Assembly) were analyzed using Monte Carlo code MVP (general purpose Monte Carlo code of neutron and photon transport calculations based on the continuous energy method). The calculation accuracy of the code was evaluated by the analysis for nuclear characteristics of a HTGR (high temperature gas-cooled reactor). The MVP code can analyze with a detailed three-dimensional core model with a few approximations. The HTGRs have following characteristics from view point of nuclear design : they have burnable poisons, many void holes, namely, the control insertion holes and so on. Taking account of these characteristics, multiplication factor at criticality, burnable poison worth, and void worth were evaluated. The maximum calculation errors were 0.8%Δk, 7%, and 25% respectively, From these results, it can be concluded that the MVP code is able to be applied to the nuclear characteristics analysis of the HTGR like the High Temperature Engineering Test Reactor (HTTR). (author)

Studies of Nuclear Fuel Performance Using On-site Gamma-ray Spectroscopy and In-pile Measurements

International Nuclear Information System (INIS)

Matsson, Ingvar

2006-01-01

Presently there is a clear trend of increasing demands on in-pile performance of nuclear fuel. Higher target burnups, part length rods and various fuel additives are some examples of this trend. Together with an increasing demand from the public for even safer nuclear power utilisation, this implies an increased focus on various experimental, preferably non-destructive, methods to characterise the fuel. This thesis focuses on the development and experimental evaluation of such methods. In its first part, the thesis presents a method based on gamma-ray spectroscopy with germanium detectors that have been used at various power reactors in Europe. The aim with these measurements is to provide information about the thermal power distribution within fuel assemblies in order to validate core physics production codes. The early closure of the Barsebaeck 1 BWR offered a unique opportunity to perform such validations before complete depletion of burnable absorbers in Gd-rods had taken place. To facilitate the measurements, a completely submersible measuring system, LOKET, was developed allowing for convenient in-pool measurements to be performed. In its second part, the thesis describes methods that utilise in-pile measurements. These methods have been used in the Halden test-reactor for determination of fission gas release, pellet-cladding interaction studies and fuel development studies. Apart from the power measurements, the LOKET device has been used for fission gas release (FGR) measurements on single fuel rods. The significant reduction in fission gas release in the modern fuel designs, in comparison with older designs, has been demonstrated in a series of experiments. A FGR database covering a wide range of burnup, power histories and fuel designs has been compiled and used for fuel performance analysis. The fission gas release has been measured on fuel rods with average burnups well above 60 MWd/kgU. The comparison between core physics calculations (PHOENIX-4/POLCA

Measurement techniques in dry-powdered processing of spent nuclear fuels

International Nuclear Information System (INIS)

Bowers, D. L.; Hong, J.-S.; Kim, H.-D.; Persiani, P. J.; Wolf, S. F.

1999-01-01

High-performance liquid chromatography (HPLC) with inductively coupled plasma mass spectrometry (ICPMS) detection, α-spectrometry (α-S), and γ-spectrometry (γ-S) were used for the determination of nuclide content in five samples excised from a high-burnup fuel rod taken from a pressurized water reactor (PWR). The samples were prepared for analysis by dissolution of dry-powdered samples. The measurement techniques required no separation of the plutonium, uranium, and fission products. The sample preparation and analysis techniques showed promise for in-line analysis of highly-irradiated spent fuels in a dry-powdered process. The analytical results allowed the determination of fuel burnup based on 148 Nd, Pu, and U content. A goal of this effort is to develop the HPLC-ICPMS method for direct fissile material accountancy in the dry-powdered processing of spent nuclear fuel

No Occasion for Pleasure: The Self-Worth Contingency of a Setback and Coping With Humor

Directory of Open Access Journals (Sweden)

Fay Caroline Mary Geisler

2014-08-01

Full Text Available Whether or not one uses humor to cope with a setback may depend on the idiosyncratic relation of the setback to feeling of self-worth. All people pursue the higher order goal of self-validation, but people differ in what domains of life their self-worth is contingent upon and to what extent. In this article based on an incongruity theory of humor we argue that the use of humor in coping with a highly self-worth-contingent setback may be impeded by two cognitive-motivational processes: goal-driven activation and goal shielding. From the outlined theory we derived the hypothesis that the more a domain is contingent upon self-worth, the less likely a person will be to use humor to deal with a setback in that domain. We tested this hypothesis in two studies employing two forms of self-report, i.e., ratings of reaction likelihood to setbacks described at an abstract domain level (Study 1, and ranking of reaction likelihood to concrete setbacks from different domains (Study 2. The hypothesis was affirmed in different domains of self-worth contingency controlling for the influence of habitual coping with humor, coping by disengagement, and global self-esteem.

Spent fuel measurements. passive neutron albedo reactivity (PNAR) and photon signatures

Energy Technology Data Exchange (ETDEWEB)

Eigenbrodt, Julia [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Menlove, Howard Olsen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-03-29

The International Atomic Energy Agency’s (IAEA) safeguards technical objective is the timely detection of a diversion of a significant quantity of nuclear material from peaceful activities to the manufacture of nuclear weapons or of other nuclear explosive devices or for purposes unknown, and deterrence of such diversion by the risk of early detection. An important IAEA task towards meeting this objective is the ability to accurately and reliably measure spent nuclear fuel (SNF) to verify reactor operating parameters and verify that the fuel has not been removed from reactors or SNF storage facilities. This dissertation analyzes a method to improve the state-of-the-art of nuclear material safeguards measurements using two combined measurement techniques: passive neutron albedo reactivity (PNAR) and passive spectral photon measurements.

Perfectionism and Contingent Self-Worth in Relation to Disordered Eating and Anxiety.

Science.gov (United States)

Bardone-Cone, Anna M; Lin, Stacy L; Butler, Rachel M

2017-05-01

Perfectionism has been proposed as a transdiagnostic risk factor linked to eating disorders and anxiety. In the current study, we examine domains of contingent self-worth as potential moderators of the relationships between maladaptive perfectionism and disordered eating and anxiety using two waves of data collection. Undergraduate females (N = 237) completed online surveys of the study's core constructs at two points separated by about 14 months. At a bivariate level, maladaptive perfectionism was positively associated with disordered eating and anxiety. Maladaptive perfectionism and both appearance and relationship contingent self-worth interacted to predict increases in disordered eating. Neither of the interactive models predicted change in anxiety. Findings highlight maladaptive perfectionism as a transdiagnostic construct related to both disordered eating and anxiety. Interactive findings suggest that targeting maladaptive perfectionism and contingent self-worth (appearance, relationship) in prevention and treatment efforts could mitigate risk for the development or increase of disordered eating. Copyright © 2016. Published by Elsevier Ltd.

Combustion and emission formation in a biomass fueled grate furnace - measurements and modelling

International Nuclear Information System (INIS)

Lindsjoe, H.

1997-06-01

A study of turbulent combustion with special emphasis on the formation of nitrous oxide emissions in a biomass fueled grate furnace has been conducted with the aid of measurements, literature studies and CFD-computations. The literature study covers nitrous oxide formation and the pyrolysis, gasification and combustion of biomass fuel. The measurements were conducted inside the furnace and at the outlet, and temperature and some major species were measured. A tool for the treatment of the bed processes (pyrolysis, gasification and combustion) has been developed. The measurements show significantly higher concentrations of oxygen above the fuel bed than expected. The gas production in the bed was shown to be very unevenly distributed over the width of the furnace. The measured temperatures were relatively low and in the same order as reported from other, similar measurements. The computational results are in good quantitative agreement with the measurements, even for the nitrous oxide emissions. It was necessary to include tar as one of the combustible species to achieve reasonable results. The computations point out that the fuel-NO mechanism is the most important reaction path for the formation of nitrous oxide in biomass combustion in grate furnaces. The thermal NO mechanism is responsible for less than 10% of the total amount of NO-emissions. Although the results are quantitatively in good agreement with the measurements, a sensitivity study showed that the fuel-NO model did not respond to changes in the distribution of secondary air as the measurements indicate. The results from this work have lead to some guidelines on how the furnace should be operated to achieve minimum NO-emissions. Some proposals of smaller changes in the construction are also given. 33 refs, 37 figs, 7 tabs

HCPWR type fuel elements design with respect to its control rods

International Nuclear Information System (INIS)

Abbate, M.J.; Sbaffoni, M.M.; Patino, N.E.; Torasso, O.

1992-01-01

The high conversion reactors (HCPWR) can improve the nuclear fuel utilization. One of its present problems is the optimization of the control rods' worth and its relationship with the void coefficient. This investigation means, starting from one reference's design, to analyze several possibilities on the number and distribution of the control rods. As main result, one design including 24 control rods in an optimized distribution, is recommended. (author)

Derivation of the acceptance and self-worth adjustment scale.

Science.gov (United States)

Tabrett, Daryl R; Latham, Keziah

2010-11-01

The original 55-item Nottingham Adjustment Scale (NAS) is a first generation self-report instrument constructed using classical test theory to evaluate adjustment to vision loss. This study assesses the function of the NAS using Rasch analysis in a sample of adults with visual impairment and presents a revised second-generation instrument. Ninety-nine subjects with established vision loss (median onset 5 years) were administered the NAS. Rasch analysis was performed to: (1) determine optimum response scale function, (2) aid item reduction, (3) determine reliability indices and item targeting, (4) assess unidimensionality using Rasch-based principal component analysis, (5) assess differential item functioning (notable defined as >1.0 logit), and (6) formulate person measures to correlate with Geriatric Depression Scale scores and distance visual acuity to indicate convergent and discriminant validity, respectively. Response categories exhibited underutilization, which when repaired improved response scale functioning and ordered structural calibrations. Misfitting items were removed iteratively until all items had mean-square infit and outfit values of 0.70 to 1.30. However, principal component analysis confirmed insufficient unidimensionality (two contrasts identified, eigenvalues 2.4 and 2.3). Removal of these contrasts and two further iterations restored unidimensionality. Despite item mistargeting (1.58 logits), the revised 19-item instrument demonstrated good person (0.85) and item (0.96) reliability coefficients, good convergent and discriminant validity, and no systematic differential item functioning. The resultant 19-item instrument was termed the Acceptance and Self-Worth Adjustment Scale (AS-WAS). In those with established vision loss, the 19-item Acceptance and Self-Worth Adjustment Scale is a reliable and valid instrument that estimates the level of adjustment concerned with acceptance, attitudes, self-esteem, self-efficacy, and locus of control. An

Comparison of thorium-based fuels with different fissile components in existing BWRs

International Nuclear Information System (INIS)

Bjoerk, Klara Insulander; Fhager, Valentin; Demaziere, Christophe

2009-01-01

Three different types of thorium based BWR fuel have been developed, in each of which thorium was combined with a different fissile component, the three components being reactor grade plutonium, uranium enriched to 20% in uranium 235 and pure uranium 233. A BWR nuclear bundle design, based on the geometrical fuel assembly design GE14, was developed for each of these fissile components. The properties and performance of the corresponding fuel assemblies were investigated via full core calculations carried out for an existing BWR and compared with the ones of an ordinary Low Enriched Uranium (LEU) fuel, which was developed for reference. The fuel assemblies and cores were designed to meet existing fuel design criteria, and were then analyzed with regards to reactivity coefficients, delayed neutron fractions, control rod worths and shutdown margins. The results show that all three alternatives seem to be feasible, although some difficulties remain with complying with the thermal limits, and with the moderator temperature and coolant void coefficients of the U-233 containing fuel being positive under some circumstances. (author)

Measurement of Weight of Kernels in a Simulated Cylindrical Fuel Compact for HTGR

International Nuclear Information System (INIS)

Kim, Woong Ki; Lee, Young Woo; Kim, Young Min; Kim, Yeon Ku; Eom, Sung Ho; Jeong, Kyung Chai; Cho, Moon Sung; Cho, Hyo Jin; Kim, Joo Hee

2011-01-01

The TRISO-coated fuel particle for the high temperature gas-cooled reactor (HTGR) is composed of a nuclear fuel kernel and outer coating layers. The coated particles are mixed with graphite matrix to make HTGR fuel element. The weight of fuel kernels in an element is generally measured by the chemical analysis or a gamma-ray spectrometer. Although it is accurate to measure the weight of kernels by the chemical analysis, the samples used in the analysis cannot be put again in the fabrication process. Furthermore, radioactive wastes are generated during the inspection procedure. The gamma-ray spectrometer requires an elaborate reference sample to reduce measurement errors induced from the different geometric shape of test sample from that of reference sample. X-ray computed tomography (CT) is an alternative to measure the weight of kernels in a compact nondestructively. In this study, X-ray CT is applied to measure the weight of kernels in a cylindrical compact containing simulated TRISO-coated particles with ZrO 2 kernels. The volume of kernels as well as the number of kernels in the simulated compact is measured from the 3-D density information. The weight of kernels was calculated from the volume of kernels or the number of kernels. Also, the weight of kernels was measured by extracting the kernels from a compact to review the result of the X-ray CT application

Measurement and evaluation of fuels and technologies for passenger rail service in North Carolina.

Science.gov (United States)

2012-08-01

The purpose of this project is to measure a baseline for fuel use and emission rates on the rebuilt or replaced engines on each locomotive in the NCDOT Rail Division fleet, using ultra-low sulfur diesel (ULSD) fuel; measure real-world, in-use over...

Nondestructive nuclear measurement in the fuel cycle. Part 2

International Nuclear Information System (INIS)

Lyoussi, A.

2005-01-01

Nondestructive measurement techniques are today widely used in practically all steps of the fuel cycle. This article is devoted to the presentation of the control and characterization needs and to the main active nondestructive nuclear methods used: 1 - main active nondestructive nuclear measurement methods: active neutronic measurement (needs and motivations, physical principle, measurement of delayed neutrons following a continuous irradiation, measurement of prompt neutrons (differential die-away technique - DDT), measurement of prompt and delayed neutrons (Sphincs method), neutronic method coupled to gamma spectroscopy), measurement by induced photo-fissions (needs and motivations, physical principle); 2 - conclusion. (J.S.)

Relations between Perceived Competence, Importance Ratings, and Self-Worth among African American School-Age Children

Science.gov (United States)

Grier, Leslie K.

2013-01-01

The purpose of this research was to investigate how domain-specific importance ratings affect relations between perceived competence and self-worth among African American school-age children. Importance ratings have been found to affect the strength of the relationship between perceived competence and self-worth and have implications for…

Burn-up measurements of LEU fuel for short cooling times

International Nuclear Information System (INIS)

Pereda B, C.; Henriquez A, C.; Klein D, J.; Medel R, J.

2005-01-01

The measurements presented in this work were made essentially at in-pool gamma-spectrometric facility, installed inside of the secondary pool of the RECH-1 research reactor, where the measured fuel elements are under 2 meters of water. The main reason for using the in-pool facility was because of its capability to measure the burning of fuel elements without having to wait so long, that is with only 5 cooling days, which are the usual times between reactor operations. Regarding these short cooling times, this work confirms again the possibility of using the 95 Zr as a promising burnup monitor, in spite of the rough approximations used to do it. These results are statistically reasonable within the range calculated using codes. The work corroborates previous results, presented in Santiago de Chile, and it suggests future improvements in that way. (author)

Temperature measurements of the aluminium claddings of fuel elements in nuclear reactor

International Nuclear Information System (INIS)

Chen Daolong

1986-01-01

A method for embedding the sheathed thermocouples in the aluminium claddings of some fuel elements of experimental reactors by ultrasonic welding technique is described. The measurement results of the cladding temperature of fuel elements in reactors are given. By means of this method, the joint between the sheathed thermocouples and the cladding of fuel elements can be made very tight, there are no bulges on the cladding surfaces, and the sheathed thermocouples are embedded strongly and reliably. Therefore an essential means is provided for acquiring the stable and dynamic state data of the cladding temperature of in-core fuel elements

Experimental verification of neutron emission method for measuring of fissile material content in spent fuel

International Nuclear Information System (INIS)

Abou-Zaid, A.A.; Pytel, K.

1999-01-01

A non-destructive method of measurement of fissile nuclides content remained in spent fuel from research reactor is presented. The method, called the neutron emission one, is based on counting of fission neutrons emitted from fissile isotopes: 235 U, 239 Pu, 241 Pu. Fissions are induced mainly by neutrons supplied by the external neutron source. Another effects contribute also to the measured neutron population, e. g. source neutrons from penetrating the fuel without being captured and scattered, neutrons (α,n) reactions and from spontaneous fissions of actinides. Complexity of phenomena occurring within the measurement facility required the detailed numerical simulation and experimental studies prior design of ultimate measurement stand. In the previous paper, the results of Monte Carlo simulation on optimisation of measuring stand for neutron emission method were presented. On the basis of those results, the experimental stand for Maria reactor fuel investigation has been designed and manufactured. The present paper, being the continuation of previous one, contains the description of experimental facility and the results of measurements for the fresh fuel (without burnup) and the fuel mock-up (without fissile materials). Although some discrepancies were found between Monte Carlo and experimental results, the main conclusions concerning the optimal geometry of measuring facility have been confirmed. (author)

Material accountancy measurement techniques in dry-powdered processing of nuclear spent fuels

International Nuclear Information System (INIS)

Wolf, S. F.

1999-01-01

The paper addresses the development of inductively coupled plasma-mass spectrometry (ICPMS), thermal ionization-mass spectrometry (TIMS), alpha-spectrometry, and gamma spectrometry techniques for in-line analysis of highly irradiated (18 to 64 GWD/T) PWR spent fuels in a dry-powdered processing cycle. The dry-powdered technique for direct elemental and isotopic accountancy assay measurements was implemented without the need for separation of the plutonium, uranium and fission product elements in the bulk powdered process. The analyses allow the determination of fuel burn-up based on the isotopic composition of neodymium and/or cesium. An objective of the program is to develop the ICPMS method for direct fissile nuclear materials accountancy in the dry-powdered processing of spent fuel. The ICPMS measurement system may be applied to the KAERI DUPIC (direct use of spent PWR fuel in CANDU reactors) experiment, and in a near-real-time mode for international safeguards verification and non-proliferation policy concerns

77 FR 39343 - Agency Information Collection (Income-Net Worth and Employment Statement) Activity Under OMB Review

Science.gov (United States)

2012-07-02

... DEPARTMENT OF VETERANS AFFAIRS [OMB Control No. 2900-0002] Agency Information Collection (Income-Net Worth and Employment Statement) Activity Under OMB Review AGENCY: Veterans Benefits Administration... . Please refer to ``OMB Control No. 2900-0002.'' SUPPLEMENTARY INFORMATION: Title: Income-Net Worth and...

75 FR 56662 - Agency Information Collection (Income-Net Worth and Employment Statement) Activity Under OMB Review

Science.gov (United States)

2010-09-16

... DEPARTMENT OF VETERANS AFFAIRS [OMB Control No. 2900-0002] Agency Information Collection (Income-Net Worth and Employment Statement) Activity Under OMB Review AGENCY: Veterans Benefits Administration... . Please refer to ``OMB Control No. 2900-0002.'' SUPPLEMENTARY INFORMATION: Title: Income-Net Worth and...

Current density distribution mapping in PEM fuel cells as an instrument for operational measurements

Energy Technology Data Exchange (ETDEWEB)

Geske, M.; Heuer, M.; Heideck, G.; Styczynski, Z. A. [Otto-von-Guericke University Magdeburg, Chair Electric Power Networks and Renewable Energy Sources, Magdeburg (Germany)

2010-07-01

A newly developed measurement system for current density distribution mapping has enabled a new approach for operational measurements in proton exchange membrane fuel cells (PEMFC). Taking into account previously constructed measurement systems, a method based on a multi layer printed circuit board was chosen for the development of the new system. This type of system consists of a sensor, a special electronic device and the control and visualization PC. For the acquisition of the current density distribution values, a sensor device was designed and installed within a multilayer printed circuit board with integrated shunt resistors. Varying shunt values can be taken into consideration with a newly developed and evaluated calibration method. The sensor device was integrated in a PEM fuel cell stack to prove the functionality of the whole measurement system. A software application was implemented to visualize and save the measurement values. Its functionality was verified by operational measurements within a PEMFC system. Measurement accuracy and possible negative reactions of the sensor device during PEMFC operation are discussed in detail in this paper. The developed system enables operational measurements for different operating phases of PEM fuel cells. Additionally, this can be seen as a basis for new opportunities of optimization for fuel cell design and operation modes. (author)

Current Density Distribution Mapping in PEM Fuel Cells as An Instrument for Operational Measurements

Directory of Open Access Journals (Sweden)

Martin Geske

2010-04-01

Full Text Available A newly developed measurement system for current density distribution mapping has enabled a new approach for operational measurements in proton exchange membrane fuel cells (PEMFC. Taking into account previously constructed measurement systems, a method based on a multi layer printed circuit board was chosen for the development of the new system. This type of system consists of a sensor, a special electronic device and the control and visualization PC. For the acquisition of the current density distribution values, a sensor device was designed and installed within a multilayer printed circuit board with integrated shunt resistors. Varying shunt values can be taken into consideration with a newly developed and evaluated calibration method. The sensor device was integrated in a PEM fuel cell stack to prove the functionality of the whole measurement system. A software application was implemented to visualize and save the measurement values. Its functionality was verified by operational measurements within a PEMFC system. Measurement accuracy and possible negative reactions of the sensor device during PEMFC operation are discussed in detail in this paper. The developed system enables operational measurements for different operating phases of PEM fuel cells. Additionally, this can be seen as a basis for new opportunities of optimization for fuel cell design and operation modes.

Assessing alternative measures of wealth in health research.

Science.gov (United States)

Cubbin, Catherine; Pollack, Craig; Flaherty, Brian; Hayward, Mark; Sania, Ayesha; Vallone, Donna; Braveman, Paula

2011-05-01

We assessed whether it would be feasible to replace the standard measure of net worth with simpler measures of wealth in population-based studies examining associations between wealth and health. We used data from the 2004 Survey of Consumer Finances (respondents aged 25-64 years) and the 2004 Health and Retirement Survey (respondents aged 50 years or older) to construct logistic regression models relating wealth to health status and smoking. For our wealth measure, we used the standard measure of net worth as well as 9 simpler measures of wealth, and we compared results among the 10 models. In both data sets and for both health indicators, models using simpler wealth measures generated conclusions about the association between wealth and health that were similar to the conclusions generated by models using net worth. The magnitude and significance of the odds ratios were similar for the covariates in multivariate models, and the model-fit statistics for models using these simpler measures were similar to those for models using net worth. Our findings suggest that simpler measures of wealth may be acceptable in population-based studies of health.

Efficacy of a group-based multimedia HIV prevention intervention for drug-involved women under community supervision: project WORTH.

Science.gov (United States)

El-Bassel, Nabila; Gilbert, Louisa; Goddard-Eckrich, Dawn; Chang, Mingway; Wu, Elwin; Hunt, Tim; Epperson, Matt; Shaw, Stacey A; Rowe, Jessica; Almonte, Maria; Witte, Susan

2014-01-01

This study is designed to address the need for evidence-based HIV/STI prevention approaches for drug-involved women under criminal justice community supervision. We tested the efficacy of a group-based traditional and multimedia HIV/STI prevention intervention (Project WORTH: Women on the Road to Health) among drug-involved women under community supervision. We randomized 306 women recruited from community supervision settings to receive either: (1) a four-session traditional group-based HIV/STI prevention intervention (traditional WORTH); (2) a four-session multimedia group-based HIV/STI prevention intervention that covered the same content as traditional WORTH but was delivered in a computerized format; or (3) a four-session group-based Wellness Promotion intervention that served as an attention control condition. The study examined whether the traditional or multimedia WORTH intervention was more efficacious in reducing risks when compared to Wellness Promotion; and whether multimedia WORTH was more efficacious in reducing risks when compared to traditional WORTH. Primary outcomes were assessed over the 12-month post-intervention period and included the number of unprotected sex acts, the proportion of protected sex acts, and consistent condom use. At baseline, 77% of participants reported unprotected vaginal or anal sex (n = 237) and 63% (n = 194) had multiple sex partners. Women assigned to traditional or multimedia WORTH were significantly more likely than women assigned to the control condition to report an increase in the proportion of protected sex acts (β = 0.10; 95% CI = 0.02-0.18) and a decrease in the number of unprotected sex acts (IRR = 0.72; 95% CI = 0.57-0.90). The promising effects of traditional and multimedia WORTH on increasing condom use and high participation rates suggest that WORTH may be scaled up to redress the concentrated epidemics of HIV/STIs among drug-involved women in the criminal justice system. Clinical

LOFT fuel rod pressure measurement

International Nuclear Information System (INIS)

Billeter, T.R.

1979-01-01

Pressure sensors selected for measuring fuel rod pressure within the LOFT reactor exhibited stable, repeatable operating characteristics during calibrations at temperatures up to 800 0 F and pressures to 2500 psig. All sensors have a nominal sensitivity of .5 millivolts per psi, decreasing monotonically with temperature. Output signal increases linearly with increasing pressure up to 2000 psig. For imposed slow and rapid temperature variations and for pressure applied during these tests, the sensor indicates a pressure at variance with the actual value by up to 15% of reading. However, the imposed temperature rates of change often exceeded the value of -10 0 F/sec. specified for LOFT. The series of tests in an autoclave permit creation of an environment most closely resembling sensor operating conditions within LOFT. For multiple blowdowns and for longtime durations the sensor continued to provide pressure-related output signals. For temperature rates up to -87 0 F/sec, the indicated pressure measurement error remained less than 13% of reading. Adverse effects caused by heating the 1/16 inch O.D. signal cable to 800 0 F contributed only insignificantly to the noted pressure measurement error

Investigating The Integral Control Rod Worth Of The Miniature Neutron Source Reactor MNSR

International Nuclear Information System (INIS)

Nguyen Hoang Hai; Do Quang Binh

2011-01-01

Determining control rod characteristics is an essential problem of nuclear reactor analysis. In this research, the integral control rod worth of the miniature neutron source reactor MNSR is investigated. Some other parameters of the nuclear reactor, such as core excess reactivity, shut down margin, are also calculated. Group constants for all reactor components are generated by the WIMSD code and then are used in the CITATION code to solve the neutron diffusion equations. The maximum relative error of the calculated results compared with the measurement data is about 3.5%. (author)

Enrichment measurement in TRIGA type fuels; Medicion de enriquecimiento en combustibles tipo Triga

Energy Technology Data Exchange (ETDEWEB)

Aguilar H, F.; Mazon R, R. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

2001-05-15

The Department of Energy of the United States of North America, through the program 'Idaho Operations Nuclear Spent Fuel Program' of the Idaho National Engineering and Environmental Laboratory (INEEL), in Idaho Falls; Idaho USA, hires to Global Technologies Inc. (GTI) to develop a prototype device of detection enrichment uranium (DEU Detection of Enrichment of Uranium) to determine quantitatively the enrichment in remainder U-235 in a TRIGA fuel element at the end of it useful life. The characteristics of the prototype developed by GTI are the following ones: It allows to carry out no-destructive measurements of TRIGA type fuel. Easily transportable due to that reduced of it size. The determination of the enrichment (in grams of U-235) it is obtained with a precision of 5%. The National Institute of Nuclear Research (ININ), in its facilities of the Nuclear Center of Mexico, it has TRIGA type fuel of high and low enrichment (standard and FLIP) fresh and with burnt, it also has the infrastructure (hot cells, armor-plating of transport, etc) and qualified personnel to carry out the necessary maneuvers to prove the operation of the DEU prototype. For this its would be used standard type fuel elements and FLIP, so much fresh as with certain burnt one. In the case of the fresh fuels the measurement doesn't represent any risk, the fuels before and after the measurement its don't contain a quantity of fission products that its represent a radiological risk in its manipulation; but in the case of the fuels with burnt the handling of the same ones represents an important radiological risk reason why for its manipulation it was used the transport armor-plating and the hot cells. (Author)

Acoustic sensor for in-pile fuel rod fission gas release measurement

International Nuclear Information System (INIS)

Fourmentel, D.; Villard, J. F.; Ferrandis, J. Y.; Augereau, F.; Rosenkrantz, E.; Dierckx, M.

2009-01-01

We have developed a specific acoustic sensor to improve the knowledge of fission gas release in Pressurized Water Reactor (PWR) fuel rods when irradiated in materials testing reactors. In order to perform experimental programs related to the study of the fission gas release kinetics, the CEA (French Nuclear Energy Commission) acquired the ability to equip a pre-irradiated PWR fuel rod with three sensors, allowing the simultaneous on-line measurements of the following parameters: - fuel temperature with a centre-line thermocouple type C, - internal pressure with a specific counter-pressure sensor, - fraction of fission gas released in the fuel rod with an innovative acoustic sensor. The third detector is the subject of this paper. This original acoustic sensor has been designed to measure the molar mass and pressure of the gas contained in the fuel rod plenum. For in-pile instrumentation, the fraction of fission gas, such as Krypton and Xenon, in Helium, can be deduced online from this measurement. The principle of this acoustical sensor is the following: a piezoelectric transducer generates acoustic waves in a cavity connected to the fuel rod plenum. The acoustic waves are propagated and reflected in this cavity and then detected by the transducer. The data processing of the signal gives the velocity of the acoustic waves and their amplitude, which can be related respectively to the molar mass and to the pressure of the gas. The piezoelectric material of this sensor has been qualified in nuclear conditions (gamma and neutron radiations). The complete sensor has also been specifically designed to be implemented in materials testing reactors conditions. For this purpose some technical points have been studied in details: - fixing of the piezoelectric sample in a reliable way with a suitable signal transmission, - size of the gas cavity to avoid any perturbation of the acoustic waves, - miniaturization of the sensor because of narrow in-pile experimental devices

Comparison between theoretical values foreseen and measured during the commissioning of second cycle of Angra-1

International Nuclear Information System (INIS)

Angelkorte, G.; Dias, A.M.; Sakai, M.

1988-01-01

The purpose of this paper is to present the most important results of the comparison of the data measured during the start-up tests for the second cycle of Angra 1 with the calculated values for this cycle. This short comparison considered the following quantities: Critical boron concentration, Control rod worths, Isothermal temperature coefficient, Reaction rates of the incore detector system. The calculated results showed good agreement with the measured ones remaining within the acceptation criteria limits, confirming the fuel management calculations for the second cycle, namely that for the reload scheme obtained, the core is capable of operating at nominal power safety and without restrictions. (author) [pt

Reciprocal links among differential parenting, perceived partiality, and self-worth: a three-wave longitudinal study.

Science.gov (United States)

Shebloski, Barbara; Conger, Katherine J; Widaman, Keith F

2005-12-01

This study examined reciprocal links between parental differential treatment, siblings' perception of partiality, and self-worth with 3 waves of data from 384 adolescent sibling dyads. Results suggest that birth-order status was significantly associated with self-worth and perception of maternal and paternal differential treatment. There was a consistent across-time effect of self-worth on perception of parental partiality for later born siblings, but not earlier born siblings, and a consistent effect of differential treatment on perception of partiality for earlier born but not later born siblings. The results contribute new insight into the associations between perception of differential parenting and adolescents' adjustment and the role of birth order. Copyright 2006 APA, all rights reserved).

Results of the initial test program for the Sandia Pulsed Reactor III (SPR III)

International Nuclear Information System (INIS)

Estes, B.F.; Reuscher, J.A.

1976-08-01

This document presents a detailed discussion of the reactor including the mechanical and nuclear design characteristics. Also presented are the complete results of the Initial Approach to Critical and the Zero-and-Low Power testing programs. Reactivity worth measurements are given for such parameters as control element integral worth, Safety Block integral worth, and various materials (polyethylene, copper, lead, etc) as a function of position relative to the core. Subcritical reactivity measurements made during the approach to critical generally proved to be in reasonably good agreement with design values due to the good source-fuel-detector geometry possible with a reactor of this type. Subsequent dynamic measurements for reactivity worths are shown to be in good agreement with calculated results

Study on Thermal Degradation Characteristics and Regression Rate Measurement of Paraffin-Based Fuel

Directory of Open Access Journals (Sweden)

Songqi Hu

2015-09-01

Full Text Available Paraffin fuel has been found to have a regression rate that is higher than conventional HTPB (hydroxyl-terminated polybutadiene fuel and, thus, presents itself as an ideal energy source for a hybrid rocket engine. The energy characteristics of paraffin-based fuel and HTPB fuel have been calculated by the method of minimum free energy. The thermal degradation characteristics were measured for paraffin, pretreated paraffin, HTPB and paraffin-based fuel in different working conditions by the using differential scanning calorimetry (DSC and a thermogravimetric analyzer (TGA. The regression rates of paraffin-based fuel and HTPB fuel were tested by a rectangular solid-gas hybrid engine. The research findings showed that: the specific impulse of paraffin-based fuel is almost the same as that of HTPB fuel; the decomposition temperature of pretreated paraffin is higher than that of the unprocessed paraffin, but lower than that of HTPB; with the increase of paraffin, the initial reaction exothermic peak of paraffin-based fuel is reached in advance, and the initial reaction heat release also increases; the regression rate of paraffin-based fuel is higher than the common HTPB fuel under the same conditions; with the increase of oxidizer mass flow rate, the regression rate of solid fuel increases accordingly for the same fuel formulation.

Intimacy development in late adolescence: Longitudinal associations with perceived parental autonomy support and adolescents' self-worth.

Science.gov (United States)

Van Petegem, Stijn; Brenning, Katrijn; Baudat, Sophie; Beyers, Wim; Zimmer-Gembeck, Melanie J

2018-03-21

The present longitudinal study tested for the role of perceived parental autonomy-support and late adolescents' self-worth in their intimacy development. A sample of 497 Belgian late adolescents (M age  = 17.9, 43.5% girls) participated in this two-wave study. Results indicated that perceived autonomy-supportive parenting did not relate significantly to change in adolescents' experienced intimacy (in terms of closeness and mutuality), but was associated with a decrease in unmitigated agency (an excessive focus on the self) and unmitigated communion (an excessive focus on the other) across time. Adolescents' self-worth predicted an increase in experienced intimacy and a decrease in unmitigated agency and communion, and the initial level of experienced intimacy predicted an increase in self-worth. Finally, results suggested that adolescents' self-worth may mediate some of the longitudinal relations between perceived parental autonomy-support and adolescents' intimate functioning. No evidence was found for moderation by romantic involvement, gender or age. Copyright © 2018. Published by Elsevier Ltd.

[Does self-worth mediate the effects of socio-environmental experiences on depression among fifth and sixth grade students?].

Science.gov (United States)

Nishino, Yasuyo; Kobayashi, Sachiko; Kitagawa, Tomoko

2009-08-01

This study investigated self-worth as a mediator between socio-environmental experiences and depression. A sample of 255 fifth and sixth grade students completed self-report questionnaires assessing self-worth, depression, and socio-environmental experiences of social support and stressors. Data for both males and females showed a direct effect of "friend support" on depression. However, for males, but not females, self-worth also mediated the influence of "friend support" on depression.

SUB-LEU-METAL-THERM-001 SUBCRITICAL MEASUREMENTS OF LOW ENRICHED TUBULAR URANIUM METAL FUEL ELEMENTS BEFORE and AFTER IRRADIATION

International Nuclear Information System (INIS)

SCHWINKENDORF, K.N.

2006-01-01

With the shutdown of the Hanford PUREX (Plutonium-Uranium Extraction Plant) reprocessing plant in the 1970s, adequate storage capacity for spent Hanford N Reactor fuel elements in the K and N Reactor pools became a concern. To maximize space utilization in the pools, accounting for fuel burnup was considered. Calculations indicated that at typical fuel exposures for N Reactor, the spent-fuel critical mass would be twice the critical mass for green fuel. A decision was reached to test the calculational result with a definitive experiment. If the results proved positive, storage capacity could be increased and N Reactor operation could be prolonged. An experiment to be conducted in the N Reactor spent-fuel storage pool was designed and assembled and the services of the Battelle Northwest Laboratories (BNWL) (now Pacific Northwest National Laboratory [PNNL]) critical mass laboratory were procured for the measurements. The experiments were performed in April 1975 in the Hanford N Reactor fuel storage pool. The fuel elements were MKIA fuel assemblies, comprising two concentric tubes of low-enriched metallic uranium. Two separate sets of measurements were performed: one with ''green'' (fresh) fuel and one with spent fuel. Both the green and spent fuel, were measured in the same geometry. The spent-fuel MKIA assemblies had an average burnup of 2865 MWd (megawatt days)/t. A constraint was imposed restricting the measurements to a subcritical limit of k eff = 0.97. Subcritical count rate data was obtained with pulsed-neutron and approach-to-critical measurements. Ten (10) configurations with green fuel and nine (9) configurations with spent fuel are described and evaluated. Of these, 3 green fuel and 4 spent fuel loading configurations were considered to serve as benchmark models. However, shortcomings in experimental data failed to meet the high standards for a benchmark problem. Nevertheless, the data provided by these subcritical measurements can supply useful

Possibility of implementation of 6-year fuel cycle at NPP with VVER-440 reactor

Energy Technology Data Exchange (ETDEWEB)

Heraltova, L., E-mail: lenka.heraltova@fjfi.cvut.cz [UJV Rez a.s., Hlavni 130, 250 68 Husinec-Rez (Czech Republic); Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, Brehova 7, 115 19 Praha 1 (Czech Republic)

2015-12-15

Highlights: • Possibility of extension of fuel cycle. • Increase of enrichment above 5% {sup 235}U. • Core properties calculated by diffusion code ANDREA. • Back end fuel cycle characteristic. - Abstract: This paper discusses possibility of an extension of a fuel cycle at a VVER-440 reactor for up to 6 years. The prolongation of a fuel cycle was realized by optimization of a fuel design and increasing of a fuel enrichment. The modified design of the fuel assembly covers change of pellet geometry, decreasing of parasitic absorption in construction materials, improved moderation of fuel pins and also increase of enrichment. Fuel assemblies with enrichment up to 7% {sup 235}U are considered for prolonged fuel batches. Three different batch lengths were considered for evaluation of core properties – 12, 18 and 24 months, and two types of burnable absorbers were included – Gd{sub 2}O{sub 3} and Er{sub 2}O{sub 3}. Comparison of proposed fuel assemblies was realized by length of a batch, average burnup, maximal power of fuel assembly or fuel pin, control fuel assembly worth, reactivity coefficients, and effective delayed neutrons fraction. Comparison of characteristics of a burned fuel discharged from a reactor core is discussed in the last part of the paper.

Development of a water boil-off spent-fuel calorimeter system. [To measure decay heat generation rate

Energy Technology Data Exchange (ETDEWEB)

Creer, J.M.; Shupe, J.W. Jr.

1981-05-01

A calorimeter system was developed to measure decay heat generation rates of unmodified spent fuel assemblies from commercial nuclear reactors. The system was designed, fabricated, and successfully tested using the following specifications: capacity of one BWR or PWR spent fuel assembly; decay heat generation range 0.1 to 2.5 kW; measurement time of < 12 h; and an accuracy of +-10% or better. The system was acceptance tested using a dc reference heater to simulate spent fuel assembly heat generation rates. Results of these tests indicated that the system could be used to measure heat generation rates between 0.5 and 2.5 kW within +- 5%. Measurements of heat generation rates of approx. 0.1 kW were obtained within +- 15%. The calorimeter system has the potential to permit measurements of heat generation rates of spent fuel assemblies and other devices in the 12- to 14-kW range. Results of calorimetry of a Turkey Point spent fuel assembly indicated that the assembly was generating approx. 1.55 kW.

The Role of Contingent Self-Worth in the Relation between Victimization and Internalizing Problems in Adolescents

Science.gov (United States)

Ghoul, Assia; Niwa, Erika Y.; Boxer, Paul

2013-01-01

Peer victimization can challenge mental health, yet limited research has considered contingent self-worth as a moderator of that relation. This study examined the relation of peer victimization to major depressive disorder, generalized anxiety disorder, and social phobia during adolescence, and contingent self-worth as a hypothesized moderator of…

Use of multi-functional flexible micro-sensors for in situ measurement of temperature, voltage and fuel flow in a proton exchange membrane fuel cell.

Science.gov (United States)

Lee, Chi-Yuan; Chan, Pin-Cheng; Lee, Chung-Ju

2010-01-01

Temperature, voltage and fuel flow distribution all contribute considerably to fuel cell performance. Conventional methods cannot accurately determine parameter changes inside a fuel cell. This investigation developed flexible and multi-functional micro sensors on a 40 μm-thick stainless steel foil substrate by using micro-electro-mechanical systems (MEMS) and embedded them in a proton exchange membrane fuel cell (PEMFC) to measure the temperature, voltage and flow. Users can monitor and control in situ the temperature, voltage and fuel flow distribution in the cell. Thereby, both fuel cell performance and lifetime can be increased.

Reactor physics tests of TRIGA Mark-II Reactor in Ljubljana

International Nuclear Information System (INIS)

Ravnik, M.; Mele, I.; Trkov, A.; Rant, J.; Glumac, B.; Dimic, V.

2008-01-01

TRIGA Mark-II Reactor in Ljubljana was recently reconstructed. The reconstruction consisted mainly of replacing the grid plates, the control rod mechanisms and the control unit. The standard type control rods were replaced by the fuelled follower type, the central grid location (A ring) was adapted for fuel element insertion, the triangular cutouts were introduced in the upper plate design. However, the main novelty in reactor physics and operational features of the reactor was the installation of a pulse rod. Having no previous operational experience in pulsing, a detailed and systematic sequence of tests was defined in order to check the predicted design parameters of the reactor with measurements. The following experiments are treated in this paper: initial criticality, excess reactivity measurements, control rod worth measurement, fuel temperature distribution, fuel temperature reactivity coefficient, pulse parameters measurement (peak power, prompt energy, peak temperature). Flux distributions in steady state and pulse mode were measured as well, however, they are treated only briefly due to the volume of the results. The experiments were performed with completely fresh fuel of 12 w% enriched Standard Stainless Steel type. The core configuration was uniform (one fuel element type, including fuelled followers) and compact (no irradiation channels or gaps), as such being particularly convenient for testing the computer codes for TRIGA reactor calculations. Comparison of analytical predictions, obtained with WIMS, SLXTUS, TRIGAP and PULSTRI codes to measured values showed agreement within the error of the measurement and calculation. The paper has the following contents: 1. Introduction; 2. Steady State Experiments; 2.1. Core loading and critical experiment; 2.2. Flux range determination for tests at zero power; 2.3. Digital reactivity meter checkout; 2.4. Control rod worth measurements; 2.5. Excess reactivity measurement; 2.6. Thermal power calibration; 2

Burnup measurements on spent fuel elements of the RP-10 research reactor

Energy Technology Data Exchange (ETDEWEB)

Vela Mora, Mariano; Gallardo Padilla, Alberto; Palomino, Jose Luis Castro, E-mail: mvela@ipen.gob.p [Instituto Peruano de Energia Nuclear (IPEN/Peru), Lima (Peru). Grupo de Calculo, Analisis y Seguridad de Reactores; Terremoto, Luis Antonio Albiac, E-mail: laaterre@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2011-07-01

This work describes the measurement, using nondestructive gamma-ray spectroscopy, of the average burnup attained by Material Testing Reactor (MTR) fuel elements irradiated in the RP-10 research reactor. Measurements were performed at the reactor storage pool area using {sup 137}Cs as the only burnup monitor, even for spent fuel elements with cooling times much shorter than two years. The experimental apparatus was previously calibrated in efficiency to obtain absolute average burnup values, which were compared against corresponding ones furnished by reactor physics calculations. The mean deviation between both values amounts to 6%. (author)

Burnup measurements on spent fuel elements of the RP-10 research reactor

International Nuclear Information System (INIS)

Vela Mora, Mariano; Gallardo Padilla, Alberto; Palomino, Jose Luis Castro

2011-01-01

This work describes the measurement, using nondestructive gamma-ray spectroscopy, of the average burnup attained by Material Testing Reactor (MTR) fuel elements irradiated in the RP-10 research reactor. Measurements were performed at the reactor storage pool area using 137 Cs as the only burnup monitor, even for spent fuel elements with cooling times much shorter than two years. The experimental apparatus was previously calibrated in efficiency to obtain absolute average burnup values, which were compared against corresponding ones furnished by reactor physics calculations. The mean deviation between both values amounts to 6%. (author)

The Relationship Between Poor Quality of Life and Desire to Hasten Death: A Multiple Mediation Model Examining the Contributions of Depression, Demoralization, Loss of Control, and Low Self-worth.

Science.gov (United States)

Robinson, Sophie; Kissane, David W; Brooker, Joanne; Hempton, Courtney; Burney, Susan

2017-02-01

The risk of suicide is elevated in palliative care patients compared with the general population. Various psychological factors, including depression, demoralization, loss of control, and low self-worth, have been associated with a desire to hasten death. The aim of this study was to investigate whether depression, demoralization, loss of control, and low self-worth mediated the relationship between global quality of life and desire to hasten death. A sample of 162 palliative care patients completed measures of global quality of life, depression, demoralization, perceived control, self-worth, and desire to hasten death. A multiple mediation model with bootstrapping sampling tested the total (combined) indirect effect and individual indirect effects of depression, the two subscales of demoralization (Meaning and Purpose, and Distress and Coping Ability), perceived control, and self-worth. Depressive symptoms, loss of meaning and purpose, loss of control, and low self-worth mediated the direct effect of global quality of life on desire to hasten death. The Distress and Coping Ability component of demoralization was not a significant mediator. Depression, loss of meaning and purpose, loss of control, and low self-worth are strong clinical markers for desire to hasten death. Targeting these symptoms through existentially oriented therapies, such as meaning-centered therapy, may ameliorate suicidal thinking. Copyright © 2016 American Academy of Hospice and Palliative Medicine. Published by Elsevier Inc. All rights reserved.

Technique of measuring and evaluating the R3E dia. 112 experimental fuel element

International Nuclear Information System (INIS)

Stanc, S.; Badiar, S.

1974-10-01

The charging is considered of the R3E experimental fuel element having a diameter of 112 mm which will serve the simultaneous temperature measurement on 15 sites and neutron flux measurement on 7 sites aimed at verifying the improved R3-type fuel element with extended cladding and a new toroidal spacing. The temperature will be measured using jacket thermocouples while neutron flux will be measured by self-powered detectors. The methods are described of checking the measuring chain by determining the loop resistance of the thermocouples and SPN detectors in order to verify the correctness of the measured data. (J.B.)

Calculation of control rod worth with mutual interaction

International Nuclear Information System (INIS)

Balthar, M.C.V.; Oliveira Vellozo, S. de; Carvalho Vital, H. de

1989-01-01

This work presents a two-dimensional model for determining the total worth of a set of N absorbing rods. The model simplifies the evaluation of the interaction coefficient among rods by analysing them in pairs and attributes to it a purely geometrical character. Comparisons with conventional calculational methods indicate that the results are in error by less than 6%. (author) [pt

Dereplications Can Amplify The Extent And Worth Of Traditional ...

African Journals Online (AJOL)

Moreover, since the talents and methods of those conducting these initial studies varied widely, little effort was made to provide adequate information on how selective processes and preferences as well as modes of collection, preparation and use were achieved. Without these data, the potential of their clinical worth, ...

Field test and evaluation of the IAEA coincidence collar for the measurement of unirradiated BWR fuel assemblies

International Nuclear Information System (INIS)

Menlove, H.O.; Keddar, A.

1982-12-01

The neutron coincidence counter has been field tested and evaluated for the measurement of boiling-water-reactor (BWR) fuel assemblies at the ASEA-ATOM Fuel Fabrication Facility. The system measures the 235 U content per unit length of full fuel assemblies using neutron interrogation and coincidence counting. The 238 U content is measured in the passive mode without the AmLi neutron interrogatioin source. The field tests included both standard production movable fuel rods to investigate enrichment and absorber variations. Results gave a response standard deviation of 0.9% for the active case and 2.1% for the passive case in 1000-s measurement times. 10 figures, 2 tables

Optimization of the binary breeder reactor. VIII annular core fueled with 233U - 238U and Pu-238U

International Nuclear Information System (INIS)

Nascimento, J.A. do; Ishiguro, Y.

1988-04-01

First cycle burnup characteristics of a 1200 MWe binary breeder reactor with annular core fueled with metallic 233 U- 238 U-Zr, Pu- 238 U-Zr and Th in the blankets have been analysed. The Doppler effect is small as expected in a metal fueled fast reactor. The sodium void reactivity is, in general, smaller than in metal fueled homogeneous fast reactors of 1 m core height. The estimates of the required and available control rod worths show a large shutdown margin throughout the operational cycle. There are flexibilities in the blanket fueling and well balanced breeding in the two cycles, uranium and thorium, with doubling times of about 20 years are possible. (author) [pt

Accuracy and computational time of a hierarchy of growth rate definitions for breeder reactor fuel

International Nuclear Information System (INIS)

Maudlin, P.J.; Borg, R.C.; Ott, K.O.

1979-01-01

For a hierarchy of four logically different definitions for calculating the asymptotic growth of fast breeder reactor fuel, an investigation is performed concerning the comparative accuracy and computational effort associated with each definition. The definition based on detailed calculation of the accumulating fuel in an expanding park of reactors asymptotically yields the most accurate value of the infinite time growth rate, γ/sup infinity/, which is used as a reference value. The computational effort involved with the park definition is very large. The definition based on the single reactor calculation of the equilibrium surplus production rate and fuel inventory gives a value for γ/sup infinity of comparable accuracy to the park definition and uses significantly less central processor unit (CPU) time. The third definition is based on a continuous treatment of the reactor fuel cycle for a single reactor and gives a value for γ/sup infinity/ that accurately approximates the second definition. The continuous definition requires very little CPU time. The fourth definition employs the isotopic breeding worths, w/sub i//sup */, for a projection of the asymptotic growth rate. The CPU time involved in this definition is practically nil if its calculation is based on the few-cycle depletion calculation normally performed for core design and critical enrichment evaluations. The small inaccuracy (approx. = 1%) of the breeding-worth-based definition is well within the inaccuracy range that results unavoidably from other sources such as nuclear cross sections, group constants, and flux calculations. This fully justifies the use of this approach in routine calculations

Evaluation of Instrumentation for Measuring Undissolved Water in Aviation Turbine Fuels per ASTM D3240

Science.gov (United States)

2015-11-05

Undissolved Water in Aviation Turbine Fuels per ASTM D3240 5a. CONTRACT NUMBER 5b. GRANT NUMBER 5c. PROGRAM ELEMENT NUMBER 6. AUTHOR(S) Joel Schmitigal... water ) in Aviation Turbine Fuels per ASTM D3240 15. SUBJECT TERMS fuel, JP-8, aviation fuel, contamination, free water , undissolved water , Aqua-Glo 16...Michigan 48397-5000 Evaluation of Instrumentation for Measuring Undissolved Water in Aviation Turbine Fuels per ASTM D3240 Joel Schmitigal Force

Measurement station for interim inspections of Lightbridge metallic fuel rods at the Halden Boiling Water Reactor

Science.gov (United States)

Hartmann, C.; Totemeier, A.; Holcombe, S.; Liverud, J.; Limi, M.; Hansen, J. E.; Navestad, E. AB(; )

2018-01-01

Lightbridge Corporation has developed a new Uranium-Zirconium based metallic fuel. The fuel rods aremanufactured via a co-extrusion process, and are characterized by their multi-lobed (cruciform-shaped) cross section. The fuel rods are also helically-twisted in the axial direction. Two experimental fuel assemblies, each containing four Lightbridge fuel rods, are scheduled to be irradiated in the Halden Boiling Water Reactor (HBWR) starting in 2018. In addition to on-line monitoring of fuel rod elongation and critical assembly conditions (e.g. power, flow rates, coolant temperatures, etc.) during the irradiation, several key parameters of the fuel will be measured out-of-core during interim inspections. An inspection measurement station for use in the irradiated fuel handling compartment at the HBWR has therefore been developed for this purpose. The multi-lobed cladding cross section combined with the spiral shape of the Lightbridge metallic fuel rods requires a high-precision guiding system to ensure good position repeatability combined with low-friction guiding. The measurement station is equipped with a combination of instruments and equipment supplied from third-party vendors and instruments and equipment developed at Institute for Energy Technology (IFE). Two sets of floating linear voltage differential transformer (LVDT) pairs are used to measure swelling and diameter changes between the lobes and the valleys over the length of the fuel rods. Eddy current probes are used to measure the thickness of oxide layers in the valleys and on the lobe tips and also to detect possible surface cracks/pores. The measurement station also accommodates gamma scans. Additionally, an eddy-current probe has been developed at IFE specifically to detect potential gaps or discontinuities in the bonding layer between the metallic fuel and the Zirconium alloy cladding. Potential gaps in the bonding layer will be hidden behind a 0.5-1.0 mm thick cladding wall. It has therefore been

Holdup measurement for nuclear fuel manufacturing plants

International Nuclear Information System (INIS)

Zucker, M.S.; Degen, M.; Cohen, I.; Gody, A.; Summers, R.; Bisset, P.; Shaub, E.; Holody, D.

The assay of nuclear material holdup in fuel manufacturing plants is a laborious but often necessary part of completing the material balance. A range of instruments, standards, and a methodology for assaying holdup has been developed. The objectives of holdup measurement are ascertaining the amount, distribution, and how firmly fixed the SNM is. The purposes are reconciliation of material unbalance during or after a manufacturing campaign or plant decommissioning, to decide security requirements, or whether further recovery efforts are justified

Equation of state for L.M.F.B.R. fuel (measurement of fission gas release during transients)

International Nuclear Information System (INIS)

Combette, P.; Barthelemy, P.

1979-01-01

A sample of fuel (UO 2 or UPuO 2 ) can be heated by fission in a heating transient up to energy deposition 4000 j/g, in the Silene reactor. The Kistler type capsule, the calorimeter device and the radiochemical analysis of fission products enable the pressure pulse and the fuel energy deposition to be measured. So, the relationship between the fuel vapour pressure and the fuel specific energy can be deduced. Peaks pressure (about 1 MPa) coming from fresh UO 2 vaporization, have been measured on a 7 milliseconds time scale. There is a good agreement with the E.O.S. for fresh UO 2 , which is well known for low pressure (1 MPa). Numerous tests have been done with 93% enriched UO 2 and a first test with highly active fuel containing plutonium (15 at %) has been performed. The capsule allows the released gas coming from the irradiated fuel to be retained for measurements and analysis. To investigate the mode of fuel disruption, in-pile fission-heated fuel pellets has been recorded by high speed cinematography

An early view of the impact of deregulation and managed care on hospital profitability and net worth.

Science.gov (United States)

Jordan, W J

2001-01-01

This study shows the impact of the removal of hospital rate regulation followed by the growth of managed care on hospitals' profitability and net worth. New Jersey emerged from a regulated prospective payment system in 1992. The transition to a freely competitive market structure had a negative impact on hospital profitability, net worth, patient length of stay, and other measures of capacity utilization. Similarly, the doubling of the HMO penetration rate in the state between 1995 and 1997 is shown to have negatively influenced hospital financial viability. Hospitals have responded in part by increasing usage of outpatient services. The use of discounted fee-for-service instead of per diem reimbursement for outpatient services provides an incentive for hospitals to favor outpatient over inpatient services. The effect of these changes is detailed, along with data showing that the larger discounts given by hospitals to managed care organizations, Medicare, and Medicaid played an important role in explaining the diminished profitability of hospitals.

Is your upgrade worth it? process mining can tell

NARCIS (Netherlands)

Genuchten, van M.J.I.M.; Mans, R.S.; Reijers, H.A.; Wismeijer, D.

2014-01-01

Software vendors typically release updates and upgrades of their software once or twice a year. Users are then faced with the question of whether the upgrade is worth the price and the trouble. The software industry doesn't provide much evidence that it's worthwhile to upgrade to new releases. The

Is your upgrade worth it? Process mining can tell

NARCIS (Netherlands)

Genuchten, M.J.I.M.; Mans, R.S.; Reijers, H.A.; Wismeijer, D.

2014-01-01

Software vendors typically release updates and upgrades of their software once or twice a year. Users are then faced with the question of whether the upgrade is worth the price and the trouble. The software industry doesn't provide much evidence that it's worthwhile to upgrade to new releases. The

A glimpse behind the mask: facets of narcissism and feelings of self-worth.

Science.gov (United States)

Zeigler-Hill, Virgil; Besser, Avi

2013-01-01

This study examined the connections that the facets of narcissism captured by the Narcissistic Personality Inventory (NPI; Raskin & Hall, 1979) and the Pathological Narcissism Inventory (PNI; Pincus et al., 2009) have with self-esteem. This was accomplished by asking 372 participants to complete measures of narcissism and self-esteem level as well as daily diary measures concerning their state self-esteem and daily experiences. Our analyses found that the facets of narcissism differed in their associations with average level of self-esteem, fluctuations in state self-esteem over time, and self-esteem reactions following daily events. These results suggest that it is important to consider specific facets of narcissism when examining feelings of self-worth rather than relying on broader composite measures of narcissistic personality features. Implications of these results for our understanding of the dynamics of the narcissistic personality are discussed.

A technique of melting temperature measurement and its application for irradiated high-burnup MOX fuels

International Nuclear Information System (INIS)

Namekawa, Takashi; Hirosawa, Takashi

1999-01-01

A melting temperature measurement technique for irradiated oxide fuels is described. In this technique, the melting temperature was determined from a thermal arrest on a heating curve of the specimen which was enclosed in a tungsten capsule to maintain constant chemical composition of the specimen during measurement. The measurement apparatus was installed in an alpha-tight steel box within a gamma-shielding cell and operated by remote handling. The temperature of the specimen was measured with a two-color pyrometer sighted on a black-body well at the bottom of the tungsten capsule. The diameter of the black-body well was optimized so that the uncertainties of measurement were reduced. To calibrate the measured temperature, two reference melting temperature materials, tantalum and molybdenum, were encapsulated and run before and after every oxide fuel test. The melting temperature data on fast reactor mixed oxide fuels irradiated up to 124 GWd/t were obtained. In addition, simulated high-burnup mixed oxide fuel up to 250 GWd/t by adding non-radioactive soluble fission products was examined. These data shows that the melting temperature decrease with increasing burnup and saturated at high burnup region. (author)

Procedure and apparatus for measuring the radial gap between fuel and surrounding cladding in a fuel rod for a nuclear reactor

International Nuclear Information System (INIS)

Olshausen, K.D.

1976-01-01

A device is described for measuring non-destructively the annular fuel-cladding gap in an irradiated or fresh fuel rod. The principle applied is that a force is applied to an arm which presses the cladding diametrically, thus deforming it until it touches the fuel pellet. By presenting the values of the force applied and the deformation produced on an XY recorder, the width of the gap is obtained. Alternatively the gap width may be obtained digitally. Since the gap is so small that the deformation is within the elastic range, the fuel rod may be reloaded in the reactor for further irradiation. (JIW)

The nuclear fuel cycle light and shadow

International Nuclear Information System (INIS)

Giraud, A.

1977-01-01

The nuclear fuel cycle industry has a far reaching effect on future world energy developments. The growth in turnover of this industry follows a known patterm; by 1985 this turnover will have reached a figure of 2 billion dollars. Furthermore, the fuel cycle plays a determining role in ensuring the physical continuity of energy supplies for countries already engaged in the nuclear domain. Finally, the development of this industry is subject to economic and political constraints which imply the availability of raw materials, technological know-how, and production facilities. Various factors which could have an adverse influence on the cycle: technical, economic, or financial difficulties, environmental impact, nuclear safety, theft or diversion of nuclear materials, nuclear weapon, proliferation risks, are described, and the interaction between the development of the cycle, energy independance, and the fulfillment of nuclear energy programs is emphasized. It is concluded that the nuclear fuel cycle industry is confronted with difficulties due to its extremely rapid growth rate (doubling every 5 years); it is a long time since such a growth rate has been experienced by any heavy industry. The task which lays before us is difficult, but the fruit is worth the toil, as it is the fuel cycle which will govern the growth of the nuclear industry [fr

The travesty of discarding used CANDU fuel

Energy Technology Data Exchange (ETDEWEB)

Ottensmeyer, P. [Univ. of Toronto, Toronto, Ontario (Canada)

2016-09-15

The current plan worldwide for virtually all used nuclear fuels is costly deep burial to attempt to isolate their long-term radiotoxicity permanently. Alternatively Canada's 50,000 tons spent CANDU fuel, of which only 0.74% of the heavy atoms have been fissioned to extract their energy, could supply 130 times more non-carbon energy using proven economical recycling and fast-neutron technologies. The result in this country alone would currently be the creation of $74 trillion of reliable electricity on demand without greenhouse gas emissions. It would avoid adding 475 billion tons CO{sub 2} to the atmosphere compared to the use of coal, to mitigate climate change. Worldwide recycling of stored spent nuclear fuel and replenishing with depleted uranium in fast-neutron reactors could avoid emitting over 20 trillion tons CO{sub 2}, or over six times the current total atmospheric CO{sub 2} content. As added bonus the long-term radiotoxicity of the used CANDU fuel is effectively eliminated, making a long-term deep geological repository unnecessary. Even the shorter-lived radioisotope fission products become valuable stable atoms and minerals that would fetch $3 million per ton. Such an alternative is certainly worth pursuing. (author)

Shock Tube Measurements for Liquid Fuels Combustion

National Research Council Canada - National Science Library

Hanson, Ronald K

2006-01-01

...) fundamental studies of fuel spray evaporation rates and ignition times of low-vapor pressure fuels such as JP-8, diesel fuel and normal alkane surrogates in a new aerosol shock tube using state...

Automatic Gamma-Scanning System for Measurement of Residual Heat in Spent Nuclear Fuel

International Nuclear Information System (INIS)

Osifo, Otasowie

2007-03-01

In Sweden, spent nuclear fuel will be encapsulated and placed in a deep geological repository. In this procedure, reliable and accurate spent fuel data such as discharge burnup, cooling time and residual heat must be available. The gamma scanning method was proposed in earlier work as a fast and reliable method for the experimental determination of such spent fuel data. This thesis is focused on the recent achievements in the development of a pilot gamma scanning system and its application in measuring spent fuel residual heat. The achievements include the development of dedicated spectroscopic data-acquisition and analysis software and the use of a specially designed calorimeter for calibrating the gamma scanning system. The pilot system is described, including an evaluation of the performance of the spectrum analysis software. Also described are the gamma-scanning measurements on 31 spent PWR fuel assemblies performed using the pilot system. The results obtained for the determination of residual heat are presented, showing an agreement of (2-3) % with both calorimetric and calculated data. In addition, the ability to verify declared data such as discharge burnup and cooling time is demonstrated

Associations between the parent-child relationship and adolescent self-worth: a genetically informed study of twin parents and their adolescent children.

Science.gov (United States)

McAdams, Tom A; Rijsdijk, Fruhling V; Narusyte, Jurgita; Ganiban, Jody M; Reiss, David; Spotts, Erica; Neiderhiser, Jenae M; Lichtenstein, Paul; Eley, Thalia C

2017-01-01

Low self-worth during adolescence predicts a range of emotional and behavioural problems. As such, identifying potential sources of influence on self-worth is important. Aspects of the parent-child relationship are often associated with adolescent self-worth but to date it is unclear whether such associations may be attributable to familial confounding (e.g. genetic relatedness). We set out to clarify the nature of relationships between parental expressed affection and adolescent self-worth, and parent-child closeness and adolescent self-worth. We used data from the Twin and Offspring Study in Sweden, a children-of-twins sample comprising 909 adult twin pairs with adolescent children. Using these data we were able to apply structural equation models with which we could examine whether associations remained after accounting for genetic transmission. Results demonstrated that parent-child closeness and parental-expressed affection were both phenotypically associated with adolescent self-worth. Associations could not be attributed to genetic relatedness between parent and child. Parent-child closeness and parental affection are associated with adolescent self-worth above and beyond effects attributable to genetic relatedness. Data were cross-sectional, so the direction of effects cannot be confirmed but findings support the notion that positive parent-child relationships increase adolescent self-worth. © 2016 The Authors. Journal of Child Psychology and Psychiatry published by John Wiley & Sons Ltd on behalf of Association for Child and Adolescent Mental Health.

Mixed Reload Design Using MOX and UOX Fuel Assemblies

International Nuclear Information System (INIS)

Ramon, Ramirez Sanchez J.; Perry, R.T.

2002-01-01

As part of the studies involved in plutonium utilization assessment for a Boiling Water Reactor, a conceptual design of MOX fuel was developed, this design is mechanically the same design of 10 X 10 BWR fuel assemblies but different fissile material. Several plutonium and gadolinium concentrations were tested to match the 18 months cycle length which is the current cycle length of LVNPP, a reference UO 2 assembly was modeled to have a full cycle length to compare results, an effective value of 0.97 for the multiplication factor was set as target for 470 Effective Full Power days for both cycles, here the gadolinium concentration was a key to find an average fissile plutonium content of 6.55% in the assembly. A reload of 124 fuel assemblies was assumed to simulate the complete core, several load fractions of MOX fuel mixed with UO 2 fresh fuel were tested to verify the shutdown margin, the UO 2 fuel meets the shutdown margin when 124 fuel assemblies are loaded into the core, but it does not happen when those 124 assemblies are replaced with MOX fuel assemblies, so the fraction of MOX was reduced step by step up to find a mixed load that meets both length cycle and shutdown margin. Finally the conclusion is that control rods losses some of their worth in presence of plutonium due to a more hardened neutron spectrum in MOX fuel and this fact limits the load of MOX fuel assemblies in the core, this results are shown in this paper. (authors)

Impact of delays in plutonium use on the stationary growth rate of fast breeder reactor fuel

International Nuclear Information System (INIS)

Borg, R.C.; Ott, K.O.

1977-07-01

The hierarchy of the four growth rate expressions originally derived from an instantaneous reuse scheme is expanded to account for finite burnup in the core and blanket, β-decay of 241 Pu, core and blanket loading schemes, reuse delays due to reprocessing and fabricating fuel and external fuel cycle losses. The most general growth rate expression, obtained from the asymptotic slope of the accumulating fuel material in an expanding park of breeder reactors, is formally the same in both cases. Formulation of the growth rate based on the condensation of the detailed information of the equilibrium fuel cycle for a single reactor, is more complicated than without delays due to the composition difference between the average residing and excess discharge material. The third growth rate expression results from a slightly more complicated fuel-cycle eigenvalue problem than without delays. The last definition employs isotopic breeding worth factors obtained from the adjoint fuel cycle eigenvalue problem

Associations of adolescent hopelessness and self-worth with pregnancy attempts and pregnancy desire.

Science.gov (United States)

Fedorowicz, Anna R; Hellerstedt, Wendy L; Schreiner, Pamela J; Bolland, John M

2014-08-01

We examined the associations of pregnancy desire (ambivalence or happiness about a pregnancy in the next year) and recent pregnancy attempts with hopelessness and self-worth among low-income adolescents. To evaluate independent associations among the study variables, we conducted gender-stratified multivariable logistic regression analyses with data derived from 2285 sexually experienced 9- to 18-year-old participants in the Mobile Youth Survey between 2006 and 2009. Fifty-seven percent of youths reported a desire for pregnancy and 9% reported pregnancy attempts. In multivariable analyses, hopelessness was positively associated and self-worth was negatively associated with pregnancy attempts among both female and male youths. Hopelessness was weakly associated (P = .05) with pregnancy desire among female youths. The negative association of self-worth and the positive association of hopelessness with pregnancy attempts among young men as well as young women and the association of hopelessness with pregnancy desire among young women raise questions about why pregnancy is apparently valued by youths who rate their social and cognitive competence as low and who live in an environment with few options for material success.

A STUDY OF THE DISCREPANCY BETWEEN FEDERAL AND STATE MEASUREMENTS OF ON-HIGHWAY FUEL CONSUMPTION

Energy Technology Data Exchange (ETDEWEB)

Hwang, HL

2003-08-11

Annual highway fuel taxes are collected by the Treasury Department and placed in the Highway Trust Fund (HTF). There is, however, no direct connection between the taxes collected by the Treasury Department and the gallons of on-highway fuel use, which can lead to a discrepancy between these totals. This study was conducted to determine how much of a discrepancy exists between the total fuel usages estimated based on highway revenue funds as reported by the Treasury Department and the total fuel usages used in the apportionment of the HTF to the States. The analysis was conducted using data from Highway Statistics Tables MF-27 and FE-9 for the years 1991-2001. It was found that the overall discrepancy is relatively small, mostly within 5% difference. The amount of the discrepancy varies from year to year and varies among the three fuel types (gasoline, gasohol, special fuels). Several potential explanations for these discrepancies were identified, including issues on data, tax measurement, gallon measurement, HTF receipts, and timing. Data anomalies caused by outside forces, such as deferment of tax payments from one fiscal year to the next, can skew fuel tax data. Fuel tax evasion can lead to differences between actual fuel use and fuel taxes collected. Furthermore, differences in data collection and reporting among States can impact fuel use data. Refunds, credits, and transfers from the HTF can impact the total fuel tax receipt data. Timing issues, such as calendar year vs. fiscal year, can also cause some discrepancy between the two data sources.

Educational worth of physical education and sport participation: a ...

African Journals Online (AJOL)

Bailey alleged that the benefits of PESS has been made in such assertive tones that a bystander might think that nothing more can be said. Bailey and Hardman believe that it has not been proven scientifically that PESS contributes to the holistic development of the child. The present article attested the educational worth of ...

76 FR 5307 - Net Worth Standard for Accredited Investors

Science.gov (United States)

2011-01-31

... affected investors who do not fund capital calls or otherwise reinvest in future rounds of financing. \\41...-3144; IC-29572; File No. S7-04-11] RIN 3235-AK90 Net Worth Standard for Accredited Investors AGENCY... accredited investor standards in our rules under the Securities Act of 1933 to reflect the requirements of...

Remotely sensed measurements of forest structure and fuel loads in the Pinelands of New Jersey

Science.gov (United States)

Nicholas Skowronski; Kenneth Clark; Ross Nelson; John Hom; Matt Patterson

2007-01-01

We used a single-beam, first return profiling LIDAR (Light Detection and Ranging) measurements of canopy height, intensive biometric measurements in plots, and Forest Inventory and Analysis (FIA) data to quantify forest structure and ladder fuels (defined as vertical fuel continuity between the understory and canopy) in the New Jersey Pinelands. The LIDAR data were...

Characterization of fuel swelling in helium-bonded carbide fuel pins

International Nuclear Information System (INIS)

Louie, D.L.Y.

1987-08-01

This work is not only the first attempt at characterizing the swelling of (U,Pu)C fuel pellets, but it also represents the only detailed examinations on carbide fuel swelling at high fuel burnups (4 to 16 at. %). This characterization includes the contributions of fission gases, cracks and solid fission products to fuel swelling. Significantly, the contributions of fission gases and cracks were determined by using the image analysis technique (IAT) which allows researchers to take areal measurements of the irradiated fuel porosity and cracks from the photographs of metallographic fuel samples. However, because areal measurements for varying depths in the fuel pellet could not be obtained, the crack areal measurements could not be converted into volumetric quantities. Consequently, in this situation, an areal fuel swelling analysis was used. The macroscopic fission-gas induced fuel swelling (MAS) caused by fission-gas bubbles and pores > 1 μm was determined using the measured irradiated fuel porosity because the measuring range of IAT is limited to bubbles and pores >1 μm. Conversely, for fuel swelling induced by fission-gas bubbles < 1 μm, the microscopic fission-gas induced fuel swelling (MIS) was estimated using an areal fuel swelling model

Self-worth mediates the effects of violent loss on PTSD symptoms.

Science.gov (United States)

Mancini, Anthony D; Prati, Gabriele; Black, Sarah

2011-02-01

Although research has confirmed that violent losses can exacerbate grief reactions, few investigations have explored underlying mechanisms. In this study, the authors used a dataset on bereaved spouses and bereaved parents at 4- and 18-months postloss to examine the mediating effects of self-worth and worldviews (benevolence and meaningfulness beliefs). Persons bereaved by violent causes had significantly more posttraumatic stress disorder (PTSD), grief, and depression symptoms at 4- and 18-months postloss than persons bereaved by natural causes. Moreover, self-worth but not worldviews mediated the effects of violent loss on PTSD and depression symptoms cross sectionally and PTSD symptoms longitudinally. Findings underscore that self-views are a critical component of problematic reactions to violent loss, but fail to support the role of "shattered" worldviews. Copyright © 2011 International Society for Traumatic Stress Studies.

Factors which could limit the nuclear fuel cycle development

International Nuclear Information System (INIS)

Pecqueur, M.; Barre, B.

1977-01-01

The nuclear fuel cycle is a most important industry for the energy future of the world. It has also a leading part as regards the physical continuity of energy supply of the countries engaged in the nuclear field. The development of this industry is subject to the economic or political constraints involved by the availability of raw materials, technologies or production means. The various limiting factors which could affect the different stages of the fuel cycle are linked with the technical, economic and financial aspects, with the impact on the environment, nuclear safety, risks of non-pacific uses and proliferation of arms. Interesting to note is also the correlation between the fuel cycle development and the problems of energy independence and security of nuclear programs. As a conclusion, the nuclear fuel cycle industry is confronted to difficulties due to its extremely rapid growth (doubling time 5 years) which only few heavy industries have encountered for long periods. It is more over submitted to the political and safety constraints always linked with nuclear matters. The task is therefore a difficult one. But the objective is worth-while since it is a condition to the development of nuclear industry [fr

In-reactor fuel cladding external corrosion measurement process and results

International Nuclear Information System (INIS)

Thomazet, J.; Musante, Y.; Pigelet, J.

1999-01-01

Analysis of the zirconium alloy cladding behaviour calls for an on-site corrosion measurement device. In the 80's, a FISCHER probe was used and allowed oxide layer measurements to be taken along the outer generating lines of the peripheral fuel rods. In order to allow measurements on inner rods, a thin Eddy current probe called SABRE was developed by FRAMATOME. The SABRE is a blade equipped with two E.C coils is moved through the assembly rows. A spring allows the measurement coil to be clamped on each of the generating lines of the scanned rods. By inserting this blade on all four assembly faces, measurements can also be performed along several generating lines of the same rod. Standard rings are fitted on the device and allow on-line calibration for each measured row. Signal acquisition and processing are performed by LAGOS, a dedicated software program developed by FRAMATOME. The measurements are generally taken at the cycle outage, in the spent fuel pool. On average, data acquisition calls for one shift per assembly (eight hours): this corresponds to more than 2500 measurement points. These measurements are processed statistically by the utility program SAN REMO. All the results are collected in a database for subsequent behaviour analysis: examples of investigated parameters are the thermal/hydraulic conditions of the reactors, the irradiation history, the cladding material, the water chemistry This analysis can be made easier by comparing the behaviour measurement and prediction by means of the COROS-2 corrosion code. (author)

An Electronic Measurement Instrumentation of the Impedance of a Loaded Fuel Cell or Battery.

Science.gov (United States)

Aglzim, El-Hassane; Rouane, Amar; El-Moznine, Reddad

2007-10-17

In this paper we present an inexpensive electronic measurement instrumentationdeveloped in our laboratory, to measure and plot the impedance of a loaded fuel cell orbattery. Impedance measurements were taken by using the load modulation method. Thisinstrumentation has been developed around a VXI system stand which controls electroniccards. Software under Hpvee ® was developed for automatic measurements and the layout ofthe impedance of the fuel cell on load. The measurement environment, like the ambienttemperature, the fuel cell temperature, the level of the hydrogen, etc..., were taken withseveral sensors that enable us to control the measurement. To filter the noise and theinfluence of the 50Hz, we have implemented a synchronous detection which filters in a verynarrow way around the useful signal. The theoretical result obtained by a simulation underPspice ® of the method used consolidates the choice of this method and the possibility ofobtaining correct and exploitable results. The experimental results are preliminary results ona 12V vehicle battery, having an inrush current of 330A and a capacity of 40Ah (impedancemeasurements on a fuel cell are in progress, and will be the subject of a forthcoming paper).The results were plotted at various nominal voltages of the battery (12.7V, 10V, 8V and 5V)and with two imposed currents (0.6A and 4A). The Nyquist diagram resulting from theexperimental data enable us to show an influence of the load of the battery on its internalimpedance. The similitude in the graph form and in order of magnitude of the valuesobtained (both theoretical and practical) enables us to validate our electronic measurementinstrumentation. One of the future uses for this instrumentation is to integrate it with several control sensors, on a vehicle as an embedded system to monitor the degradation of fuel cell membranes.

An Electronic Measurement Instrumentation of the Impedance of a Loaded Fuel Cell or Battery

Directory of Open Access Journals (Sweden)

Reddad El-Moznine

2007-10-01

Full Text Available In this paper we present an inexpensive electronic measurement instrumentationdeveloped in our laboratory, to measure and plot the impedance of a loaded fuel cell orbattery. Impedance measurements were taken by using the load modulation method. Thisinstrumentation has been developed around a VXI system stand which controls electroniccards. Software under Hpvee® was developed for automatic measurements and the layout ofthe impedance of the fuel cell on load. The measurement environment, like the ambienttemperature, the fuel cell temperature, the level of the hydrogen, etc..., were taken withseveral sensors that enable us to control the measurement. To filter the noise and theinfluence of the 50Hz, we have implemented a synchronous detection which filters in a verynarrow way around the useful signal. The theoretical result obtained by a simulation underPspice® of the method used consolidates the choice of this method and the possibility ofobtaining correct and exploitable results. The experimental results are preliminary results ona 12V vehicle battery, having an inrush current of 330A and a capacity of 40Ah (impedancemeasurements on a fuel cell are in progress, and will be the subject of a forthcoming paper.The results were plotted at various nominal voltages of the battery (12.7V, 10V, 8V and 5Vand with two imposed currents (0.6A and 4A. The Nyquist diagram resulting from theexperimental data enable us to show an influence of the load of the battery on its internalimpedance. The similitude in the graph form and in order of magnitude of the valuesobtained (both theoretical and practical enables us to validate our electronic measurementinstrumentation. One of the future uses for this instrumentation is to integrate it with several control sensors, on a vehicle as an embedded system to monitor the degradation of fuel cell membranes.