Measurement and analysis of CEFR safety and shim rod worth

International Nuclear Information System (INIS)

Chen Yiyu; Yang Yong; Gang Zhi; Xu Li; Yang Xiaoyan; Zhou Keyuan; Hu Dingsheng

2013-01-01

The reactivity worth of safety rods and shim rods in critical phase and operating phase was calculated respectively using Monte Carlo program in this paper. In addition, the reactivity worth of safety rods and shim rods was measured by the rod drop-off method and period method. The experimental results are in good agreement with the calculated values with less than 5% error. It illustrates the high calculation precision of Monte Carlo program, which provides a practical reference for subsequent application of Monte Carlo program in future demonstration fast reactors. (authors)

Reactivity-worth estimates of the OSMOSE samples in the MINERVE reactor R1-MOX, R2-UO2 and MORGANE/R configurations.

Energy Technology Data Exchange (ETDEWEB)

Zhong, Z.; Klann, R. T.; Nuclear Engineering Division

2007-08-03

An initial series of calculations of the reactivity-worth of the OSMOSE samples in the MINERVE reactor with the R2-UO2 and MORGANE/R core configuration were completed. The calculation model was generated using the lattice physics code DRAGON. In addition, an initial comparison of calculated values to experimental measurements was performed based on preliminary results for the R1-MOX configuration.

An approach to estimate the reactivity worth of R-5 poison tube system and experimental verification in ZERLINA reactor

International Nuclear Information System (INIS)

Khosla, S.K.; Paul, O.P.K.; Sengupta, S.N.

1976-01-01

It is proposed to employ a liquid poison injection system as an emergency shut down device for R-5 reactor. The liquid poison consists of gadolinium nitrate solution, which is injected into twenty poison tubes made of zircaloy that are located in between the regular lattice positions in R-5 reactor. The calculational model adopted to estimate the reactivity worth of the poison tubes so as to hold the reactor subcritical by 50 mk at full tank, is described. Similar reactivity estimates have also been carried out for R-5 poison tubes installed in Zerlina reactor in order to assess the adequacy of the calculational mode. The results of the calculations are compared with experimental values for single poison tubes. (author)

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

Development of a digital reactivity meter for criticality prediction and control rod worth evaluation in pressurized water reactors

Energy Technology Data Exchange (ETDEWEB)

Kuramoto, Renato Y.R.; Miranda, Anselmo F.; Valladares, Gastao Lommez; Prado, Adelk C. [Eletrobras Termonuclear S.A. - ELETRONUCLEAR, Angra dos Reis, RJ (Brazil). Central Nuclear Almirante Alvaro Alberto], e-mail: kuramot@eletronuclear.gov.br

2009-07-01

In this work, we have proposed the development of a digital reactivity meter in order to monitor subcriticality continuously during criticality approach in a PWR. A subcritical reactivity meter can provide an easy prediction of the estimated critical point prior to reactor criticality, without complicated hand calculation. Moreover, in order to reduce the interval of the Physics Tests from the economical point of view, a subcritical reactivity meter can evaluate the control rod worth from direct subcriticality measurement. In other words, count rate of Source Range (SR) detector recorded during the criticality approach could be used for subcriticality evaluation or control rod worth evaluation. Basically, a digital reactivity meter is based on the inverse solution of the kinetic equations of a reactor with the external neutron source in one-point reactor model. There are some difficulties in the direct application of a digital reactivity meter to the subcriticality measurement. When the Inverse Kinetic method is applied to a sufficiently high power level or to a core without an external neutron source, the neutron source term may be neglected. When applied to a lower power level or in the sub-critical domain, however, the source effects must be taken in account. Furthermore, some treatments are needed in using the count rate of Source Range (SR) detector as input signal to the digital reactivity meter. To overcome these difficulties, we have proposed a digital reactivity meter combined with a methodology of the modified Neutron Source Multiplication (NSM) method with correction factors for subcriticality measurements in PWR. (author)

Development of a digital reactivity meter for criticality prediction and control rod worth evaluation in pressurized water reactors

International Nuclear Information System (INIS)

Kuramoto, Renato Y.R.; Miranda, Anselmo F.; Valladares, Gastao Lommez; Prado, Adelk C.

2009-01-01

In this work, we have proposed the development of a digital reactivity meter in order to monitor subcriticality continuously during criticality approach in a PWR. A subcritical reactivity meter can provide an easy prediction of the estimated critical point prior to reactor criticality, without complicated hand calculation. Moreover, in order to reduce the interval of the Physics Tests from the economical point of view, a subcritical reactivity meter can evaluate the control rod worth from direct subcriticality measurement. In other words, count rate of Source Range (SR) detector recorded during the criticality approach could be used for subcriticality evaluation or control rod worth evaluation. Basically, a digital reactivity meter is based on the inverse solution of the kinetic equations of a reactor with the external neutron source in one-point reactor model. There are some difficulties in the direct application of a digital reactivity meter to the subcriticality measurement. When the Inverse Kinetic method is applied to a sufficiently high power level or to a core without an external neutron source, the neutron source term may be neglected. When applied to a lower power level or in the sub-critical domain, however, the source effects must be taken in account. Furthermore, some treatments are needed in using the count rate of Source Range (SR) detector as input signal to the digital reactivity meter. To overcome these difficulties, we have proposed a digital reactivity meter combined with a methodology of the modified Neutron Source Multiplication (NSM) method with correction factors for subcriticality measurements in PWR. (author)

Evaluation of fission product worth margins in PWR spent nuclear fuel burnup credit calculations

International Nuclear Information System (INIS)

Blomquist, R.N.; Finck, P.J.; Jammes, C.; Stenberg, C.G.

1999-01-01

Current criticality safety calculations for the transportation of irradiated LWR fuel make the very conservative assumption that the fuel is fresh. This results in a very substantial overprediction of the actual k eff of the transportation casks; in certain cases, this decreases the amount of spent fuel which can be loaded in a cask, and increases the cost of transporting the spent fuel to the repository. Accounting for the change of reactivity due to fuel depletion is usually referred to as ''burnup credit.'' The US DOE is currently funding a program aimed at establishing an actinide only burnup credit methodology (in this case, the calculated reactivity takes into account the buildup or depletion of a limited number of actinides). This work is undergoing NRC review. While this methodology is being validated on a significant experimental basis, it implicitly relies on additional margins: in particular, the absorption of neutrons by certain actinides and by all fission products is not taken into account. This provides an important additional margin and helps guarantee that the methodology is conservative provided these neglected absorption are known with reasonable accuracy. This report establishes the accuracy of fission product absorption rate calculations: (1) the analysis of European fission product worth experiments demonstrates that fission product cross-sections available in the US provide very good predictions of fission product worth; (2) this is confirmed by a direct comparison of European and US cross section evaluations; (3) accuracy of Spent Nuclear Fuel (SNF) fission product content predictions is established in a recent ORNL report where several SNF isotopic assays are analyzed; and (4) these data are then combined to establish in a conservative manner the fraction of the predicted total fission product absorption which can be guaranteed based on available experimental data

Depletion calculations of adjuster rods in Darlington

Energy Technology Data Exchange (ETDEWEB)

Arsenault, B.; Tsang, K., E-mail: benoit.arsenault@amecfw.com, E-mail: kwok.tsang@amecfw.com [AMEC Foster Wheeler, Toronto, ON (Canada)

2015-07-01

This paper describes the simulation methodology and reactivity worth calculated for aged adjuster rods in the Darlington core. ORIGEN-S IST was applied to simulate the isotope transmutation process of the stainless steel and titanium adjusters. The compositions were used in DRAGON-IST to calculate the change in incremental properties of aged adjusters. Pre-simulations of the reactivity worth of the stainless steel and titanium adjusters in Darlington were performed using RFSP-IST and the results showed that the titanium adjuster rods exhibit faster reactivity-worth drop than that of stainless steel rods. (author)

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

Analysis of reactivity worth for xenon poisoning during restart-up of reactor in iodine pit

International Nuclear Information System (INIS)

Li Xaofeng; Chen Wenzhen; Zhu Qian; Xu Guojun

2009-01-01

The reactivity worth of xenon poisoning and the densities of 135 I and 135 Xe were derived when the reactor was restarted up in iodine pit. Through the expressions obtained we can find the physics characteristics of reactor restarted up in iodine pit comprehensively and essentially. The results were analyzed and discussed. The reactor power before shutdown, the start-up power, the position where the reactor starts up in iodine pit, and so on, all have effect on the reactivity worth of xenon poisoning, and the different conditions can lead to totally different physics characteristics. In addition, the time when the reactor starts up in iodine pit is a very important factor for nuclear reactors safety. The conclusions are very important to the maneuverability and operation safety of ship nuclear reactors. (authors)

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.

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

Calculations of steady-state and reactivity insertion transients in a research reactor simulating the PWR

International Nuclear Information System (INIS)

Mladin, Mirea; Mladin, Daniela; Prodea, Ilie

2010-01-01

In 2008, IAEA started a Coordinated Research Project for benchmarking the thermalhydraulic and neutronic computer codes for research reactor analysis against the experimental data. In this framework, for the first year of research contract, the Institute for Nuclear Research engaged in steady-state analysis of SPERT-III reactor and also in the simulation of the reactivity insertion tests performed in this reactor during mid sixties. In the first part, the paper describes a Monte Carlo input model of the oxide core selected for investigation and the results of the steady-state neutronic calculations with respect to hot and cold core reactivity excess and control rods worth. Also, prompt neutron life and reactivity feed-back coefficients were examined. These results were compared with the data provided in the reactor specification document concerning neutronic design calculated data. The second part of the paper is dedicated to calculation of the reactivity insertion transients with RELAP5 and CATHARE2 thermalhydraulic codes, both including point reactor kinetics models, and to comparison with experimental data. (authors)

Evaluation of differential shim rod worth measurements in the OAK Ridge research reactor

International Nuclear Information System (INIS)

Bretscher, M.M.

1987-01-01

Reasonable agreement between calculated and measured differential shim rod worths in the Oak Ridge Research Reactor (ORR) has been achieved by taking into account the combined effects of negative reactivity contributions from changing fuel-moderator temperatures and of delayed photo-neutrons. A method has been developed for extracting the asymptotic period from the shape of the initial portion of the measured time-dependent neutron flux profile following a positive reactivity insertion. In this region of the curve temperature related reactivity feedback effects are negligibly small. Results obtained by applying this technique to differential shim rod worth measurements made in a wide variety of ORR cores are presented. (Author)

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

Reactivity control system of a passively safe thorium breeder pebble bed reactor

International Nuclear Information System (INIS)

Wols, F.J.; Kloosterman, J.L.; Lathouwers, D.; Hagen, T.H.J.J. van der

2014-01-01

Highlights: • A worth of over 15,000 pcm ensures achieving long-term cold shutdown in thorium PBR. • Control rod worth in side reflector is insufficient due to low-power breeder zone. • 20 control rods, just outside the driver zone, can achieve long-term cold shutdown. • BF 3 gas can be inserted for reactor shutdown, but only in case of emergency. • Perturbation theory accurately predicts absorber gas worth for many concentrations. - Abstract: This work investigates the neutronic design of the reactivity control system for a 100 MW th passively safe thorium breeder pebble bed reactor (PBR), a conceptual design introduced previously by the authors. The thorium PBR consists of a central driver zone of 100 cm radius, surrounded by a breeder zone with 300 cm outer radius. The fissile content of the breeder zone is low, leading to low fluxes in the radial reflector region. Therefore, a significant decrease of the control rod worth at this position is anticipated. The reactivity worth of control rods in the side reflector and at alternative in-core positions is calculated using different techniques, being 2D neutron diffusion, perturbation theory and more accurate 3D Monte Carlo models. Sensitivity coefficients from perturbation theory provide a first indication of effective control rod positions, while the 2D diffusion models provide an upper limit on the reactivity worth achievable at a certain radial position due to the homogeneous spreading of the absorber material over the azimuthal domain. Three dimensional forward calculations, e.g. in KENO, are needed for an accurate calculation of the total control rod worth. The two dimensional homogeneous calculations indicate that the reactivity worth in the radial reflector is by far insufficient to achieve cold reactor shutdown, which requires a control rod worth of over 15 000 pcm. Three dimensional heterogeneous KENO calculations show that placing 20 control rods just outside the driver channel, between 100 cm

Reactivity control system of a passively safe thorium breeder pebble bed reactor

Energy Technology Data Exchange (ETDEWEB)

Wols, F.J., E-mail: f.j.wols@tudelft.nl; Kloosterman, J.L.; Lathouwers, D.; Hagen, T.H.J.J. van der

2014-12-15

Highlights: • A worth of over 15,000 pcm ensures achieving long-term cold shutdown in thorium PBR. • Control rod worth in side reflector is insufficient due to low-power breeder zone. • 20 control rods, just outside the driver zone, can achieve long-term cold shutdown. • BF{sub 3} gas can be inserted for reactor shutdown, but only in case of emergency. • Perturbation theory accurately predicts absorber gas worth for many concentrations. - Abstract: This work investigates the neutronic design of the reactivity control system for a 100 MW{sub th} passively safe thorium breeder pebble bed reactor (PBR), a conceptual design introduced previously by the authors. The thorium PBR consists of a central driver zone of 100 cm radius, surrounded by a breeder zone with 300 cm outer radius. The fissile content of the breeder zone is low, leading to low fluxes in the radial reflector region. Therefore, a significant decrease of the control rod worth at this position is anticipated. The reactivity worth of control rods in the side reflector and at alternative in-core positions is calculated using different techniques, being 2D neutron diffusion, perturbation theory and more accurate 3D Monte Carlo models. Sensitivity coefficients from perturbation theory provide a first indication of effective control rod positions, while the 2D diffusion models provide an upper limit on the reactivity worth achievable at a certain radial position due to the homogeneous spreading of the absorber material over the azimuthal domain. Three dimensional forward calculations, e.g. in KENO, are needed for an accurate calculation of the total control rod worth. The two dimensional homogeneous calculations indicate that the reactivity worth in the radial reflector is by far insufficient to achieve cold reactor shutdown, which requires a control rod worth of over 15 000 pcm. Three dimensional heterogeneous KENO calculations show that placing 20 control rods just outside the driver channel

Reactivity anomalies in the FFTF [Fast Flux Test Facility

International Nuclear Information System (INIS)

Knutson, B.J.; Harris, R.A.

1987-04-01

Experience using an automated core reactivity monitoring technique at the Fast Flux Test Facility (FFTF) through eight operating cycles is described. This technique relies on comparing predicted to measured rod positions to detect any anomalous (or unpredicted) core reactivity changes. Reactivity worth predictions of core state changes (e.g., temperature and irradiation changes) and compensating control rod movements are required for the rod position comparison. A substantial data base now exists to evaluate changes in temperature reactivity feedback effects operational in the FFTF, rod worth changes due to core loading, temperature and irradiation effects and burnup effects associated with transmutation of fuel materials. This report summarizes preliminary work of correlating zero power and at-power rod worth measurement data, calculated burnup rates and rod worths using the latest ENDF/B-V cross section set for each cycle to evaluate the prediction models and attempt to resolve observed reactivity anomalies. 2 figs., 2 tabs

Whole core calculations of power reactors by Monte Carlo method

International Nuclear Information System (INIS)

Nakagawa, Masayuki; Mori, Takamasa

1993-01-01

Whole core calculations have been performed for a commercial size PWR and a prototype LMFBR by using vectorized Monte Carlo codes. Geometries of cores were precisely represented in a pin by pin model. The calculated parameters were k eff , control rod worth, power distribution and so on. Both multigroup and continuous energy models were used and the accuracy of multigroup approximation was evaluated through the comparison of both results. One million neutron histories were tracked to considerably reduce variances. It was demonstrated that the high speed vectorized codes could calculate k eff , assembly power and some reactivity worths within practical computation time. For pin power and small reactivity worth calculations, the order of 10 million histories would be necessary. Required number of histories to achieve target design accuracy were estimated for those neutronic parameters. (orig.)

Study of graphite reactivity worth on well-defined cores assembled on LR-0 reactor

International Nuclear Information System (INIS)

Košťál, Michal; Rypar, Vojtěch; Milčák, Ján; Juříček, Vlastimil; Losa, Evžen; Forget, Benoit; Harper, Sterling

2016-01-01

Highlights: • A light water critical facility for graphite reactivity worth measurements. • Comparison of calculated and measured k eff . • Effect of graphite description on k eff . - Abstract: Graphite is an often-used moderating material on the basis of its good moderating power and very low absorption cross section. This small absorption cross section permits the use of natural or low-enriched uranium in graphite moderated reactors. Graphite is now being considered as the moderator for Fluoride-salt-cooled High Temperature Reactors (FHR). The critical moderator level was measured for various graphite block configurations in an experimental dry assembly of the LR-0 reactor. Comparisons with experiments were performed between Monte Carlo simulation tools for which satisfactory agreement was obtained with the exception of some systematic discrepancies. The larger discrepancies were observed when using the ENDF/B-VII.0 library. To decrease the uncertainties, based on conservative assumptions, relative comparisons were done. The results provided by the different nuclear data libraries are within 3 sigma interval of experimental uncertainties. It has been determined that differences between the results of calculations are caused by variations in the (n,n), (n,n′), (n,g) reactions and also by various angular distributions, while the (n,g) cross section variations play only a minor role for these configurations.

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)

Analysis of void reactivity measurements in full MOX BWR physics experiments

International Nuclear Information System (INIS)

Ando, Yoshihira; Yamamoto, Toru; Umano, Takuya

2008-01-01

In the full MOX BWR physics experiments, FUBILA, four 9x9 test assemblies simulating BWR full MOX assemblies were located in the center of the core. Changing the in-channel moderator condition of the four assemblies from 0% void to 40% and 70% void mock-up, void reactivity was measured using Amplified Source Method (ASM) technique in the subcritical cores, in which three fission chambers were located. ASM correction factors necessary to express the consistency of the detector efficiency between measured core configurations were calculated using collision probability cell calculation and 3D-transport core calculation with the nuclear data library, JENDL-3.3. Measured reactivity worth with ASM correction factor was compared with the calculated results obtained through a diffusion, transport and continuous energy Monte Carlo calculation respectively. It was confirmed that the measured void reactivity worth was reproduced well by calculations. (author)

Core calculational techniques and procedures

International Nuclear Information System (INIS)

Romano, J.J.

1977-10-01

Described are the procedures and techniques employed by B and W in core design analyses of power peaking, control rod worths, and reactivity coefficients. Major emphasis has been placed on current calculational tools and the most frequently performed calculations over the operating power range

Investigation of reactivity changes due to flooding the irradiation sites of the MNSR reactor using the MCNP code and comparison with experimental results

Directory of Open Access Journals (Sweden)

A Shirani

2010-06-01

Full Text Available In this work, the Isfahan Miniature Neutron Source Reactor (MNSR has been simulated using the MCNP code, and reactivity worth of flooding the inner irradiation sites of this reactor in an accident has been calculated. Also, by inserting polyethylene capsules containing water inside the inner irradiation sites, reactivity changes of this reactor in same such accident have been measured, the results of which are in good agreements with the calculated results. In this work, the reactivity worth due to flooding one inner irradiation site is 0.53mk , and reactivity worth due to flooding of the whole 5 inner irradiation sites is 2.61 mk.

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)

Establishment of analysis procedure for control rod reactivity worth

Energy Technology Data Exchange (ETDEWEB)

Song, Hoon; Kim, Young Il; Kim, Sang Ji; Kim, Young In

2001-03-01

As to the calculation method of control rod reactivity relating to hexagonal assembly, which are used generally in fast reactor, we have investigated the calculation method, the problems to rise during calculation, the degrees of calculation and the enhancement of calculation modeling so on, and estimated the application of calculation method through comparison and analysis of calculation result using the effective cross section generation system, TRANSX/TWODANT, and neutron flux calculation system, diffusion theory code DIF-3D, which are belonged to K-CORE System, and determined the basic calculation method, and extracted the present calculation problem in case of application in K-CORE System and the future improvement items so on.

Establishment of analysis procedure for control rod reactivity worth

International Nuclear Information System (INIS)

Song, Hoon; Kim, Young Il; Kim, Sang Ji; Kim, Young In

2001-03-01

As to the calculation method of control rod reactivity relating to hexagonal assembly, which are used generally in fast reactor, we have investigated the calculation method, the problems to rise during calculation, the degrees of calculation and the enhancement of calculation modeling so on, and estimated the application of calculation method through comparison and analysis of calculation result using the effective cross section generation system, TRANSX/TWODANT, and neutron flux calculation system, diffusion theory code DIF-3D, which are belonged to K-CORE System, and determined the basic calculation method, and extracted the present calculation problem in case of application in K-CORE System and the future improvement items so on

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)

Physics calculations for the Clinch River Breeder Reactor

International Nuclear Information System (INIS)

Kalimullah; Kier, P.H.; Hummel, H.H.

1977-06-01

Calculations of distributions of power and sodium void reactivity, unvoided and voided Doppler coefficients and steel and fuel worths have been performed using diffusion theory and first-order perturbation theory for the LWR discharge Pu-fueled CRBR at BOL, the FFTF-grade Pu-fueled CRBR at BOL and for the beginning and end of equilibrium cycle of the LWR-Pu-fueled CRBR. The results of the burnup and breeding ratio calculations performed for obtaining the reactor compositions during the equilibrium cycle are also reported. Effects of sodium and steel contents on the distributions of sodium void reactivity and steel worth have also been studied. Errors and uncertainties in the reactivity coefficients due to cross-sections and the two-dimensional geometric representations of the reactor used in the calculations have also been estimated. Comparisons of the results with those in the CRBR PSAR are also discussed

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)

Development of external coupling for calculation of the control rod worth in terms of burn-up for a WWER-1000 nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Noori-Kalkhoran, Omid, E-mail: o_noori@yahoo.com [Reactor Research School, Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of); Yarizadeh-Beneh, Mehdi [Faculty of Engineering, Shahid Beheshti University, Tehran (Iran, Islamic Republic of); Ahangari, Rohollah [Reactor Research School, Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of)

2016-08-15

Highlights: • Calculation of control rod worth in term of burn-up. • Calculation of differential and integral control rod worth. • Developing an external couple. • Modification of thermal-hydraulic profiles in calculations. - Abstract: One of the main problems relating to operation of a nuclear reactor is its safety and controlling system. The most widely used control systems for thermal reactors are neutron absorbent rods. In this study a code based method has been developed for calculation of integral and differential control rod worth in terms of burn-up for a WWER-1000 nuclear reactor. External coupling of WIMSD-5B, PARCS V2.7 and COBRA-EN has been used for this purpose. WIMSD-5B has been used for cell calculation and handling burn-up of the core in various days. PARCS V2.7 has been used for neutronic calculation of core and critical boron concentration search. Thermal-hydraulic calculation has been performed by COBRA-EN. An external coupling algorithm has been developed by MATLAB to couple and transfer suitable data between these codes in each step. Steady-State Power Picking Factors (PPFs) of the core and control rod worth for different control rod groups have been calculated from Beginning Of Cycle (BOC) to 289.7 Effective Full Power Days (EFPDs) in some steps. Results have been compared with the results of Bushehr Nuclear Power Plant (BNPP) Final Safety Analysis Report (FSAR). The results show a good agreement and confirm the ability of developed coupling in calculation of control rod worth in terms of burn-up.

Reactor perturbation calculations by Monte Carlo methods

International Nuclear Information System (INIS)

Gubbins, M.E.

1965-09-01

Whilst Monte Carlo methods are useful for reactor calculations involving complicated geometry, it is difficult to apply them to the calculation of perturbation worths because of the large amount of computing time needed to obtain good accuracy. Various ways of overcoming these difficulties are investigated in this report, with the problem of estimating absorbing control rod worths particularly in mind. As a basis for discussion a method of carrying out multigroup reactor calculations by Monte Carlo methods is described. Two methods of estimating a perturbation worth directly, without differencing two quantities of like magnitude, are examined closely but are passed over in favour of a third method based on a correlation technique. This correlation method is described, and demonstrated by a limited range of calculations for absorbing control rods in a fast reactor. In these calculations control rod worths of between 1% and 7% in reactivity are estimated to an accuracy better than 10% (3 standard errors) in about one hour's computing time on the English Electric KDF.9 digital computer. (author)

Reactivity feedback evaluation of material relocations in the CABRI-1 experiments with fuel worth distributions from SNR-300

International Nuclear Information System (INIS)

Royl, P.; Pfrang, W.; Struwe, D.

1991-01-01

The fuel relocations from the CABRI-1 experiments with irradiated fuel that had been evaluated from the hodoscope measurements were used together with fuel reactivity worth distributions from the SNR-300 to estimate the reactivity effect which these motions would have if they occurred in SNR-300 at the same relative distance to the peak power as in CABRI. The procedure for the reactivity evaluation is outlined including the assumptions made for fuel mass conservation. The results show that the initial fuel motion yields always negative reactivities. They also document the mechanism for a temporary reactivity increase by in-pin fuel flow in some transient overpower tests. This mechanism, however, never dominates, because material accumulates always sufficiently above the peak power point. Thus, the late autocatalytic amplifications of voiding induced power excursions by compactive in-pin fuel flow, that had been simulated in bounding loss of flow analyses for SNR-300, have no basis at all when considering the results from the CABRI-1 experiments

Core physics calculation and analysis for SNRE

International Nuclear Information System (INIS)

Xie Jiachun; Zhao Shouzhi; Jia Baoshan

2010-01-01

Five different precise calculation models have been set up for Small Nuclear Rocket Engine (SNRE) core based on MCNP code, and then the effective multiplication constant, drum control worth and power distribution were calculated. The results from different models indicate that the model in which elements are homogeneous could be used in the reactivity calculation, but a detailed description of elements have to be used in the element internal power distribution calculation. The results of physics parameters show that the basic characteristics of SNRE are reasonable. The drum control worth is sufficient. The power distribution is symmetrical and reasonable. All of the parameters can satisfy the design requirement. (authors)

Positive void reactivity

International Nuclear Information System (INIS)

Diamond, D.J.

1992-09-01

This report is a review of some of the important aspects of the analysis of large loss-of-coolant accidents (LOCAs). One important aspect is the calculation of positive void reactivity. To study this subject the lattice physics codes used for void worth calculations and the coupled neutronic and thermal-hydraulic codes used for the transient analysis are reviewed. Also reviewed are the measurements used to help validate the codes. The application of these codes to large LOCAs is studied with attention focused on the uncertainty factor for the void worth used to bias the results. Another aspect of the subject dealt with in the report is the acceptance criteria that are applied. This includes the criterion for peak fuel enthalpy and the question of whether prompt criticality should also be a criterion. To study the former, fuel behavior measurements and calculations are reviewed. (Author) (49 refs., 2 figs., tab.)

Multi-group diffusion perturbation calculation code. PERKY (2002)

Energy Technology Data Exchange (ETDEWEB)

Iijima, Susumu; Okajima, Shigeaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2002-12-01

Perturbation calculation code based on the diffusion theory ''PERKY'' is designed for nuclear characteristic analyses of fast reactor. The code calculates reactivity worth on the multi-group diffusion perturbation theory in two or three dimensional core model and kinetics parameters such as effective delayed neutron fraction, prompt neutron lifetime and absolute reactivity scale factor ({rho}{sub 0} {delta}k/k) for FCA experiments. (author)

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)

Hamming generalized corrector for reactivity calculation

International Nuclear Information System (INIS)

Suescun-Diaz, Daniel; Ibarguen-Gonzalez, Maria C.; Figueroa-Jimenez, Jorge H.

2014-01-01

This work presents the Hamming method generalized corrector for numerically resolving the differential equation of delayed neutron precursor concentration from the point kinetics equations for reactivity calculation, without using the nuclear power history or the Laplace transform. A study was carried out of several correctors with their respective modifiers with different time step calculations, to offer stability and greater precision. Better results are obtained for some correctors than with other existing methods. Reactivity can be calculated with precision of the order h 5 , where h is the time step. (orig.)

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.

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

Calculational model based on influence function method for power distribution and control rod worth in fast reactors

International Nuclear Information System (INIS)

Sanda, T.; Azekura, K.

1983-01-01

A model for calculating the power distribution and the control rod worth in fast reactors has been developed. This model is based on the influence function method. The characteristics of the model are as follows: Influence functions for any changes in the control rod insertion ratio are expressed by using an influence function for an appropriate control rod insertion in order to reduce the computer memory size required for the method. A control rod worth is calculated on the basis of a one-group approximation in which cross sections are generated by bilinear (flux-adjoint) weighting, not the usual flux weighting, in order to reduce the collapse error. An effective neutron multiplication factor is calculated by adjoint weighting in order to reduce the effect of the error in the one-group flux distribution. The results obtained in numerical examinations of a prototype fast reactor indicate that this method is suitable for on-line core performance evaluation because of a short computing time and a small memory size

Specification of phase 3 benchmark (Hex-Z heterogeneous and burnup calculation)

International Nuclear Information System (INIS)

Kim, Y.I.

2002-01-01

During the second RCM of the IAEA Co-ordinated Research Project Updated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effects the following items were identified as important. Heterogeneity will affect absolute core reactivity. Rod worths could be considerably reduced by heterogeneity effects depending on their detailed design. Heterogeneity effects will affect the resonance self-shielding in the treatment of fuel Doppler, steel Doppler and sodium density effects. However, it was considered more important to concentrate on the sodium density effect in order to reduce the calculational effort required. It was also recognized that burnup effects will have an influence on fuel Doppler and sodium worths. A benchmark for the assessment of heterogeneity effect for Phase 3 was defined. It is to be performed for the Hex-Z model of the reactor only. No calculations will be performed for the R-Z model. For comparison with heterogeneous evaluations, the control rod worth will be calculated at the beginning of the equilibrium cycle, based on the homogeneous model. The definitions of rod raised and rod inserted for SHR are given, using the composition numbers

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

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)

Development of a parallel processing couple for calculations of control rod worth in terms of burn-up in a WWER-1000 reactor

Energy Technology Data Exchange (ETDEWEB)

Noori-Kalkhoran, Omid; Ahangari, R. [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of). Reactor Research school; Shirani, A.S. [Shahid Beheshti Univ., Tehran (Iran, Islamic Republic of). Faculty of Engineering

2017-03-15

In this study a code based method has been developed for calculation of integral and differential control rod worth in terms of burn-up for a WWER-1000 reactor. Parallel processing of WIMSD-5B, PARCS V2.7 and COBRA-EN has been used for this purpose. WIMSD-5B has been used for cell calculation and handling burn-up of core at different days. PARCS V2.7?has been used for neutronic calculation of core and critical boron concentration search. Thermal-hydraulic calculation has been performed by COBRA-EN. A Parallel processing algorithm has been developed by MATLAB to couple and transfer suitable data between these codes in each step. Steady-State Power Picking Factors (PPFs) of the core and Control rod worth have been calculated from Beginning Of Cycle (BOC) to 289.7 Effective full Power Days (EFPDs) in some steps. Results have been compared with Bushehr Nuclear Power Plant (BNPP) Final Safety Analysis Report (FSAR) results. The results show great similarity and confirm the ability of developed coupling in calculation of control rod worth in terms of burn-up.

Reactivity calculation with reduction of the nuclear power fluctuations

International Nuclear Information System (INIS)

Suescun Diaz, Daniel; Senra Martinez, Aquilino

2009-01-01

A new formulation is presented in this paper for the calculation of reactivity, which is simpler than the formulation that uses the Laplace and Z transforms. A treatment is also made to reduce the intensity of the noise found in the nuclear power signal used in the calculation of reactivity. Two classes of different filters are used for that. This treatment is based on the fact that the reactivity can be written by using the compose Simpson's rule resulting in a sum of two convolution terms with response to the impulse that is characteristic of a linear system. The linear part is calculated by using the filter named finite impulse response filter (FIR). The non-linear part is calculated using the filter exponentially adjusted by the least squares method, which does not cause attenuation in the reactivity calculation.

Reactivity calculation with reduction of the nuclear power fluctuations

Energy Technology Data Exchange (ETDEWEB)

Suescun Diaz, Daniel [COPPE/UFRJ, Programa de Engenharia Nuclear, Caixa Postal 68509, CEP 21941-914 RJ (Brazil)], E-mail: dsuescun@hotmail.com; Senra Martinez, Aquilino [COPPE/UFRJ, Programa de Engenharia Nuclear, Caixa Postal 68509, CEP 21941-914 RJ (Brazil)

2009-05-15

A new formulation is presented in this paper for the calculation of reactivity, which is simpler than the formulation that uses the Laplace and Z transforms. A treatment is also made to reduce the intensity of the noise found in the nuclear power signal used in the calculation of reactivity. Two classes of different filters are used for that. This treatment is based on the fact that the reactivity can be written by using the compose Simpson's rule resulting in a sum of two convolution terms with response to the impulse that is characteristic of a linear system. The linear part is calculated by using the filter named finite impulse response filter (FIR). The non-linear part is calculated using the filter exponentially adjusted by the least squares method, which does not cause attenuation in the reactivity calculation.

SCRAM reactivity calculations with the KIKO3D code

International Nuclear Information System (INIS)

Hordosy, G.; Kerszturi, A.; Maraczy, Cs.; Temesvari, E.

1999-01-01

Discrepancies between calculated static reactivities and measured reactivities evaluated with reactivity meters led to investigating SCRAM with the KIKO3D dynamic code, The time and space dependent neutron flux in the reactor core during the rod drop measurement was calculated by the KIKO3D nodal diffusion code. For calculating the ionisation chamber signals the Green function technique was applied. The Green functions of ionisation chambers were evaluated via solving the neutron transport equation in the reflector regions with the MCNP Monte Carlo code. The detector signals during asymmetric SCRAM measurements were calculated and compared with measured data using the inverse point kinetics transformation. The sufficient agreement validates the KIKO3D code to determine the reactivities after SCRAM. (Authors)

Calculational model based on influence function method for power distribution and control rod worth in fast reactors

International Nuclear Information System (INIS)

Toshio, S.; Kazuo, A.

1983-01-01

A model for calculating the power distribution and the control rod worth in fast reactors has been developed. This model is based on the influence function method. The characteristics of the model are as follows: 1. Influence functions for any changes in the control rod insertion ratio are expressed by using an influence function for an appropriate control rod insertion in order to reduce the computer memory size required for the method. 2. A control rod worth is calculated on the basis of a one-group approximation in which cross sections are generated by bilinear (flux-adjoint) weighting, not the usual flux weighting, in order to reduce the collapse error. 3. An effective neutron multiplication factor is calculated by adjoint weighting in order to reduce the effect of the error in the one-group flux distribution. The results obtained in numerical examinations of a prototype fast reactor indicate that this method is suitable for on-line core performance evaluation because of a short computing time and a small memory size

Reactivity balance for a soluble boron-free small modular reactor

Directory of Open Access Journals (Sweden)

Lezani van der Merwe

2018-06-01

Full Text Available Elimination of soluble boron from reactor design eliminates boron-induced reactivity accidents and leads to a more negative moderator temperature coefficient. However, a large negative moderator temperature coefficient can lead to large reactivity feedback that could allow the reactor to return to power when it cools down from hot full power to cold zero power. In soluble boron-free small modular reactor (SMR design, only control rods are available to control such rapid core transient.The purpose of this study is to investigate whether an SMR would have enough control rod worth to compensate for large reactivity feedback. The investigation begins with classification of reactivity and completes an analysis of the reactivity balance in each reactor state for the SMR model.The control rod worth requirement obtained from the reactivity balance is a minimum control rod worth to maintain the reactor critical during the whole cycle. The minimum available rod worth must be larger than the control rod worth requirement to manipulate the reactor safely in each reactor state. It is found that the SMR does have enough control rod worth available during rapid transient to maintain the SMR at subcritical below k-effectives of 0.99 for both hot zero power and cold zero power. Keywords: Control Rod Worth, Reactivity Balance, Reactivity Feedback, Small Modular Reactor, Soluble Boron Free

Moderator temperature effects on reactivity of HEU core of MNSR

International Nuclear Information System (INIS)

Ahmad, Siraj-ul-Islam; Sahibzada, Tasveer Muhammad

2012-01-01

Highlights: ► The MNSR core was analyzed to see the cross section effects on moderator temperature coefficient of reactivity. ► WIMS-D code was used for cell calculations. ► The 3D diffusion theory code PRIDE was first validated using IAEA benchmark problem and then used for analysis of MNSR. ► The differences among results for various libraries were discussed. -- Abstract: In this article we report on analyses that were performed to investigate the influence of cross section differences among libraries released by various centers on reactivity of Miniature Neutron Source Reactors. The 3D model of the core was developed with WIMS-D and PRIDE codes and six cross section libraries were used including JENDL-3.2, JEF-2.2, JEFF-3.3, ENDF/B-VI and ENDF/B-VII, and IAEA library. It was observed that all the libraries predict the reactivity within 10%, with IAEA library giving minimum reactivity worth, and JEF-2.2 data library resulted in highest worth.

Measurements and calculation of reactivity in the IEA-R1 nuclear reactor

International Nuclear Information System (INIS)

Ferreira, P.S.B.

1988-01-01

Techniques and experimentals procedures utilized in the measurement of some nuclear parameters related to reactivity are presented. Measurements of reactivity coefficients, such as void, temperature and power, and control rod worth were made in the IEA-R1 Research Reactor. The techniques used to perform the measurements were: i) stable period (control rod calibration), ii) inverse kinetics (digital reactivity meter), iii) aluminium slab insertion in the fuel element coolant channels (void reactivity), iv) nuclear reactor core temperature changes by means of the changes in the coolant systems of reactor core (isothermal reactivity coefficient) and v) by making perturbation in the core through the control rod motions (power reactivity coefficient and control rod calibration). By using the computer codes HAMMER, HAMMER-TECHNION and CITATION, the experiments realized in the IEA-R1 reactor were simulated. From this simulation, the theoretical reactivity parameters were estimated and compared with the respective experimental results. Furthermore, in the second fuel load of Angra-1 Nuclear Power Station, the IPEN-CNEN/SP digital reactivity - meter were used in the lower power test with the aim to assess the equipment performance. Among several tests, the reacticity-meter were used in parallel with a Westinghouse analogic reativimeter-meter) to measure the heat additiona point, critical boron concentration, control rod calibration, isothermal and moderator reactivity coefficient. These tests, and the results obtained by the digital reactivity-meter are described. The results were compared with those obtained by Westinghouse analogic reactivity meter, showing excellent agreement. (author) [pt

MTR fuel element burn-up measurements by the reactivity method

International Nuclear Information System (INIS)

Zuniga, A.; Cuya, T.R.; Ravnik, M.

2003-01-01

Fuel element burn-up was measured by the reactivity method in the 10 MW Peruvian MTR reactor RP-10. The main purpose of the experiment was testing the reactivity method for an MTR reactor as the reactivity method was originally developed for TRIGA reactors. The reactivity worth of each measured fuel element was measured in its original core position in order to measure the burn-up of the fuel elements that were part of the experimental core. The burn-up of each measured fuel element was derived by interpolating its reactivity worth from the reactivity worth of two reference fuel elements of known burn-up, whose reactivity worth was measured in the position of the measured fuel element. The accuracy of the method was improved by separating the reactivity effect of burn-up from the effect of the position in the core. The results of the experiment showed that the modified reactivity method for fuel element burn-up determination could be applied also to MTR reactors. (orig.)

Void coefficient of reactivity calculation for AP-600 core

International Nuclear Information System (INIS)

Suparlina, L.; Budiono, T.A.; Mardha, A.; Tukiran

1998-01-01

Void coefficient of reactivity as one of reactor kinetics parameters has been carried out. The calculation was done into two steps which is cell calculation using WIMSD/4 and core calculation using Batan-2DIFF code programs with the condition of beginning of cycle with all fresh fuels elements and all control rods withdrawn. The one dimension transport program in four neutron energy groups is used to calculate the cell generation of various core materials cell has been calculated in 1/4 fuel element with cluster model and square pitch arrange. Moderator density have been reduced until 20% for the void coefficient of reactivity calculation. Macroscopic cross-section as the out put of WIMSD/4 is being used as the input at the diffusion neutron program for core calculation. The void coefficient of reactivity of the AP-600 core can be determined with regular neutron flux and adjoint in four energy groups and X-Y geometry. The results is shown that the K eff calculation value is different 5.2% from the design data

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)

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.

BN-600 Phase III benchmark calculations

International Nuclear Information System (INIS)

Hill, R.N.; Grimm, K.N.

2002-01-01

Calculations for a Hexagonal-Z model of the BN-600 reactor with a partial mixed oxide loading, based on a joint IPPE/OBMK loading configuration that contained three uranium enrichment zones and one plutonium enrichment zone in the core, have been performed at ANL. Control-rod worths and reactivity feedback coefficients were calculated using both homogeneous and heterogeneous models. These values were calculated with either first-order perturbation theory methods (Triangle-Z geometry), nodal eigenvalue differences (Hexagonal-Z geometry), or Monte Carlo eigenvalue differences. Both spatially-dependent and region integrated values are shown

Advances in supercell calculation methods and comparison with measurements

Energy Technology Data Exchange (ETDEWEB)

Arsenault, B [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Baril, R; Hotte, G [Hydro-Quebec, Central Nucleaire Gentilly, Montreal, Quebec (Canada)

1996-07-01

In the last few years, modelling techniques have been developed in new supercell computer codes. These techniques have been used to model the CANDU reactivity devices. One technique is based on one- and two-dimensional transport calculations with the WIMS-AECL lattice code followed by super homogenization and three-dimensional flux calculations in a modified version of the MULTICELL code. The second technique is based on two- and three-dimensional transport calculations in DRAGON. The code calculates the lattice properties by solving the transport equation in a two-dimensional geometry followed by supercell calculations in three dimensions. These two calculation schemes have been used to calculate the incremental macroscopic properties of CANDU reactivity devices. The supercell size has also been modified to define incremental properties over a larger region. The results show improved agreement between the reactivity worth of zone controllers and adjusters. However, at the same time the agreement between measured and simulated flux distributions deteriorated somewhat. (author)

Formation for the calculation of reactivity without nuclear power history

International Nuclear Information System (INIS)

Suescun Diaz, Daniel; Senra Martinez, Aquilino; Carvalho Da Silva, Fernando

2007-01-01

This paper presents a new method for the solution of the inverse point kinetics equation. This method is based on the integration by parts of the integral of the inverse point kinetics equation, which results in a power series in terms of the nuclear power in time dependence. With the imposition of conditions to the nuclear power, the reactivity is represented as first and second derivatives of this nuclear power. This new calculation method for reactivity has very special characteristics, amongst which the possibility of using longer sampling period, and the possibility of restarting the calculation, after its interruption, allowing the calculation of reactivity in a non-continuous way. Beside that, the reactivity can be obtained independent of the nuclear power memory. (author)

Reactivity Coefficient Calculation for AP1000 Reactor Using the NODAL3 Code

Science.gov (United States)

Pinem, Surian; Malem Sembiring, Tagor; Tukiran; Deswandri; Sunaryo, Geni Rina

2018-02-01

The reactivity coefficient is a very important parameter for inherent safety and stability of nuclear reactors operation. To provide the safety analysis of the reactor, the calculation of changes in reactivity caused by temperature is necessary because it is related to the reactor operation. In this paper, the temperature reactivity coefficients of fuel and moderator of the AP1000 core are calculated, as well as the moderator density and boron concentration. All of these coefficients are calculated at the hot full power condition (HFP). All neutron diffusion constant as a function of temperature, water density and boron concentration were generated by the SRAC2006 code. The core calculations for determination of the reactivity coefficient parameter are done by using NODAL3 code. The calculation results show that the fuel temperature, moderator temperature and boron reactivity coefficients are in the range between -2.613 pcm/°C to -4.657pcm/°C, -1.00518 pcm/°C to 1.00649 pcm/°C and -9.11361 pcm/ppm to -8.0751 pcm/ppm, respectively. For the water density reactivity coefficients, the positive reactivity occurs at the water temperature less than 190 °C. The calculation results show that the reactivity coefficients are accurate because the results have a very good agreement with the design value.

IAEA sodium void reactivity benchmark calculations

International Nuclear Information System (INIS)

Hill, R.N.; Finck, P.J.

1992-01-01

In this paper, the IAEA-1 992 ''Benchmark Calculation of Sodium Void Reactivity Effect in Fast Reactor Core'' problem is evaluated. The proposed design is a large axially heterogeneous oxide-fueled fast reactor as described in Section 2; the core utilizes a sodium plenum above the core to enhance leakage effects. The calculation methods used in this benchmark evaluation are described in Section 3. In Section 4, the calculated core performance results for the benchmark reactor model are presented; and in Section 5, the influence of steel and interstitial sodium heterogeneity effects is estimated

Calculation of reactivity without Lagrange interpolation

International Nuclear Information System (INIS)

Suescun D, D.; Figueroa J, J. H.; Rodriguez R, K. C.; Villada P, J. P.

2015-09-01

A new method to solve numerically the inverse equation of punctual kinetics without using Lagrange interpolating polynomial is formulated; this method uses a polynomial approximation with N points based on a process of recurrence for simulating different forms of nuclear power. The results show a reliable accuracy. Furthermore, the method proposed here is suitable for real-time measurements of reactivity, with step sizes of calculations greater that Δt = 0.3 s; due to its precision can be used to implement a digital meter of reactivity in real time. (Author)

Continuous reactivity calculation for subcritical system

International Nuclear Information System (INIS)

Silva, Cristiano; Goncalves, Alessandro C.; Martinez, Aquilino S.; Silva, Fernando C. da

2011-01-01

With the rise of a new generation of nuclear reactors as for existence the ADS (Accelerator-Driven System), it is important to have a fast and accurate prediction of the variation in reactivity during a possible variation in the intensity of external sources. This paper presents a formulation for the calculation of reactivity in subcritical systems using the inverse method related only to nuclear power derivatives. One of the applications of the proposed method is the possibility of developing reactimeters that allow the continuous monitoring of subcritical systems. (author)

Continuous reactivity calculation for subcritical system

Energy Technology Data Exchange (ETDEWEB)

Silva, Cristiano; Goncalves, Alessandro C.; Martinez, Aquilino S.; Silva, Fernando C. da, E-mail: cristiano@herzeleid.net, E-mail: aquilino@lmp.ufrj.br, E-mail: fernando@con.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Palma, Daniel A.P., E-mail: dapalma@cnen.gov.br [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

2011-07-01

With the rise of a new generation of nuclear reactors as for existence the ADS (Accelerator-Driven System), it is important to have a fast and accurate prediction of the variation in reactivity during a possible variation in the intensity of external sources. This paper presents a formulation for the calculation of reactivity in subcritical systems using the inverse method related only to nuclear power derivatives. One of the applications of the proposed method is the possibility of developing reactimeters that allow the continuous monitoring of subcritical systems. (author)

Calculation of reactivity of control rods in graphite moderated reactors

International Nuclear Information System (INIS)

Nakata, H.

1978-01-01

A study about the method of calculation for the reactivity of control rods in graphite-moderated critical assemblies, is presented. The result of theoretical calculation, developed by super celles and Nordheim-Scalettar methods are compared with experimental results for the critical Assembly of General Atomic. The two methods are then applicable to reactivity calculation of the control rods of graphite moderated critical assemblies [pt

Influence of external source location in the reactivity calculation

International Nuclear Information System (INIS)

Silva, Adilson Costa da; Silva, Fernando Carvalho da; Martinez, Aquilino Senra

2011-01-01

We used the neutron diffusion equation with external neutron sources, in cartesian geometry and the two groups of energy, to verify the influence of external neutron source locations in the reactivity calculation. For this, a coarse mesh finite difference method was developed for the adjoint flux calculation and simplifies reactivity calculation in PWR type reactor, which uses the output of the nodal expansion method. The results were obtained for different locations on the two-dimensional plane, as well as for different types of fuel elements in the reactor core. (author)

Influence of external source location in the reactivity calculation

Energy Technology Data Exchange (ETDEWEB)

Silva, Adilson Costa da; Silva, Fernando Carvalho da; Martinez, Aquilino Senra, E-mail: asilva@con.ufrj.b, E-mail: fernando@con.ufrj.b, E-mail: Aquilino@lmp.ufrj.b [Coordenacao dos Programas de Pos-Graduacao de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

2011-07-01

We used the neutron diffusion equation with external neutron sources, in cartesian geometry and the two groups of energy, to verify the influence of external neutron source locations in the reactivity calculation. For this, a coarse mesh finite difference method was developed for the adjoint flux calculation and simplifies reactivity calculation in PWR type reactor, which uses the output of the nodal expansion method. The results were obtained for different locations on the two-dimensional plane, as well as for different types of fuel elements in the reactor core. (author)

ALPHA/PHOENIX-P/ANC system validation for Angra-1 neutronic calculations

International Nuclear Information System (INIS)

Ponzoni Filho, Pedro; Sato, Sadakatu; Santos, Teresinha Ipojuca Cardoso; Fernandes, Vanderlei Borba; Fetterman, R.J.

1995-01-01

The ALPHA/PHOENIX-P/ANC (APA) code package is an advanced neutronic calculation system for pressurized water reactor (PWR). PHOENIX-P generates the required cross sections for the fuel, burnable absorbers, control rods and baffle/reflector region. The ALPHA code is used to automate the generation of these cross-sections as well as process the PHOENIX-P results to generate the ANC model input. ANC is a three dimensional advanced nodal code used for the modeling of the, depletion of the fuel in the core, and for the calculation of power distributions, rod worths and other reactivity parameters. This paper provides brief overview of the APA methodology for reload core design of Angra Unit 1 Cycles 1 and 2. Results included are predicted power distributions, control rod worths and other reactivity parameters compared to plant measurements. These results demonstrate that the APA system can be used for the reload core design. (author). 7 refs, 9 figs

ALPHA/PHOENIX-P/ANC system validation for Angra-1 neutronic calculations

Energy Technology Data Exchange (ETDEWEB)

Ponzoni Filho, Pedro; Sato, Sadakatu; Santos, Teresinha Ipojuca Cardoso; Fernandes, Vanderlei Borba [FURNAS, Rio de Janeiro, RJ (Brazil); Fetterman, R.J. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

1995-12-31

The ALPHA/PHOENIX-P/ANC (APA) code package is an advanced neutronic calculation system for pressurized water reactor (PWR). PHOENIX-P generates the required cross sections for the fuel, burnable absorbers, control rods and baffle/reflector region. The ALPHA code is used to automate the generation of these cross-sections as well as process the PHOENIX-P results to generate the ANC model input. ANC is a three dimensional advanced nodal code used for the modeling of the, depletion of the fuel in the core, and for the calculation of power distributions, rod worths and other reactivity parameters. This paper provides brief overview of the APA methodology for reload core design of Angra Unit 1 Cycles 1 and 2. Results included are predicted power distributions, control rod worths and other reactivity parameters compared to plant measurements. These results demonstrate that the APA system can be used for the reload core design. (author). 7 refs, 9 figs.

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.

Integral test of JENDL-3.2 data by re-analysis of sample reactivity measurements at SEG and STEK facilities

International Nuclear Information System (INIS)

Dietze, Klaus

2001-01-01

Sample reactivity measurements, which have been performed at the fast-thermal coupled facilities RRR/SEG and STEK, have been re-analyzed using the JNC route for reactor calculation JENDL-3.2 // SLAROM / CITATION / PERKY. C/E-values of central reactivity worths (CRW) of FP nuclides, structural materials, and standards are given. (author)

Application of reactivity method to MTR fuel burn-up measurement

International Nuclear Information System (INIS)

Zuniga, A.; Ravnik, M.; Cuya, R.

2001-01-01

Fuel element burn-up has been measured for the first time by reactivity method in a MTR reactor. The measurement was performed in RP-10 reactor of Peruvian Institute for Nuclear Energy (IPEN) in Lima. It is a pool type 10MW material testing reactor using standard 20% enriched uranium plate type fuel elements. A fresh element and an element with well defined burn-up were selected as reference elements. Several elements in the core were selected for burn-up measurement. Each of them was replaced in its original position by both reference elements. Change in excess reactivity was measured using control rod calibration curve. The burn-up reactivity worth of fuel elements was plotted as a function of their calculated burnup. Corrected burn-up values of the measured fuel elements were calculated using the fitting function at experimental reactivity for all elements. Good agreement between measured and calculated burn-up values was observed indicating that the reactivity method can be successfully applied also to MTR fuel element burn-up determination.(author)

Calculation of homogenized Pickering NGS stainless steel adjuster rod neutron cross-sections using conservation of reaction rates

Energy Technology Data Exchange (ETDEWEB)

Robinson, R C [Atlantic Nuclear Services Ltd. (Canada); Tran, F [Ontario Hydro, Pickering, ON (Canada). Pickering Generating Station

1996-12-31

A homogenization methodology for calculation of reactivity device incremental cross-sections has been developed using reaction rate conservation (RRC). A heterogeneous transport calculation of flux was utilised to produce the homogenized cross-sections for a finite difference two group diffusion code. The RRC cross-sections have been shown to improve significantly the prediction of reactivity worth for stainless steel adjuster rods installed in Pickering NGS reactors. (author). 10 refs., 3 tabs., 6 figs.

Calculational and experimental experience on core management of experimental fast reactor 'JOYO'

International Nuclear Information System (INIS)

Yoshida, A.; Arii, Y.; Shono, A.; Suzuki, S.; Kinjo, K.

1992-01-01

For the core management of JOYO Mark-II, many core characteristics have been calculated with the core management code system 'MAGI', and measurements have also been carried out at each duty operation cycle. From the evaluation of these results, the characteristics of core parameters such as criticality, reactivity coefficients, and control rod worth can be predicted accurately as followings; excess reactivity: ± 0.1% Δk/k, outlet temperature of subassembly: ±10degC, fuel burn-up: ±5%, control rod worth: ±5%. As a result, we can not only get steady operation of JOYO but also perform various irradiation tests with satisfied conditions. This paper presents experience obtained until now through twenty three duty cycle operations of Mark-II core in JOYO. (author)

Uncertainty evaluatins of CASMO-3/MASTER system for PWR core neutronics calculations

International Nuclear Information System (INIS)

Song, Jae Seung; Kim, Kang Seog; Lee, Kibog; Park, Jin Ha; Zee, Sung Quun

1996-01-01

Uncertainties in core neutronic calculations of CASMO-3/MASTER, which is a KAERI developed core nuclear design code system, were evaluated via comparisons with measured data. Comparisons were performed with plant measurement data from one Westinghouse type and one ABB-CE type plant and two Korean standard type plants. The CASMO-3/MASTER capability and levels of accuracy are concluded to be sufficient for the neutronics design including safety related parameters related with reactivity, power distributions, temperature and power coefficients, inverse boron worth and control bank worth

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...

Calculation of the void reactivity of CANDU lattices using the SCALE code system

Energy Technology Data Exchange (ETDEWEB)

Valko, J. [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.; Feher, S. [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.; Hoogenboom, J.E. [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.; Slobben, J. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands)

1995-11-01

The reactivity effect of coolant voiding in CANDU-type fuel lattices has been calculated with different methods using the SCALE code system. The known positive void reactivity coefficient of the original lattice was correctly obtained. A modified fuel bundle containing dysprosium and slightly enriched uranium to eliminate the positive reactivity effect was also calculated. Owing to the increased heterogeneity of this modified fuel the one-dimensional cylindrical calculation with XSDRN proved to be inadequate. Code options allowing bundle geometry were successfully used for the calculation of the strongly space dependent flux and spectrum changes which determine the void reactivity. (orig.).

Assessment of CANDU-6 reactivity devices for DUPIC fuel

International Nuclear Information System (INIS)

Jeong, Chang Joon; Choi, Hang Bok

1998-11-01

Reactivity device characteristics for a CANDU 6 reactor loaded with DUPIC fuel have been assessed. The lattice parameters were generated by WIMS-AECL code and the core calculations were performed by RFSP code with a 3-dimensional full core model. The reactivity devices studied are the zone controller, adjusters, mechanical control absorber and shutoff rods. For the zone controller system, damping capability for spatial oscillation was investigated. For the adjusters, the restart capability was investigated. For the adjusters, the restart capability was investigated. The shin operation and power stepback calculation were also performed to confirm the compatibility of the current adjuster system. The mechanical control absorber was assessed for the function of compensating temperature reactivity feedback following a power reduction. And shutoff rods were also assessed to investigate the following a power reduction. And shutoff rods were also assessed to investigate the static reactivity worth. This study has shown that the current reactivity device system of CANDU-6 core with the DUPIC fuel. (author). 9 refs., 17 tabs., 7 figs

An approach of SFR safety study through the most penalizing sodium void reactivity - 105

International Nuclear Information System (INIS)

Tiberi, V.; Ivanov, E.; Pignet, S.

2010-01-01

Sodium void reactivity effects can affect the plant safety significantly during accidental transients. Accordingly, they have to be accurately investigated for any new sodium cooled fast reactor concept, even if a fuel with a melting point lower than the sodium boiling temperature is adopted. Thus all new reactor concepts should be compared to each - others adopting the value of the maximal possible sodium void reactivity as a discrimination parameter. However, taking into account that the sodium void worth is spatially depended, it is not evident which volume could be voided in order to obtain the maximal reactivity increase. Typically the complete active core voiding (zones initially loaded with 235 U or 239 Pu) is taken into account. This paper summarizes the extensive work carried-out in the IRSN to investigate the sodium-void reactivity spatial profiles of a fast sodium-cooled reactor core in the aim of defining a methodology to search for the area where the void contribution to the reactivity is strictly positive. Perturbation theory design approach available in the ERANOS 2.1 code has been adopted to evaluate the 'area of positive void worth'. To do that, the code has been previously validated against experimental based benchmarks (IRPhEP) and reference calculations. The evaluation of the absolute values of reactivity profiles can be improved later-on adopting more sophisticated methodologies to perform more accurate calculations of the sample with the voided area determined adopting the rough procedure described here. It has been demonstrated that even the non-reference way of ERANOS calculations could be used to provide the basis for different core concepts inter-comparison. (authors)

Comparison of Wims-Aecl / Dragon / RFSP and MCNP results with Zed-2 measurements for control device worth and reactor kinetics - 037

International Nuclear Information System (INIS)

Pencer, J.; Choy Wong, F.; Bromley, B.P.; Atfield, J.; Zeller, M.

2010-01-01

This paper summarizes comparisons between MCNP5 and WIMS-AECL / DRAGON / RFSP calculations and experimental results obtained from the Zero Energy Deuterium (ZED-2) critical facility at AECL Chalk River Laboratories. MCNP5 and WIMS-AECL / DRAGON / RFSP were used to calculate reactivity worths for two reactivity devices, a mechanical zone controller (MZC) and shut-off rod (SOR) in a lattice similar to that of the ACR-1000 R . WIMS-AECL / DRAGON / RFSP was also used to obtain kinetics parameters for a transient based on a rod drop of a ZED-2 standby absorber rod (SAR). ZED-2 experiments were performed using 43-element ACR Low Enriched Uranium (ACR-LEU) fuel bundles with H 2 O- or air-cooled fuel bundles arranged in a 24-cm pitch square lattice. Calculations with MCNP5 gave biases in device worths that were within 0.2 mk of measured values, while WIMS-AECL / DRAGON / RFSP gave values that were within 0.3 mk of measured values. Transient analyses using the CERBERUS module within RFSP yielded a total delayed neutron fraction (β) that was within 4% of the value derived by point kinetics analysis of experimental data. The corresponding delayed photo-neutron fraction (β photo-neutron ) from CERBERUS was within 5% of that derived by point kinetics. This study has helped quantify the agreement between calculation and measurement for codes that are used in the safety analysis of the ACR-1000 reactor. Results demonstrate good agreement in code predictions. (authors)

Proceedings of the NEACRP/IAEA Specialists meeting on the international comparison calculation of a large sodium-cooled fast breeder reactor at Argonne National Laboratory on February 7-9, 1978

International Nuclear Information System (INIS)

LeSage, L.G.; McKnight, R.D.; Wade, D.C.; Freese, K.E.; Collins, P.J.

1980-08-01

The results of an international comparison calculation of a large (1250 MWe) LMFBR benchmark model are presented and discussed. Eight reactor configurations were calculated. Parameters included with the comparison were: eigenvalue, k/sub infinity/, neutron balance data, breeding reaction rate ratios, reactivity worths, central control rod worth, regional sodium void reactivity, core Doppler and effective delayed neutron fraction. Ten countries participated in the comparison, and sixteen solutions were contributed. The discussion focuses on the variation in parameter values, the degree of consistency among the various parameters and solutions, and the identification of unexpected results. The results are displayed and discussed both by individual participants and by groupings of participants

Application of bias factor method with use of virtual experimental value to prediction uncertainty reduction in void reactivity worth of breeding light water reactor

International Nuclear Information System (INIS)

Kugo, Teruhiko; Mori, Takamasa; Kojima, Kensuke; Takeda, Toshikazu

2007-01-01

We have carried out the critical experiments for the MOX fueled tight lattice LWR cores using FCA facility and constructed the XXII-1 series cores. Utilizing the critical experiments carried out at FCA, we have evaluated the reduction of prediction uncertainty in the coolant void reactivity worth of the breeding LWR core based on the bias factor method with focusing on the prediction uncertainty due to cross section errors. In the present study, we have introduced a concept of a virtual experimental value into the conventional bias factor method to overcome a problem caused by the conventional bias factor method in which the prediction uncertainty increases in the case that the experimental core has the opposite reactivity worth and the consequent opposite sensitivity coefficients to the real core. To extend the applicability of the bias factor method, we have adopted an exponentiated experimental value as the virtual experimental value and formulated the prediction uncertainty reduction by the use of the bias factor method extended by the concept of the virtual experimental value. From the numerical evaluation, it has been shown that the prediction uncertainty due to cross section errors has been reduced by the use of the concept of the virtual experimental value. It is concluded that the introduction of virtual experimental value can effectively utilize experimental data and extend applicability of the bias factor method. (author)

Investigating heavy water zero power reactors with a new core configuration based on experiment and calculation results

Energy Technology Data Exchange (ETDEWEB)

Nasrazadani, Zahra; Salimi, Raana; Askari, Afrooz; Khorsandi, Jamshid; Mirvakili, Mohammad; Mashayekh, Mohammad [Reactor Research School, Nuclear Science and Technology Research Institute, Atomic Energy Organization of Iran, Esfahan (Iran, Islamic Republic of)

2017-02-15

The heavy water zero power reactor (HWZPR), which is a critical assembly with a maximum power of 100 W, can be used in different lattice pitches. The last change of core configuration was from a lattice pitch of 18-20 cm. Based on regulations, prior to the first operation of the reactor, a new core was simulated with MCNP (Monte Carlo N-Particle)-4C and WIMS (Winfrith Improved Multigroup Scheme)-CITATON codes. To investigate the criticality of this core, the effective multiplication factor (Keff) versus heavy water level, and the critical water level were calculated. Then, for safety considerations, the reactivity worth of D2O, the reactivity worth of safety and control rods, and temperature reactivity coefficients for the fuel and the moderator, were calculated. The results show that the relevant criteria in the safety analysis report were satisfied in the new core. Therefore, with the permission of the reactor safety committee, the first criticality operation was conducted, and important physical parameters were measured experimentally. The results were compared with the corresponding values in the original core.

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

Analysis of reactivity worths of highly-burnt PWR fuel samples measured in LWR-PROTEUS Phase II

Energy Technology Data Exchange (ETDEWEB)

Grimm, Peter; Murphy, Michael F.; Jatuff, Fabian; Seiler, Rudolf [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland)

2008-07-01

The reactivity loss of PWR fuel with burnup has been determined experimentally by inserting fresh and highly-burnt fuel samples in a PWR test lattice in the framework of the LWR-PROTEUS Phase II programme. Seven UO{sub 2} samples irradiated in a Swiss PWR plant with burnups ranging from approx40 to approx120 MWd/kg and four MOX samples with burnups up to approx70 MWd/kg were oscillated in a test region constituted of actual PWR UO{sub 2} fuel rods in the centre of the PROTEUS zero-power experimental facility. The measurements were analyzed using the CASMO-4E fuel assembly code and a cross section library based on the ENDF/B-VI evaluation. The results show close proximity between calculated and measured reactivity effects and no trend for a deterioration of the quality of the prediction at high burnup. The analysis thus demonstrates the high accuracy of the calculation of the reactivity of highly-burnt fuel. (authors)

Calculation of reactivity using a finite impulse response filter

Energy Technology Data Exchange (ETDEWEB)

Suescun Diaz, Daniel [COPPE/UFRJ, Programa de Engenharia Nuclear, Caixa Postal 68509, CEP 21941-914, RJ (Brazil); Senra Martinez, Aquilino [COPPE/UFRJ, Programa de Engenharia Nuclear, Caixa Postal 68509, CEP 21941-914, RJ (Brazil)], E-mail: aquilino@lmp.ufrj.br; Carvalho Da Silva, Fernando [COPPE/UFRJ, Programa de Engenharia Nuclear, Caixa Postal 68509, CEP 21941-914, RJ (Brazil)

2008-03-15

A new formulation is presented in this paper to solve the inverse kinetics equation. This method is based on the Laplace transform of the point kinetics equations, resulting in an expression equivalent to the inverse kinetics equation as a function of the power history. Reactivity can be written in terms of the summation of convolution with response to impulse, characteristic of a linear system. For its digital form the Z-transform is used, which is the discrete version of the Laplace transform. This new method of reactivity calculation has very special features, amongst which it can be pointed out that the linear part is characterized by a filter named finite impulse response (FIR). The FIR filter will always be, stable and non-varying in time, and, apart from this, it can be implemented in the non-recursive form. This type of implementation does not require feedback, allowing the calculation of reactivity in a continuous way.

Simulation error propagation for a dynamic rod worth measurement technique

International Nuclear Information System (INIS)

Kastanya, D.F.; Turinsky, P.J.

1996-01-01

KRSKO nuclear station, subsequently adapted by Westinghouse, introduced the dynamic rod worth measurement (DRWM) technique for measuring pressurized water reactor rod worths. This technique has the potential for reduced test time and primary loop waste water versus alternatives. The measurement is performed starting from a slightly supercritical state with all rods out (ARO), driving a bank in at the maximum stepping rate, and recording the ex-core detectors responses and bank position as a function of time. The static bank worth is obtained by (1) using the ex-core detectors' responses to obtain the core average flux (2) using the core average flux in the inverse point-kinetics equations to obtain the dynamic bank worth (3) converting the dynamic bank worth to the static bank worth. In this data interpretation process, various calculated quantities obtained from a core simulator are utilized. This paper presents an analysis of the sensitivity to the impact of core simulator errors on the deduced static bank worth

Finite differences with exponential filtering in the calculation of reactivity

International Nuclear Information System (INIS)

Suescun Diaz, Daniel; Senra Martinez, Aquilino

2010-01-01

A formulation for the calculation of reactivity using a recursive process is presented in this paper, as well as the treatment to reduce noise intensity that is found in the nuclear power signal. Using the history of nuclear power considered as the memory of such power and the filter exponentially adjusted with the least squares method, it is possible to reduce the nuclear power fluctuations without causing attenuation for the calculation of reactivity and with a smaller delay than that for low-pass filter of first order delay filter. (orig.)

Finite differences with exponential filtering in the calculation of reactivity

Energy Technology Data Exchange (ETDEWEB)

Suescun Diaz, Daniel; Senra Martinez, Aquilino [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). COPPE - Programa de Engenharia Nuclear

2010-08-15

A formulation for the calculation of reactivity using a recursive process is presented in this paper, as well as the treatment to reduce noise intensity that is found in the nuclear power signal. Using the history of nuclear power considered as the memory of such power and the filter exponentially adjusted with the least squares method, it is possible to reduce the nuclear power fluctuations without causing attenuation for the calculation of reactivity and with a smaller delay than that for low-pass filter of first order delay filter. (orig.)

Analysis of radially heterogeneous ZPPR-13A benchmark for investigating the spatial dependence of the calculated-to-experiment ratio for control rod worths

International Nuclear Information System (INIS)

Mahalakshmi, B.; Mohanakrishnan, P.

1993-01-01

Investigation were performed on the ZPPR-13A critical assembly to determine the cause of the radial variation of the calculated-to-experimental (C/E) ratio for control rod worth in large heterogeneous cores. The effects of errors in cross section, mesh size, group condensation, transport, and modeling were studied by studied by using two- and three-dimensional diffusion calculations and three-dimensional transport calculations. In that process, the cross-section set and the calculation scheme that are being used for fast reactor design in India have been revalidated. The cross-section set was found to yield satisfactory results. Three-dimensional calculations with adjusted and unadjusted cross sections confirmed that the error in cross sections was largely responsible for the radial dependence of the C/E ratios. The contributions from group condensation and mesh size errors were < 2%, and from modeling errors and transport correction, < 1%. The effect of these errors is insignificant when compared with the effect of the cross-section error. The analysis also showed that even without the adjustment in diffusion coefficient suggested in earlier studies, a satisfactory prediction is found, at least for this benchmark. The diffusion-to-transport correction for control rod worth was found to be -7%

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)

Hamming method for solving the delayed neutron precursor concentration for reactivity calculation

International Nuclear Information System (INIS)

Díaz, Daniel Suescún; Ospina, Juan Felipe Flórez; Sarasty, Jesús Andrés Rodríguez

2012-01-01

Highlights: ► We present a new formulation to calculate the reactivity using the Hamming method. ► This method shows better accuracy than existing methods for reactivity calculation. ► The reactivity is calculated without limitation of the nuclear power form. ► The method can be implemented in reactivity meters with time step of up to 0.1 s. - Abstract: We propose a new method for numerically solving the inverse point kinetic equation for a nuclear reactor using the Hamming method, without requiring the nuclear power history and without using the Laplace transform. This new method converges with accuracy of order h 5 , where h is the step in the computation time. The procedure is validated for different forms of the nuclear power and with different time steps. The results indicate that this method has a better accuracy and lower computational effort compared with other conventional methods that use the nuclear power history.

Determination of the most reactivity control rod by pseudo-harmonics perturbation method

International Nuclear Information System (INIS)

Freire, Fernando S.; Silva, Fernando C.; Martinez, Aquilino S.

2005-01-01

Frequently it is necessary to compute the change in core multiplication caused by a change in the core temperature or composition. Even when this perturbation is localized, such as a control rod inserted into the core, one does not have to repeat the original criticality calculation, but instead we can use the well-known pseudo-harmonics perturbation method to express the corresponding change in the multiplication factor in terms of the neutron flux expanded in the basis vectors characterizing the unperturbed core. Therefore we may compute the control rod worth to find the most reactivity control rod to calculate the fast shutdown margin. In this thesis we propose a simple and precise method to identify the most reactivity control rod. (author)

Experimental critical loadings and control rod worths in LWR-PROTEUS configurations compared with MCNPX results

International Nuclear Information System (INIS)

Plaschy, M.; Murphy, M.; Jatuff, F.; Seiler, R.; Chawla, R.

2006-01-01

The PROTEUS research reactor at the Paul Scherrer Inst. (PSI) has been operating since the sixties and has already permitted, due to its high flexibility, investigation of a large range of very different nuclear systems. Currently, the ongoing experimental programme is called LWR-PROTEUS. This programme was started in 1997 and concerns large-scale investigations of advanced light water reactors (LWR) fuels. Until now, the different LWR-PROTEUS phases have permitted to study more than fifteen different configurations, each of them having to be demonstrated to be operationally safe, in particular, for the Swiss safety authorities. In this context, recent developments of the PSI computer capabilities have made possible the use of full-scale SD-heterogeneous MCNPX models to calculate accurately different safety related parameters (e.g. the critical driver loading and the shutdown rod worth). The current paper presents the MCNPX predictions of these operational characteristics for seven different LWR-PROTEUS configurations using a large number of nuclear data libraries. More specifically, this significant benchmarking exercise is based on the ENDF/B6v2, ENDF/B6v8, JEF2.2, JEFF3.0, JENDL3.2, and JENDL3.3 libraries. The results highlight certain library specific trends in the prediction of the multiplication factor k eff (e.g. the systematically larger reactivity calculated with JEF2.2 and the smaller reactivity associated with JEFF3.0). They also confirm the satisfactory determination of reactivity variations by all calculational schemes, for instance, due to the introduction of a safety rod pair, these calculations having been compared with experiments. (authors)

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.

Calculation methods of reactivity using derivatives of nuclear power and Filter fir

International Nuclear Information System (INIS)

Diaz, Daniel Suescun

2007-01-01

This work presents two new methods for the solution of the inverse point kinetics equation. The first method is based on the integration by parts of the integral of the inverse point kinetics equation, which results in a power series in terms of the nuclear power in time dependence. Applying some conditions to the nuclear power, the reactivity is represented as first and second derivatives of this nuclear power. This new calculation method for reactivity has special characteristics, amongst which the possibility of using different sampling periods, and the possibility of restarting the calculation, after its interruption associated it with a possible equipment malfunction, allowing the calculation of reactivity in a non-continuous way. Apart from this reactivity can be obtained with or without dependency on the nuclear power memory. The second method is based on the Laplace transform of the point kinetics equations, resulting in an expression equivalent to the inverse kinetics equation as a function of the power history. The reactivity can be written in terms of the summation of convolution with response to impulse, characteristic of a linear system. For its digital form the Z-transform is used, which is the discrete version of the Laplace transform. In this method it can be pointed out that the linear part is equivalent to a filter named Finite Impulse Response (Fir). The Fir filter will always be, stable and non-varying in time, and, apart from this, it can be implemented in the non-recursive way. This type of implementation does not require feedback, allowing the calculation of reactivity in a continuous way. The proposed methods were validated using signals with random noise and showing the relationship between the reactivity difference and the degree of the random noise. (author)

Reactivity and reaction rate studies on the fourth loading of ZENITH

International Nuclear Information System (INIS)

Cameron, I.R.; Freemantle, R.G.; Reed, D.L.; Wilson, D.J.

1963-08-01

The determination of the excess reactivity, control rod worths, prompt neutron lifetime, flux fine structure, and reaction rates of various nuclides for the fourth loading of the heated zero energy reactor ZENITH is described. The core contains 7.76 kg of U235, giving a carbon/U235 atom ratio of 7578, and forms the most dilute of the range studied. Comparisons of the experimental results with calculations using multigroup diffusion codes are presented. (author)

Reactivity and reaction rate studies on the fourth loading of ZENITH

Energy Technology Data Exchange (ETDEWEB)

Cameron, I R; Freemantle, R G; Reed, D L; Wilson, D J [General Reactor Physics Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

1963-08-15

The determination of the excess reactivity, control rod worths, prompt neutron lifetime, flux fine structure, and reaction rates of various nuclides for the fourth loading of the heated zero energy reactor ZENITH is described. The core contains 7.76 kg of U235, giving a carbon/U235 atom ratio of 7578, and forms the most dilute of the range studied. Comparisons of the experimental results with calculations using multigroup diffusion codes are presented. (author)

Accurate reactivity void coefficient calculation for the fast spectrum reactor FBR-IME

Energy Technology Data Exchange (ETDEWEB)

Lima, Fabiano P.C.; Vellozo, Sergio de O.; Velozo, Marta J., E-mail: fabianopetruceli@outlook.com, E-mail: vellozo@cbpf.br, E-mail: martajann@gmail.com [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Secao de Engenharia Militar

2017-07-01

This paper aims to present an accurate calculation of the void reactivity coefficient for the FBR-IME, a fast spectrum reactor in development at the Engineering Military Institute (IME). The main design peculiarity lies in using mixed oxide [MOX - PuO{sub 2} + U(natural uranium)O{sub 2}] as fuel core. For this task, SCALE system was used to calculate the reactivity for several voids distributions generated by bubbles in the sodium beyond its boiling point. The results show that although the void reactivity coefficient is positive and location dependent, they are offset by other feedback effects, resulting in a negative overall coefficient. (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)

Method of allowing for resonances in calculating reactivity values

International Nuclear Information System (INIS)

Kumpf, H.

1985-01-01

On the basis of the integral transport equation for the source density an expression has been derived for calculating reactivity values taking resonances in the core and in the sample into account. The model has been used for evaluating reactivities measured in the Rossendorf SEG IV configuration. It is shown that the influence of resonances in the core can be kept tolerable, if a sufficiently thick buffer zone of only slightly absorbing non-resonant material is arranged between the sample and the core. (author)

Study on dynamic rod worth measurement method and its test verification

International Nuclear Information System (INIS)

Wu Lei; Liu Tongxian; Zhao Wenbo; Li Songling; Yu Yingrui

2015-01-01

An advanced rod worth measurement technique, the dynamic rod worth measurement method (DRWM) has been developed. Static Spatial Factors (SSF) and Dynamic Spatial Factor (DSF) were introduced to improve the inverse kinetics method. The three dimensional steady and transient simulations for the measurement process was carried out to calculate the modification factors. The rod worth measurement, test was performed on a research reactor to verify DRWM. The results showed that the DRWM method provided the improved accuracy and could be a replacement of the traditional methods. (authors)

Coupled core criticality calculations with control rods located in the central reflector region

Energy Technology Data Exchange (ETDEWEB)

Sobhy, M [Reactor depatrment, nuclear research center, Inshaas (Egypt)

1995-10-01

The reactivity of a coupled core is controlled by a set of control rods distributed in the central reflector region. The reactor contains two compact cores cooled and moderated by light water. Control rods are designed to have reactivity worths sufficient to start, control and shutdown the coupled system. Each core in a coupled system is in subcritical conditions without any absorber then each core needs to the other core to fulfill nuclear chain reaction and to approach the criticality. In this case, each core is considered clean which is suitable for research reactor with low flux disturbance and better neutron economy, in addition to the advantage of disappearing the cut corner fuel baskets. This facilitate the in core fuel management with identical fuel baskets. Hot spots will disappear. This leads to a good heat transfer process. the excess reactivity and the shutdown margin are calculated for some of reflector as coupling region gives sufficient area for coupled core are calculated cost. The fluctuations of reactivity for coupled core are calculated by noise analysis technique and compared with that for rode core. The results show low reactivity perturbation associated with coupled core.

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)

Model for calculating the boron concentration in PWR type reactors

International Nuclear Information System (INIS)

Reis Martins Junior, L.L. dos; Vanni, E.A.

1986-01-01

A PWR boron concentration model has been developed for use with RETRAN code. The concentration model calculates the boron mass balance in the primary circuit as the injected boron mixes and is transported through the same circuit. RETRAN control blocks are used to calculate the boron concentration in fluid volumes during steady-state and transient conditions. The boron reactivity worth is obtained from the core concentration and used in RETRAN point kinetics model. A FSAR type analysis of a Steam Line Break Accident in Angra I plant was selected to test the model and the results obtained indicate a sucessfull performance. (Author) [pt

Determination of reactivity coefficients from measurable effects of small external perturbations using a bank of Kalman filters

International Nuclear Information System (INIS)

Racz, A.

1990-12-01

The goal of this paper is to present a method for the determination of reactivity coefficients in a nuclear power reactor in operation. A method based on Kalman filtering technique and the Magill-Bogler test is proposed for the determination of reactivity coefficients from measured effects of small external perturbation introduced into a steady-state power reactor. Numerical experiments are presented to justify the procedure. A realistic problem is considered: the calculation of the control rod worth. Finally a possible way is given to check the goodness of the estimation. (author) 16 refs.; 4 figs

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)

Analysis of the Rossendorf SEG experiments using the JNC route for reactor calculation

International Nuclear Information System (INIS)

Dietze, Klaus

1999-11-01

The integral experiments performed at the Rossendorf fast-thermal coupled reactor RRR/SEG have been reanalyzed using the JNC route for reactor calculation JENDL3.2/SLAROM/CITATION/JOINT/PERKY. The Rossendorf experiments comprise sample reactivity measurements with pure fission products and structural material in five configurations with different neutron and adjoint spectra. The shapes of the adjoint spectra have been designed to get high sensitivities to neutron capture or the scattering effect. The calculated neutron and adjoint spectra are in good agreement with former results obtained with the European route JEF2.2/ECCO/ERANOS. The C/E-values of the central reactivity worths of samples under investigation are given. Deviations in the results of both routes are due to the different libraries, codes, and self-shielding treatments used in the calculations. Results exceeding the experimental error are discussed. (author)

Calculation of research reactor RA power at uncontrolled reactivity changes

International Nuclear Information System (INIS)

Cupac, S.

1978-01-01

The safety analysis of research reactor RA involves also the calculation of reactor power at uncontrolled reactivity changes. The corresponding computer code, based on Point Kinetics Model has been made. The short review of method applied for solving kinetic equations is given and several examples illustrating the reactor behaviour at various reactivity changes are presented. The results already obtained are giving rather rough picture of reactor behaviour in considered situations. This is the consequence of using simplified feed back and reactor cooling models, as well as temperature reactivity coefficients, which do not correspond to the actual reactor RA structure (which is now only partly fulfilled with 80% enriched uranium fuel). (author) [sr

Application of a Virtual Reactivity Feedback Control Loop in Non-Nuclear Testing of a Fast Spectrum Reactor

International Nuclear Information System (INIS)

Bragg-Sitton, Shannon M.; Forsbacka, Matthew

2004-01-01

For a compact, fast-spectrum reactor, reactivity feedback is dominated by core deformation at elevated temperature. Given the use of accurate deformation measurement techniques, it is possible to simulate nuclear feedback in non-nuclear electrically heated reactor tests. Implementation of simulated reactivity feedback in response to measured deflection is being tested at the Nasa Marshall Space Flight Center Early Flight Fission Test Facility (EFF-TF). During tests of the SAFE-100 reactor prototype, core deflection was monitored using a high resolution camera. 'Virtual' reactivity feedback was accomplished by applying the results of Monte Carlo calculations (MCNPX) to core deflection measurements; the computational analysis was used to establish the reactivity worth of various core deformations. The power delivered to the SAFE-100 prototype was then adjusted accordingly via kinetics calculations. The work presented in this paper will demonstrate virtual reactivity feedback as core power was increased from 1 kWt to 10 kWt, held approximately constant at 10 kWt, and then allowed to decrease based on the negative thermal reactivity coefficient. (authors)

Application of the Firefly and Luus-Jaakola algorithms in the calculation of a double reactive azeotrope

Science.gov (United States)

Mendes Platt, Gustavo; Pinheiro Domingos, Roberto; Oliveira de Andrade, Matheus

2014-01-01

The calculation of reactive azeotropes is an important task in the preliminary design and simulation of reactive distillation columns. Classically, homogeneous nonreactive azeotropes are vapor-liquid coexistence conditions where phase compositions are equal. For homogeneous reactive azeotropes, simultaneous phase and chemical equilibria occur concomitantly with equality of compositions (in the Ung-Doherty transformed space). The modeling of reactive azeotrope calculation is represented by a nonlinear algebraic system with phase equilibrium, chemical equilibrium and azeotropy equations. This nonlinear system can exhibit more than one solution, corresponding to a double reactive azeotrope. In a previous paper (Platt et al 2013 J. Phys.: Conf. Ser. 410 012020), we investigated some numerical aspects of the calculation of reactive azeotropes in the isobutene + methanol + methyl-tert-butyl-ether (with two reactive azeotropes) system using two metaheuristics: the Luus-Jaakola adaptive random search and the Firefly algorithm. Here, we use a hybrid structure (stochastic + deterministic) in order to produce accurate results for both azeotropes. After identifying the neighborhood of the reactive azeotrope, the nonlinear algebraic system is solved using Newton's method. The results indicate that using metaheuristics and some techniques devoted to the calculation of multiple minima allows both azeotropic coordinates in this reactive system to be obtains. In this sense, we provide a comprehensive analysis of a useful framework devoted to solving nonlinear systems, particularly in phase equilibrium problems.

Application of the Firefly and Luus–Jaakola algorithms in the calculation of a double reactive azeotrope

International Nuclear Information System (INIS)

Platt, Gustavo Mendes; Domingos, Roberto Pinheiro; Andrade, Matheus Oliveira de

2014-01-01

The calculation of reactive azeotropes is an important task in the preliminary design and simulation of reactive distillation columns. Classically, homogeneous nonreactive azeotropes are vapor–liquid coexistence conditions where phase compositions are equal. For homogeneous reactive azeotropes, simultaneous phase and chemical equilibria occur concomitantly with equality of compositions (in the Ung–Doherty transformed space). The modeling of reactive azeotrope calculation is represented by a nonlinear algebraic system with phase equilibrium, chemical equilibrium and azeotropy equations. This nonlinear system can exhibit more than one solution, corresponding to a double reactive azeotrope. In a previous paper (Platt et al 2013 J. Phys.: Conf. Ser. 410 012020), we investigated some numerical aspects of the calculation of reactive azeotropes in the isobutene + methanol + methyl-tert-butyl-ether (with two reactive azeotropes) system using two metaheuristics: the Luus–Jaakola adaptive random search and the Firefly algorithm. Here, we use a hybrid structure (stochastic + deterministic) in order to produce accurate results for both azeotropes. After identifying the neighborhood of the reactive azeotrope, the nonlinear algebraic system is solved using Newton's method. The results indicate that using metaheuristics and some techniques devoted to the calculation of multiple minima allows both azeotropic coordinates in this reactive system to be obtains. In this sense, we provide a comprehensive analysis of a useful framework devoted to solving nonlinear systems, particularly in phase equilibrium problems. (paper)

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)

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

Analysis of rod drop and pulsed source measurements of reactivity in the Winfrith SGHWR

International Nuclear Information System (INIS)

Brittain, I.

1970-05-01

Reactivity measurements by the rod-drop and pulsed source methods in the Winfrith SGHWR are seriously affected by spatial harmonics. A method of calculation is described which enables the spatial harmonics to be calculated in non-uniform cores in two or three dimensions, and thus allows a much more rigorous analysis of the experimental results than the usual point model. The method is used to analyse all the rod-drop measurements made during commissioning of the Winfrith SGHWR, and to comment on the results of pulsed source measurements. The reactivity worths of banks of ten and twelve shut-down tubes deduced from rod-drop and pulsed source experiments are in satisfactory agreement with each other and also with AIMAZ calculated values. The ability to calculate higher spatial harmonics in nonuniform cores is thought to be new, and may have a wider application to reactor kinetics through the method of Modal Analysis. (author)

EPRI depletion benchmark calculations using PARAGON

International Nuclear Information System (INIS)

Kucukboyaci, Vefa N.

2015-01-01

Highlights: • PARAGON depletion calculations are benchmarked against the EPRI reactivity decrement experiments. • Benchmarks cover a wide range of enrichments, burnups, cooling times, and burnable absorbers, and different depletion and storage conditions. • Results from PARAGON-SCALE scheme are more conservative relative to the benchmark data. • ENDF/B-VII based data reduces the excess conservatism and brings the predictions closer to benchmark reactivity decrement values. - Abstract: In order to conservatively apply burnup credit in spent fuel pool criticality analyses, code validation for both fresh and used fuel is required. Fresh fuel validation is typically done by modeling experiments from the “International Handbook.” A depletion validation can determine a bias and bias uncertainty for the worth of the isotopes not found in the fresh fuel critical experiments. Westinghouse’s burnup credit methodology uses PARAGON™ (Westinghouse 2-D lattice physics code) and its 70-group cross-section library, which have been benchmarked, qualified, and licensed both as a standalone transport code and as a nuclear data source for core design simulations. A bias and bias uncertainty for the worth of depletion isotopes, however, are not available for PARAGON. Instead, the 5% decrement approach for depletion uncertainty is used, as set forth in the Kopp memo. Recently, EPRI developed a set of benchmarks based on a large set of power distribution measurements to ascertain reactivity biases. The depletion reactivity has been used to create 11 benchmark cases for 10, 20, 30, 40, 50, and 60 GWd/MTU and 3 cooling times 100 h, 5 years, and 15 years. These benchmark cases are analyzed with PARAGON and the SCALE package and sensitivity studies are performed using different cross-section libraries based on ENDF/B-VI.3 and ENDF/B-VII data to assess that the 5% decrement approach is conservative for determining depletion uncertainty

Is it worth to calculate the dose of radioiodine?

International Nuclear Information System (INIS)

Mikalauskas, V.; Kuprionis, G.; Vajauskas, D.

2005-01-01

Full text: Administration of empirical doses of radioiodine (RAI) has been preferred to calculated doses in many hospitals, because the need to measure the size and the iodine uptake in the thyroid involves considerable inconvenience to the patient and additional costs. The preparation of RAI of varying activities also means extra work. Today there is no general consensus on whether radioiodine should be given as a fixed dose or should be calculated. There is also no consensus regarding the question of which radiation burden should be administered to a given volume of thyroid if the activity is calculated. However, while it is possible to deliver a relatively precise dose of radiation to the thyroid gland, maybe it is worth doing this?The aim of this study was to investigate the results of different uptake and volume dependent target doses on clinical outcome of patients with hyperthyroidism in Graves' disease, multi-nodular toxic goiter or toxic adenoma after radioiodine therapy. We reviewed the records of 428 patients (389 women and 39 men, mean age 56.8±12.9 years) who had received radioiodine treatment for Graves' disease and multinodular toxic goiter (n=312) or toxic adenoma (n=116) during the period of 2000-2004 in Kaunas Medical University Hospital. Most patients were given antithyroid drug therapy in order to achieve euthyroidism before treatment with RAI. Radioiodine uptake test with repeated measurements at 2, 6, 24, 48 and/or 72 and/or 96 hr to define the effective half-life was performed. In addition, all the patients underwent thyroid ultrasonography and scintigraphy to define the volume of the thyroid. The 131I activities were calculated according to the formula of Marinelli. In addition to the normal calculation individual target doses were adjusted to the thyroid volumes of each patient before therapy. For statistical evaluation, the patients were divided into four groups: group I included those with a thyroid volume 51 ml. Statistical analysis was

Tradeoff of sodium void worth and burnup reactivity swing: Impacts on balance safety position in metallic-fueled cores

International Nuclear Information System (INIS)

Wigeland, R.A.; Turski, R.B.; Pizzica, P.A.

1994-01-01

A study has been conducted to investigate the effect of a lower sodium void worth on the consequences of severe accidents in metallic-fueled sodium-cooled reactors. Four 900 MWth designs were used for the study, where all of the reactor cores were designed based on the metallic fuel of the Integral Fast Reactor (IFR) concept. The four core designs each have different sodium void worth, in the range of -3$ to 5$. The purpose of the investigation was to determine the differences in severe accident response for the four core designs, in order to estimate the improvement in overall safety that could be achieved from a reduction in the sodium void worth for reactor cores which use a metallic fuel form

The problem of reactivity and reaction-rate calculations for mixed graphite lattices

International Nuclear Information System (INIS)

Pitcher, H.H.W.

1963-05-01

The dependence of reactor physics quantities, such as η f and Pu239/U235 fission ratio, in a single cell on the environment of the cell, and the relationship of the reactivity of a mixed lattice to the reactivity of its components, in graphite-moderated reactors are investigated. In a particular case, a mixed lattice fuelled with uranium at 0 and 3000 MWD/Te showed at 8 cm. pitch a small but appreciable change (∼ 1%) in cell quantities, and at 25 cm. pitch a smaller change. It is found that the present method of calculating lattice reactivity, ignoring intercell effects, is probably adequate for standard-pitch metal-fuelled graphite-moderated systems. More general mixed-lattice systems, particularly if accurate values of cell quantities are required, may need special calculation techniques; these are discussed, and techniques adequate for most systems are presented. (author)

High 240Pu FTR/EMC experiments and analysis: Carbide fuel and UO2 blanket subassembly worths

International Nuclear Information System (INIS)

Ombrellaro, P.A.

1977-06-01

Carbide-plutonium fuel and UO 2 blanket subassembly worth measurements performed at ANL in the EMC/LWR were analyzed. Composition exchange worth calculations were performed for: (a) the replacement of high- 240 Pu fuel composition for low- 240 Pu fuel composition and carbide-plutonium fuel composition, successively, in the center subassembly of the core; (b) the replacement of low- 240 Pu fuel composition for carbide--plutonium fuel composition in one outer driver subassembly; and (c) the replacement of the radial reflector composition with UO 2 blanket composition in one subassembly of the radial reflector. The composition exchange worth calculations were performed in two-dimensional x,y geometry, using diffusion theory and perturbation theory. Each method produces about the same calculated-to-experimental bias factors

Investigation of the Buckling-Reactivity Conversion Coefficient using SRAC and MVP codes for UO2 Lattices in TCA experiments

International Nuclear Information System (INIS)

Le Dai Dien

2008-01-01

Benchmark experiments for International Reactor Physics Benchmark Experiments (IRPhE) Project carried out at TCA, the temperature effects on reactivity were studied for light water moderated and reflected UO 2 cores with/without soluble poisons. The buckling coefficient method using the measured critical water levels was proposed by Suzaki et al. The temperature dependence of buckling coefficient of reactivity and its variance by the core configurations of the benchmark experiments was investigated using SRAC and MVP calculations. From the calculations by SRAC as well as by MVP it is seen that the K-value can be taken as an average value only for each core with temperature changes which are considered as perturbation parameter. The difference between our calculations and benchmark results which uses constant K-value for all cores proves that the results depend on K-value and it play important role in defining reactivity effect using the water level worth method. (author)

Nuclear data for the calculation of thermal reactor reactivity coefficients

International Nuclear Information System (INIS)

1989-01-01

On its 15th meeting in Vienna, 16-20 June 1986, the International Nuclear Data Committee (INDC) considered it important to review the accuracy with which changes in thermal reactor reactivity resulting from changes in temperature and coolant density can be predicted. It was noted that reactor physicists in several countries had to adjust the thermal neutron cross-section data base in order to reproduce measured reactivity coefficients. Consequently, it appeared to be essential to examine the consistency of the integral and differential cross-section data and to make all the information available which has a bearing on reactivity coefficient prediction. Following the recommendation of the INDC, the Nuclear Data Section of the International Atomic Energy Agency, therefore, convened the Advisory Group Meeting on Nuclear Data for the Calculation of Thermal Reaction Reactivity Coefficients, in Vienna, Austria, 7-10 Dec. 1987. The Conclusions and Recommendations of the meeting together with the papers presented, are submitted in the present document. A separate abstract was prepared for each of these 12 papers. Refs, figs and tabs

Advances in methods of commercial FBR core characteristics analyses. Investigations of a treatment of the double-heterogeneity and a method to calculate homogenized control rod cross sections

Energy Technology Data Exchange (ETDEWEB)

Sugino, Kazuteru [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center; Iwai, Takehiko

1998-07-01

A standard data base for FBR core nuclear design is under development in order to improve the accuracy of FBR design calculation. As a part of the development, we investigated an improved treatment of double-heterogeneity and a method to calculate homogenized control rod cross sections in a commercial reactor geometry, for the betterment of the analytical accuracy of commercial FBR core characteristics. As an improvement in the treatment of double-heterogeneity, we derived a new method (the direct method) and compared both this and conventional methods with continuous energy Monte-Carlo calculations. In addition, we investigated the applicability of the reaction rate ratio preservation method as a advanced method to calculate homogenized control rod cross sections. The present studies gave the following information: (1) An improved treatment of double-heterogeneity: for criticality the conventional method showed good agreement with Monte-Carlo result within one sigma standard deviation; the direct method was consistent with conventional one. Preliminary evaluation of effects in core characteristics other than criticality showed that the effect of sodium void reactivity (coolant reactivity) due to the double-heterogeneity was large. (2) An advanced method to calculate homogenize control rod cross sections: for control rod worths the reaction rate ratio preservation method agreed with those produced by the calculations with the control rod heterogeneity included in the core geometry; in Monju control rod worth analysis, the present method overestimated control rod worths by 1 to 2% compared with the conventional method, but these differences were caused by more accurate model in the present method and it is considered that this method is more reliable than the conventional one. These two methods investigated in this study can be directly applied to core characteristics other than criticality or control rod worth. Thus it is concluded that these methods will

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

Elements of calculation of reactivity by numerical processing

International Nuclear Information System (INIS)

Hedde, J.

1968-01-01

In order to explore the new opportunities provided by numerical techniques, the author describes the theoretical optimal conditions of a calculation in real time of reactivity from counting samples produced by a nuclear reactor. These optimal conditions can be the better approached if a more complex processing is adopted. A compromise is to be searched between the desired precision and simplicity of the numerical processing hardware. An example is reported to assess result accuracy on a wide power evolution range with a structure of reduced complexity [fr

Determination of reactivity of multiplying systems filled with spherical HTGR-fuel elements using kinetic methods with regard to the pulsed-neutron method

International Nuclear Information System (INIS)

Drueke, V.

1978-06-01

At three critical or subcritical facilities - two of them filled with spherical HTGR-fuel elements - the reactivity is determined using kinetic methods. Besides the inverskinetic method the applicability of the pulsed-neutron method is investigated. The experimental results using the pulsed-neutron method are compared partly with the inverskinetic method and partly with diffusion-calculations. It is shown, that in the HTGR the space dependence of the reactivity in radial direction is not remarkable in spite of the 'kinetic distortion'; on the contrary in axial direction - the direction of the external neutron source - space dependent reactivity worths are measured. The results of the pulsed-neutron methods of Sjoestrand and Simmons-King are rather good applicable in all configurations. For the method of Sjoestrand it is necessary to select the detector positions, whereas for Simmons-King the calculated life-time determines the results. Therefore it is proposed to compare calculated and measured decay constants of the prompt neutron field in future. (orig.) [de

Implementation of CTRLPOS, a VENTURE module for control rod position criticality searches, control rod worth curve calculations, and general criticality searches

Energy Technology Data Exchange (ETDEWEB)

Smith, L.A.; Renier, J.P.

1994-06-01

A module in the VENTURE reactor analysis code system, CTRLPOS, is developed to position control rods and perform control rod position criticality searches. The module is variably dimensioned so that calculations can be performed with any number of control rod banks each having any number of control rods. CTRLPOS can also calculate control rod worth curves for a single control rod or a bank of control rods. Control rod depletion can be calculated to provide radiation source terms. These radiation source terms can be used to predict radiation doses to personnel and estimate the shielding and long-term storage requirements for spent control rods. All of these operations are completely automated. The numerous features of the module are discussed in detail. The necessary input data for the CTRLPOS module is explained. Several sample problems are presented to show the flexibility of the module. The results presented with the sample problems show that the CTRLPOS module is a powerful tool which allows a wide variety of calculations to be easily performed.

Small Sample Reactivity Measurements in the RRR/SEG Facility: Reanalysis using TRIPOLI-4

Energy Technology Data Exchange (ETDEWEB)

Hummel, Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States); Palmiotti, Guiseppe [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-08-01

This work involved reanalyzing the RRR/SEG integral experiments performed at the Rossendorf facility in Germany throughout the 1970s and 80s. These small sample reactivity worth measurements were carried out using the pile oscillator technique for many different fission products, structural materials, and standards. The coupled fast-thermal system was designed such that the measurements would provide insight into elemental data, specifically the competing effects between neutron capture and scatter. Comparing the measured to calculated reactivity values can then provide adjustment criteria to ultimately improve nuclear data for fast reactor designs. Due to the extremely small reactivity effects measured (typically less than 1 pcm) and the specific heterogeneity of the core, the tool chosen for this analysis was TRIPOLI-4. This code allows for high fidelity 3-dimensional geometric modeling, and the most recent, unreleased version, is capable of exact perturbation theory.

Preliminary physics calculations for the Clinch River Breeder Reactor

International Nuclear Information System (INIS)

Kalimullah.

1975-01-01

Calculations of sodium void, fuel, and clad worths, power distribution, and control rod worths have been carried out for an R-Z model of the CRBR, using diffusion theory and first-order perturbation theory for material worths. The power distribution and control rod worths have also been calculated in two-dimensional triangular mesh geometry. The present results are preliminary because of inaccuracy of the reactor model and the cross sections used, but the final results are not expected to be greatly different. (U.S.)

A benchmark on the calculation of kinetic parameters based on reactivity effect experiments in the CROCUS reactor

International Nuclear Information System (INIS)

Paratte, J.M.; Frueh, R.; Kasemeyer, U.; Kalugin, M.A.; Timm, W.; Chawla, R.

2006-01-01

Measurements in the CROCUS reactor at EPFL, Lausanne, are reported for the critical water level and the inverse reactor period for several different sets of delayed supercritical conditions. The experimental configurations were also calculated by four different calculation methods. For each of the supercritical configurations, the absolute reactivity value has been determined in two different ways, viz.: (i) through direct comparison of the multiplication factor obtained employing a given calculation method with the corresponding value for the critical case (calculated reactivity: ρ calc ); (ii) by application of the inhour equation using the kinetic parameters obtained for the critical configuration and the measured inverse reactor period (measured reactivity: ρ meas ). The calculated multiplication factors for the reference critical configuration, as well as ρ calc for the supercritical cases, are found to be in good agreement. However, the values of ρ meas produced by two of the applied calculation methods differ appreciably from the corresponding ρ calc values, clearly indicating deficiencies in the kinetic parameters obtained from these methods

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 and validation of calculation schemes dedicated to the interpretation of small reactivity effects for nuclear data improvement

International Nuclear Information System (INIS)

Gruel, A.

2011-01-01

Reactivity measurements by the oscillation technique, as those performed in the Minerve reactor, enable to access various neutronic parameters on materials, fuels or specific isotopes. Usually, expected reactivity effects are small, about ten pcm at maximum. Then, the modeling of these experiments should be very precise, to obtain reliable feedback on the pointed parameters. Especially, calculation biases should be precisely identified, quantified and reduced to get precise information on nuclear data. The goal of this thesis is to develop a reference calculation scheme, with well quantified uncertainties, for in-pile oscillation experiments. In this work are presented several small reactivity calculation methods, based on deterministic and/or stochastic calculation codes. Those method are compared thanks to a numerical benchmark, against a reference calculation. Three applications of these methods are presented here: a purely deterministic calculation with exact perturbation theory formalism is used for the experimental validation of fission product cross sections, in the frame of reactivity loss studies for irradiated fuel; an hybrid method, based on a stochastic calculation and the exact perturbation theory is used for the readjustment of nuclear data, here 241 Am; and a third method, based on a perturbative Monte Carlo calculation, is used in a conception study. (author) [fr

Use of the 'DRAGON' program for the calculation of reactivity devices

International Nuclear Information System (INIS)

Mollerach, Ricardo; Fink, Jose

2003-01-01

DRAGON is a computer program developed at the Ecole Polytechnique of the University of Montreal and adopted by AECL for the transport calculations associated to reactivity devices. This report presents aspects of the implementation in NASA of the DRAGON program. Some cases of interest were evaluated. Comparisons with results of known programs as WIMS D5, and with experiments were done. a) Embalse (CANDU 6) cell without burnup and leakage. Calculations of macroscopic cross sections with WIMS and DRAGON show very good agreement with smaller differences in the thermal constants. b) Embalse fresh cell with different leakage options. c) Embalse cell with leakage and burnup. A comparison of k-infinity and k-effective with WIMS and DRAGON as a function of burnup shows that the differences ((D-W)/D) for fresh fuel are -0.17% roughly constant up to about 2500 MWd/tU, and then decrease to -0.06 % for 8500 MWd/tU. Experiments made in 1977 in ZED-2 critical facility, reported in [3], were used as a benchmark for the cell and supercell DRAGON calculations. Calculated fluxes were compared with experimental values and the agreement is so good. d) ZED-2 cell calculation. The measured buckling was used as geometric buckling. This case can be considered an experimental verification. The calculated reactivity with DRAGON is about 2 mk, and can be considered satisfactory. WIMS k-effective value is about one mk higher. e) Supercell calculations for ZED-2 vertical and horizontal tube and rod adjuster using 2D and 3D models were done. Comparisons between measured and calculated fluxes in the vicinity of the adjuster rods. Incremental cross sections for these adjusters were calculated using different options. f) ZED-2 reactor calculations with PUMA reveal a good concordance with critical heights measured in experiments. The report describes also particular features of the code and recommendations regarding its use that may be useful for new users. (author)

Calculational study on reactivity effect of pipe intersections

International Nuclear Information System (INIS)

Okuno, Hiroshi; Naito, Yoshitaka; Kaneko, Toshiyuki.

1995-03-01

A simple formulation was proposed for evaluating the increment of reactivity due to the attachment of pipes to a vessel filled with fuel solution, and its validity was checked by numerical calculations. The formulation was based on the neutron balance equation which had been applied to the criticality safety analysis code MUTUAL for multi-unit systems, and the current formulation considered further the deviation of the representative neutron source point from the center of each pipe. The formulation was validated for models of 2- and 3-dimensional fuel systems by comparison with the precise calculations using the Monte Carlo code KENO-IV. For systems of pipes attached perpendicularly to the side of a cylindrical vessel, the size and number of negligible pipes were shown that corresponded to a very small increment (e.g. 0.3% Δk/k) of the neutron multiplication factor. (author)

Calculations of Changes in Reactivity during some regular periods of operation of JEN-1 MOD Reactor

International Nuclear Information System (INIS)

Alcala Ruiz, F.

1973-01-01

By a Point-Reactor model and Perturbation Theory, changes in reactivity during some regular operating periods of JEN-1 MOD Reactor have been calculated and compared with available measured values. they were in good agreement. Also changes in reactivity have been calculated during operations at higher power levels than the present one, concluding some practical consequences for the case of increasing the present power of this reactor. (Author)

The Coordination of calculation and experimental procedures in the determination of high-negative reactivities

International Nuclear Information System (INIS)

Pinegin, A.A.; Tsyganov, S.V.

1999-01-01

Usually three sources of information about the value of inserted negative reactivity (ρ) are used: dynamical experiment with reactimeters, solution of conventionally critical problems, and dynamical calculation of the process of reactivity insertion with the reactimer model. Each of them gives they own estimation of ρ. The discrepancy between these estimation could be significant, particularly noticeable in dissymmetric insertion of perturbation. The paper discusses origin of problems of estimation high negative reactivity with reactivity meter. Authors believe that correct method of high negative reactivity estimation have to include three dimensional dynamic core model for taking to account spatial effect. Moreover, some special process, such a removal of delayed neutron emitters, change in the fraction of delayed neutrons, inner source etc. (Authors)

The coordination of calculation and experimental procedures in the determination of high-negative reactivities

International Nuclear Information System (INIS)

Pinegin, A.A.; Tsyganov, S.V.

1999-01-01

Usually three sources of information about the value of inserted negative reactivity (ρ) are used: dynamical experiment with reactimeters, solution of conventionally critical problems, and dynamical calculation of the process of reactivity insertion with the reactimeter model. Each of them gives they own estimation of ρ. The discrepancy between these estimations could be significant, particularly noticeable in dissymmetric insertion of perturbation. Origin of problems of estimation high negative reactivity is discussed using the reactivity meter. A correct method of high negative reactivity estimation have to include three dimensional dinamic core model for taking to account spatial effect. Moreover, some special processes, such as removal of delayed neutron emitters, change in the fraction of delayed neutrons, inner sources are considered etc. (author)

Calculation of reactivity for safety in nuclear reactors; Calculo de la reactividad para seguridad en reactores nucleares

Energy Technology Data Exchange (ETDEWEB)

Suescun D, D. [Universidad Surcolombiana, Av. Pastrana Borrero - Carrera 1, Neiva, Huila (Colombia); Rojas A, O., E-mail: daniel.suescun@usco.edu.co [Universidad Popular Autonoma del Estado de Puebla, Av. 9 Pte 1908, Barrio de Santiago, 72410 Puebla (Mexico)

2017-09-15

The measurement of reactivity is a function of time and its calculation results from the variation in nuclear power from the inverse equation of punctual kinetics. This equation is a differential integral, where the term of the integral conserves the historical power and the differential part is directly related to the period of the reactor. In practice, in a nuclear plant, sensors are required to record the signals. For example, the movements of the control rods that cause the fluctuations of nuclear power over time commonly generate signals with noise, an event that makes difficult to estimate the reactivity. Thus is necessary and very useful to build digital reactivity meters in real time, since allows a reactor to be operated with greater security. The calculation of the reactivity is carried out using punctual kinetics, especially the concentration of delayed neutron precursors. In this work we present a new way to reduce the fluctuations in the calculation of the reactivity, for the high precision we propose the generalization of the predictor and corrector of the Adams-Bashforth-Moulton (ABM) method of order 4 to solve numerically the equations of the point kinetics for the calculation of the reactivity, without using the power history, due to the nature of the equations of the punctual kinetics, the modifiers of the different predictors are used to increase the accuracy in the approximation obtained accompanied by the filter known as Savitzky-Golay (Sg), allow to reduce the fluctuations of reactivity. It is known that the Sg filter softens and does not attenuate the nuclear power regardless of its shape, guarantees to reduce noise levels up to σ = 0.01, with a calculation time step of σ = 0.01, s. This formulation uses a polynomial approximation of Gram, with a degree d = 2, to calculate the convolution coefficients by means of an analytical formula that is implemented computationally and avoids problems of bad conditioning, caused by the inversion of a

Measurement of reactivity worths of burnable poison rods in enriched uranium graphite-moderated core simulated to high temperature gas cooled reactor

International Nuclear Information System (INIS)

Akino, Fujiyoshi; Takeuchi, Motoyoshi; Kitadate, Kenji; Yoshifuji, Hisashi; Kaneko, Yoshihiko

1980-11-01

As the core design for the Experimental Very High Temperature Gas Cooled Reactor progresses, evaluation of design precision has become increasingly important. For a high precision design, it is required to have adequate group constants based on accurate nuclear data, as well as calculation methods properly describing the physical behavior of neutrons. We, therefore, assembled a simulation core for VHTR, SHE-14, using a graphite-moderated 20%-enriched uranium Semi-Homogeneous Experimental Critical Facility (SHE), and obtained useful experimental data in evaluating the design precision. The VHTR is designed to accommodate burnable poison and control rods for reactivity compensation. Accordingly, the experimental burnable poison rods which are similar to those to be used in the experimental reactor were prepared, and their reactivity values were measured in the SHE-14 core. One to three rods of the above experimental burnable poison rods were inserted into the central column of the SHE-14 core, and the reactivity values were measured by the period and fuel rod substitution method. The results of the measurements have clearly shown that due to the self-shielding effect of B 4 C particles the reactivity value decreases with increasing particle diameter. For the particle diameter, the reactivity value is found to increase linearly with the logarithm of boron content. The measured values and those calculated are found to agree with each other within 5%. These results indicate that the reactivity of the burnable poison rod can be estimated fairly accurately by taking into account the self-shielding effect of B 4 C particles and the heterogeneity of the lattice cell. (author)

One-run Monte Carlo calculation of effective delayed neutron fraction and area-ratio reactivity

Energy Technology Data Exchange (ETDEWEB)

Zhaopeng Zhong; Talamo, Alberto; Gohar, Yousry, E-mail: zzhong@anl.gov, E-mail: alby@anl.gov, E-mail: gohar@anl.gov [Nuclear Engineering Division, Argonne National Laboratory, IL (United States)

2011-07-01

The Monte Carlo code MCNPX has been utilized to calculate the effective delayed neutron fraction and reactivity by using the area-ratio method. The effective delayed neutron fraction β{sub eff} has been calculated with the fission probability method proposed by Meulekamp and van der Marck. MCNPX was used to calculate separately the fission probability of the delayed and the prompt neutrons by using the TALLYX user subroutine of MCNPX. In this way, β{sub eff} was obtained from the one criticality (k-code) calculation without performing an adjoint calculation. The traditional k-ratio method requires two criticality calculations to calculate β{sub eff}, while this approach utilizes only one MCNPX criticality calculation. Therefore, the approach described here is referred to as a one-run method. In subcritical systems driven by a pulsed neutron source, the area-ratio method is used to calculate reactivity (in dollar units) as the ratio between the prompt and delayed areas. These areas represent the integral of the reaction rates induced from the prompt and delayed neutrons during the pulse period. Traditionally, application of the area-ratio method requires two separate fixed source MCNPX simulations: one with delayed neutrons and the other without. The number of source particles in these two simulations must be extremely high in order to obtain accurate results with low statistical errors because the values of the total and prompt areas are very close. Consequently, this approach is time consuming and suffers from the statistical errors of the two simulations. The present paper introduces a more efficient method for estimating the reactivity calculated with the area method by taking advantage of the TALLYX user subroutine of MCNPX. This subroutine has been developed for separately scoring the reaction rates caused by the delayed and the prompt neutrons during a single simulation. Therefore the method is referred to as a one run calculation. These methodologies have

One-run Monte Carlo calculation of effective delayed neutron fraction and area-ratio reactivity

International Nuclear Information System (INIS)

Zhaopeng Zhong; Talamo, Alberto; Gohar, Yousry

2011-01-01

The Monte Carlo code MCNPX has been utilized to calculate the effective delayed neutron fraction and reactivity by using the area-ratio method. The effective delayed neutron fraction β_e_f_f has been calculated with the fission probability method proposed by Meulekamp and van der Marck. MCNPX was used to calculate separately the fission probability of the delayed and the prompt neutrons by using the TALLYX user subroutine of MCNPX. In this way, β_e_f_f was obtained from the one criticality (k-code) calculation without performing an adjoint calculation. The traditional k-ratio method requires two criticality calculations to calculate β_e_f_f, while this approach utilizes only one MCNPX criticality calculation. Therefore, the approach described here is referred to as a one-run method. In subcritical systems driven by a pulsed neutron source, the area-ratio method is used to calculate reactivity (in dollar units) as the ratio between the prompt and delayed areas. These areas represent the integral of the reaction rates induced from the prompt and delayed neutrons during the pulse period. Traditionally, application of the area-ratio method requires two separate fixed source MCNPX simulations: one with delayed neutrons and the other without. The number of source particles in these two simulations must be extremely high in order to obtain accurate results with low statistical errors because the values of the total and prompt areas are very close. Consequently, this approach is time consuming and suffers from the statistical errors of the two simulations. The present paper introduces a more efficient method for estimating the reactivity calculated with the area method by taking advantage of the TALLYX user subroutine of MCNPX. This subroutine has been developed for separately scoring the reaction rates caused by the delayed and the prompt neutrons during a single simulation. Therefore the method is referred to as a one run calculation. These methodologies have been

Reactivity calculation using the Euler–Maclaurin formula

International Nuclear Information System (INIS)

Suescún-Díaz, Daniel; Rodríguez-Sarasty, Jesús A.; Figueroa-Jiménez, Jorge H.

2013-01-01

Highlights: ► Euler–Maclaurin formula has high precision and low computational cost. ► This method can be implemented in reactivity meters with time step of up to 0.1 s. ► This approach has not limitation of the nuclear power form. - Abstract: We develop an approximation method based on the Euler–Maclaurin formula for numerically solving the integral of the inverse point kinetic equation for nuclear reactor power. Due to its greater precision, this method requires fewer history points than other methods based on the nuclear power history. The approximation is validated with different forms of the nuclear power and with different time step calculations. Results suggest that this method, though easier to implement, has a better precision and lower computational costs than other methods that require the nuclear power history

Benchmark calculations on nuclear characteristics of JRR-4 HEU core by SRAC code system

International Nuclear Information System (INIS)

Arigane, Kenji

1987-04-01

The reduced enrichment program for the JRR-4 has been progressing based on JAERI's RERTR (Reduced Enrichment Research and Test Reactor) program. The SRAC (JAERI Thermal Reactor Standard Code System for Reactor Design and Analysis) is used for the neutronic design of the JRR-4 LEU Core. This report describes the benchmark calculations on the neutronic characteristics of the JRR-4 HEU Core in order to validate the calculation method. The benchmark calculations were performed on the various kind of neutronic characteristics such as excess reactivity, criticality, control rod worth, thermal neutron flux distribution, void coefficient, temperature coefficient, mass coefficient, kinetic parameters and poisoning effect by Xe-135 build up. As the result, it was confirmed that these calculated values are in satisfactory agreement with the measured values. Therefore, the calculational method by the SRAC was validated. (author)

Determination of the most reactivity control rod by pseudo-harmonics perturbation method; Determinacao da barra de controle mais reativa usando o metodo de pseudo-harmonicos

Energy Technology Data Exchange (ETDEWEB)

Freire, Fernando S.; Silva, Fernando C.; Martinez, Aquilino S. [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear]. E-mail: ffreire@con.ufrj.br; fernando@con.ufrj.br; aquilino@.con.ufrj.br

2005-07-01

Frequently it is necessary to compute the change in core multiplication caused by a change in the core temperature or composition. Even when this perturbation is localized, such as a control rod inserted into the core, one does not have to repeat the original criticality calculation, but instead we can use the well-known pseudo-harmonics perturbation method to express the corresponding change in the multiplication factor in terms of the neutron flux expanded in the basis vectors characterizing the unperturbed core. Therefore we may compute the control rod worth to find the most reactivity control rod to calculate the fast shutdown margin. In this thesis we propose a simple and precise method to identify the most reactivity control rod. (author)

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...

Calculation of the Reactivity Equivalence of Control Rods in the Second Charge of the HBWR

International Nuclear Information System (INIS)

Weissglas, P.

1960-11-01

Full text: Using current methods the reactivity equivalence of 19 31 and 37 centrally located control rods in the second charge of the HBWR has been calculated. An estimate of the available excess reactivity with clean cold core has also been made. Insertion depth was taken as 0, l/3, 2/3 and 3/3 of the core length

Calculation of the Reactivity Equivalence of Control Rods in the Second Charge of the HBWR.

Energy Technology Data Exchange (ETDEWEB)

Weissglas, P [The Swedish State Power Board, Stockholm (Sweden)

1960-11-15

Full text: Using current methods the reactivity equivalence of 19 31 and 37 centrally located control rods in the second charge of the HBWR has been calculated. An estimate of the available excess reactivity with clean cold core has also been made. Insertion depth was taken as 0, l/3, 2/3 and 3/3 of the core length.

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)

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.

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

PERL-2 and LAVR-2 programs for Monte Carlo calculation of reactivity disturbances with trajectory correlation using random numbers

International Nuclear Information System (INIS)

Kamaeva, O.B.; Polevoj, V.B.

1983-01-01

Realization of BESM-6 computer of a technique is described for calculating a wide class of reactivity disturbances by plotting trajectories in undisturbed and disturbed systems using one sequence of random numbers. The technique was realized on the base of earlier created programs of calculation of widespreed (PERL) and local (LAVR) reactivity disturbances. The efficiency of the technique and programs is demonstrated by calculation of change of effective neutron-multiplication factor when absorber is substituted for fuel element in a BFS-40 critical assembly and by calculation of control drum characteristics

Calculation-measurement comparison for control rods reactivity in RA-3 nuclear reactor

International Nuclear Information System (INIS)

Estryk, Guillermo; Gomez, Angel

2002-01-01

The RA-3 Nuclear Reactor of the Atomic Energy National Commission from Argentina, begun working with high enrichment fuel elements in 1967, and turned to low enrichment by 1990. During 1999 it was found out that several fuel elements had problems, so more than 50 % of them had to be removed from the core. Because of this, it was planned to go from core 93 to core 94 with special care from nuclear safety point of view. Core 94 was preceded by other five, T-1 to T-5, only as transitory ones. The care implied several nuclear parameters measurements: core reactivity excess, calibration of control rods, etc. Calculations were performed afterwards to simulate those measurements using the neutron diffusion code PUMA. The comparison shows a good agreement for more than 80% of the cases with differences lower than 10% in reactivity. The greatest differences were found in the last part of the control rods calibration and a better calculation of cell constants is planned to be done in order to improve the adjustment. (author)

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…

Temperature and void reactivity coefficient calculations for the high flux isotope reactor safety analysis report

International Nuclear Information System (INIS)

Engle, W.W. Jr.; Williams, L.R.

1994-07-01

This report provides documentation of a series of calculations performed in 1991 in order to provide input for the High Flux Isotope Reactor Safety Analysis Report. In particular, temperature and void reactivity coefficients were calculated for beginning-of-life, end-of-life, and xenon equilibrium (29 h) conditions. Much of the data used to prepare the computer models for these calculations was derived from the original HFIR nuclear design study

Moroccan TRIGA nuclear reactor, an important tool for the development of research, education and training

International Nuclear Information System (INIS)

St Aubin, E.; Marleau, G.

2011-01-01

Full text: We use the DRAGON and DONJON code in an optimization scheme for selecting alternative fuels in CANDU-6 reactors to develop devices reactivity worth adjustment procedure based on a coupled transport-diffusion calculation scheme that uses 3D supercell calculations and the time-average discrete refueling model. This low computer cost methodology provides various fuel management properties such as average exit burnup and maximal power peaks and also adjuster bank reactivity worth. The method is based on geometrical modifications of the adjuster rods configuration within conservative margins in order to match the total adjuster reactivity worth or the operator's action and decision time when the reactor is spuriously tripped. For the total adjuster reactivity worth optimization, we modify the pure geometrical procedure by doping the stainless steel adjuster rods with cadmium in order to achieve our goal for advanced fuel cycles. For the operator's action and decision time reactivity worth optimization, we implemented an infinite lattice model with neutron leakage in order to follow the xenon-135 built-up in out-of-core condition and to determine how much compensation time the adjuster's reactivity worth provides to operators. This model provides xenon reactivity transient in such a way that we can estimate when the xenon peaks occur, its height and also how long the core is poisoned. This method is applied to reference natural uranium fuel cycle and to a Thorium-DUPIC and a Thorium-SEU fuel cycles. Results show that our goals are achievable, albeit small fuel management penalties.

Chapter 10: Calculation of the temperature coefficient of reactivity of a graphite-moderated reactor

International Nuclear Information System (INIS)

Brown, G.; Richmond, R.; Stace, R.H.W.

1963-01-01

The temperature coefficients of reactivity of the BEPO, Windscale and Calder reactors are calculated, using the revised methods given by Lockey et al. (1956) and by Campbell and Symonds (1962). The results are compared with experimental values. (author)

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.

Calculation of reactivity without Lagrange interpolation; Calculo de la reactividad sin interpolacion de Lagrange

Energy Technology Data Exchange (ETDEWEB)

Suescun D, D.; Figueroa J, J. H. [Pontificia Universidad Javeriana Cali, Departamento de Ciencias Naturales y Matematicas, Calle 18 No. 118-250, Cali, Valle del Cauca (Colombia); Rodriguez R, K. C.; Villada P, J. P., E-mail: dsuescun@javerianacali.edu.co [Universidad del Valle, Departamento de Fisica, Calle 13 No. 100-00, Cali, Valle del Cauca (Colombia)

2015-09-15

A new method to solve numerically the inverse equation of punctual kinetics without using Lagrange interpolating polynomial is formulated; this method uses a polynomial approximation with N points based on a process of recurrence for simulating different forms of nuclear power. The results show a reliable accuracy. Furthermore, the method proposed here is suitable for real-time measurements of reactivity, with step sizes of calculations greater that Δt = 0.3 s; due to its precision can be used to implement a digital meter of reactivity in real time. (Author)

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

Impact of mesh points number on the accuracy of deterministic calculations of control rods worth for Tehran research reactor

International Nuclear Information System (INIS)

Boustani, Ehsan; Amirkabir University of Technology, Tehran; Khakshournia, Samad

2016-01-01

In this paper two different computational approaches, a deterministic and a stochastic one, were used for calculation of the control rods worth of the Tehran research reactor. For the deterministic approach the MTRPC package composed of the WIMS code and diffusion code CITVAP was used, while for the stochastic one the Monte Carlo code MCNPX was applied. On comparing our results obtained by the Monte Carlo approach and those previously reported in the Safety Analysis Report (SAR) of Tehran research reactor produced by the deterministic approach large discrepancies were seen. To uncover the root cause of these discrepancies, some efforts were made and finally was discerned that the number of spatial mesh points in the deterministic approach was the critical cause of these discrepancies. Therefore, the mesh optimization was performed for different regions of the core such that the results of deterministic approach based on the optimized mesh points have a good agreement with those obtained by the Monte Carlo approach.

Impact of mesh points number on the accuracy of deterministic calculations of control rods worth for Tehran research reactor

Energy Technology Data Exchange (ETDEWEB)

Boustani, Ehsan [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of); Amirkabir University of Technology, Tehran (Iran, Islamic Republic of). Energy Engineering and Physics Dept.; Khakshournia, Samad [Amirkabir University of Technology, Tehran (Iran, Islamic Republic of). Energy Engineering and Physics Dept.

2016-12-15

In this paper two different computational approaches, a deterministic and a stochastic one, were used for calculation of the control rods worth of the Tehran research reactor. For the deterministic approach the MTRPC package composed of the WIMS code and diffusion code CITVAP was used, while for the stochastic one the Monte Carlo code MCNPX was applied. On comparing our results obtained by the Monte Carlo approach and those previously reported in the Safety Analysis Report (SAR) of Tehran research reactor produced by the deterministic approach large discrepancies were seen. To uncover the root cause of these discrepancies, some efforts were made and finally was discerned that the number of spatial mesh points in the deterministic approach was the critical cause of these discrepancies. Therefore, the mesh optimization was performed for different regions of the core such that the results of deterministic approach based on the optimized mesh points have a good agreement with those obtained by the Monte Carlo approach.

Criticality experiment for No.2 core of DF-VI fast neutron criticality facility

International Nuclear Information System (INIS)

Yang Lijun; Liu Zhenhua; Yan Fengwen; Luo Zhiwen; Chu Chun; Liang Shuhong

2007-01-01

At the completion of the DF-VI fast neutron criticality facility, its core changed, and it was restarted and a series of experiments and measurements were made. According to the data from 29 criticality experiments, the criticality element number and mass were calculated, the control rod reactivity worth were measured by period method and rod compensate method, reactivity worth of safety rod and safety block were measured using reactivity instrument; the reactivity worth of outer elements and radial distribution of elements were measured too. Based on all the measurements mentioned above, safety operation parameters for core 2 in DF-VI fast neutron criticality facility were conformed. (authors)

Digital reactivity meter

International Nuclear Information System (INIS)

Akkus, B.; Anac, H.; Alsan, S.; Erk, S.

1991-01-01

Nowadays, various digital methods making use of microcomputers for neutron detector signals and determining the reactivity by numerical calculations are used in reactor control systems in place of classical reactivity meters. In this work, a calculation based on the ''The Time Dependent Transport Equation'' has been developed for determining the reactivity numerically. The reactivity values have been obtained utilizing a computer-based data acquisition and control system and compared with the analog reactivity meter values as well as the values calculated from the ''Inhour Equation''

Calculation of the fuel temperature coefficient of reactivity considering non-uniform radial temperature distribution in the fuel rod

Energy Technology Data Exchange (ETDEWEB)

Pazirandeh, Ali [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Science and Research Branch; Hooshyar Mobaraki, Almas

2017-07-15

The safe operation of a reactor is based on feedback models. In this paper we attempted to discuss the influence of a non-uniform radial temperature distribution on the fuel rod temperature coefficient of reactivity. The paper demonstrates that the neutron properties of a reactor core is based on effective temperature of the fuel to obtain the correct fuel temperature feedback. The value of volume-averaged temperature being used in the calculations of neutron physics with feedbacks would result in underestimating the probable event. In the calculation it is necessary to use the effective temperature of the fuel in order to provide correct accounting of the fuel temperature feedback. Fuel temperature changes in different zones of the core and consequently reactivity coefficient change are an important parameter for analysis of transient conditions. The restricting factor that compensates the inserted reactivity is the temperature reactivity coefficient and effective delayed neutron fraction.

Reactivity and neutron flux measurements in IPEN/MB-01 reactor with B4C burnable poison

International Nuclear Information System (INIS)

Fer, Nelson Custodio; Moreira, Joao Manoel Losada

2000-01-01

Burnable poison rods, made of B 4 C- Al 2 O 3 pellets with 5.01 mg/cm 3 10 B concentration, have been manufactured for a set of experiments in the IPEN/MB-01 zero-power reactor. Several core parameters which are affected by the burnable poisons rods have been measured. The principal results, for the situation in which the burnable poison rods are located near the absorber rods of a control rod, are they cause a 29% rod worth shadowing, a reduction of 39% in the local void coefficient of reactivity, a reduction of 4.8% in the isothermal temperature coefficient of reactivity, and a reduction of 9% in the thermal neutron flux in the region where the burnable poison rods are located. These experimental results will be used for the validation of burnable poison calculation methods in the CTMSP. (author)

Procedure for calculating estimated ultimate recoveries of wells in the Mississippian Barnett Shale, Bend Arch–Fort Worth Basin Province of north-central Texas

Science.gov (United States)

Leathers-Miller, Heidi M.

2017-11-28

In 2015, the U.S. Geological Survey published an assessment of technically recoverable continuous oil and gas resources of the Mississippian Barnett Shale in the Bend Arch–Fort Worth Basin Province of north-central Texas. Of the two assessment units involved in the overall assessment, one included a roughly equal number of oil wells and gas wells as classified by the U.S. Geological Survey’s standard of gas wells having production greater than or equal to 20,000 cubic feet of gas per barrel of oil and oil wells having production less than 20,000 cubic feet of gas per barrel of oil. As a result, estimated ultimate recoveries (EURs) were calculated for both oil wells and gas wells in one of the assessment units. Generally, only gas EURs or only oil EURs are calculated for an assessment unit. These EURs were calculated with data from IHS MarkitTM using DeclinePlus software in the Harmony interface and were a major component of the quantitative resource assessment. The calculated mean EURs ranged from 235 to 2,078 million cubic feet of gas and 21 to 39 thousand barrels of oil for various subsets of wells.

Critical experiment tests of bowing and expansion reactivity calculations for LMRS

International Nuclear Information System (INIS)

Schaefer, R.W.

1988-01-01

Experiments done in several LMR-type critical assemblies simulated core axial expansion, core radial expansion and bowing, coolant expansion, and control driveline expansion. For the most part new experimental techniques were developed to do these experiments. Calculations of the experiments basically used design-level methods, except when it was necessary to investigate complexities peculiar to the experiments. It was found that these feedback reactivities generally are overpredicted, but the predictions are within 30% of the experimental values. 14 refs., 2 figs., 4 tabs

Burn up calculations for the Iranian miniature reactor: A reliable and safe research reactor

International Nuclear Information System (INIS)

Faghihi, F.; Mirvakili, S.M.

2009-01-01

Presenting neutronic calculations pertaining to the Iranian miniature research reactor is the main goal of this article. This is a key to maintaining safe and reliable core operation. The following reactor core neutronic parameters were calculated: clean cold core excess reactivity (ρ ex ), control rod and shim worth, shut down margin (SDM), neutron flux distribution of the reactor core components, and reactivity feedback coefficients. Calculations for the fuel burnup and radionuclide inventory of the Iranian miniature neutron source reactor (MNSR), after 13 years of operational time, are carried out. Moreover, the amount of uranium burnup and produced plutonium, the concentrations and activities of the most important fission products, the actinide radionuclides accumulated, and the total radioactivity of the core are estimated. Flux distribution for both water and fuel temperature increases are calculated and changes of the central control rod position are investigated as well. Standard neutronic simulation codes WIMS-D4 and CITATION are employed for these studies. The input model was validated by the experimental data according to the final safety analysis report (FSAR) of the reactor. The total activity of the MNSR core is calculated including all radionuclides at the end of the core life and it is found to be equal to 1.3 x 10 3 Ci. Our investigation shows that the reactor is operating under safe and reliable conditions.

Burn up calculations for the Iranian miniature reactor: A reliable and safe research reactor

Energy Technology Data Exchange (ETDEWEB)

Faghihi, F. [Department of Nuclear Engineering, School of Engineering, Shiraz University, Shiraz 71345 (Iran, Islamic Republic of); Research Center for Radiation Protection, Shiraz University, Shiraz (Iran, Islamic Republic of)], E-mail: faghihif@shirazu.ac.ir; Mirvakili, S.M. [Department of Nuclear Engineering, School of Engineering, Shiraz University, Shiraz 71345 (Iran, Islamic Republic of)

2009-06-15

Presenting neutronic calculations pertaining to the Iranian miniature research reactor is the main goal of this article. This is a key to maintaining safe and reliable core operation. The following reactor core neutronic parameters were calculated: clean cold core excess reactivity ({rho}{sub ex}), control rod and shim worth, shut down margin (SDM), neutron flux distribution of the reactor core components, and reactivity feedback coefficients. Calculations for the fuel burnup and radionuclide inventory of the Iranian miniature neutron source reactor (MNSR), after 13 years of operational time, are carried out. Moreover, the amount of uranium burnup and produced plutonium, the concentrations and activities of the most important fission products, the actinide radionuclides accumulated, and the total radioactivity of the core are estimated. Flux distribution for both water and fuel temperature increases are calculated and changes of the central control rod position are investigated as well. Standard neutronic simulation codes WIMS-D4 and CITATION are employed for these studies. The input model was validated by the experimental data according to the final safety analysis report (FSAR) of the reactor. The total activity of the MNSR core is calculated including all radionuclides at the end of the core life and it is found to be equal to 1.3 x 10{sup 3}Ci. Our investigation shows that the reactor is operating under safe and reliable conditions.

A hierarchical procedure for calculation of risk importance measures

International Nuclear Information System (INIS)

Poern, K.; Dinsmore, S.C.

1987-01-01

Starting with a general importance definition based on conditional probabilities, a hierarchical process for calculating risk importance measures from a PSA's numerical results is developed. By the appropriate choice of events in the general definition, measures such as the risk achievement worth and the risk reduction worth can be calculated without requantifying the PSA's models. Required approximations are clearly defined and the subsequent constraints on the applicability of the process discussed. (orig.)

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

Reactivity determination of the Al2O3-B4C burnable poison as a function of its concentration in the IPEN/MB-01 reactor

International Nuclear Information System (INIS)

Giada, Marino Reis

2005-01-01

Burnable poison rods made of Al 2 O 3 -B 4 C pellets with different concentrations of 10 B have been manufactured for a set of experiments in the IPEN/MB-01 zero-power reactor. The experiments evaluated the reactivity of the burnable poison rods as a function of the 10 B concentration, and the shadowing effect on the control rod reactivity worth as a function of the distance between the burnable position rods and the control rod. The results showed that the burnable poison rods have a non-linear behavior as function of the 10 B concentration, starting to reach an asymptotic value for concentrations higher than 7 g/cm 3 of 10 B. The shadowing effect on the control rods was substantial. When the burnable poison rods were beside the control rod, its reactivity worth decreased as much as 30 %, and when they were 10,5 cm distant, the control rod worth decreased by 7 %. The MCNP results for the burnable poison reactivity effects agreed within experimental errors with the measured values. (author)

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

Development of a three dimensional homogeneous calculation model for the BFS-62 critical experiment. Preparation of adjusted equivalent measured values for sodium void reactivity values. Final report

International Nuclear Information System (INIS)

Manturov, G.; Semenov, M.; Seregin, A.; Lykova, L.

2004-01-01

The BFS-62 critical experiments are currently used as 'benchmark' for verification of IPPE codes and nuclear data, which have been used in the study of loading a significant amount of Pu in fast reactors. The BFS-62 experiments have been performed at BFS-2 critical facility of IPPE (Obninsk). The experimental program has been arranged in such a way that the effect of replacement of uranium dioxied blanket by the steel reflector as well as the effect of replacing UOX by MOX on the main characteristics of the reactor model was studied. Wide experimental program, including measurements of the criticality-keff, spectral indices, radial and axial fission rate distributions, control rod mock-up worth, sodium void reactivity effect SVRE and some other important nuclear physics parameters, was fulfilled in the core. Series of 4 BFS-62 critical assemblies have been designed for studying the changes in BN-600 reactor physics from existing state to hybrid core. All the assemblies are modeling the reactor state prior to refueling, i.e. with all control rod mock-ups withdrawn from the core. The following items are chosen for the analysis in this report: Description of the critical assembly BFS-62-3A as the 3rd assembly in a series of 4 BFS critical assemblies studying BN-600 reactor with MOX-UOX hybrid zone and steel reflector; Development of a 3D homogeneous calculation model for the BFS-62-3A critical experiment as the mock-up of BN-600 reactor with hybrid zone and steel reflector; Evaluation of measured nuclear physics parameters keff and SVRE (sodium void reactivity effect); Preparation of adjusted equivalent measured values for keff and SVRE. Main series of calculations are performed using 3D HEX-Z diffusion code TRIGEX in 26 groups, with the ABBN-93 cross-section set. In addition, precise calculations are made, in 299 groups and Ps-approximation in scattering, by Monte-Carlo code MMKKENO and discrete ordinate code TWODANT. All calculations are based on the common system

Measurement of xenon reactivity in the reactor of the nuclear ship 'MUTSU'

International Nuclear Information System (INIS)

Itagaki, Masafumi; Miyoshi, Yoshinori; Gakuhari, Kazuhiko; Okada, Noboru.

1993-01-01

This report deals with the measurement of reactivity changes caused by the increase and decrease of xenon concentration in the reactor core of the nuclear ship 'MUTSU' after a change from long-term operation at 70 % to zero power. The change in xenon reactivity was compensated by control-rod movements and the compensated reactivity was measured using a digital reactivity meter. The xenon override peak was recognized five and half hours after the start of power reduction. The equilibrium and peak reactivities of xenon were estimated by reading the initial and peak values of a theoretical curve which was fitted to the measured variation in xenon reactivity. The xenon reactivity results obtained by the present method can be considered to be accurate since no control-rod worth data were used and the measured quantity was the reactivity itself. (author)

Calculation of neutron flux and reactivity by perturbation theory at high order

International Nuclear Information System (INIS)

Silva, W.L.P. da; Silva, F.C. da; Thome Filho, Z.D.

1982-01-01

A high order pertubation theory is studied, independent of time, applied to integral parameter calculation of a nuclear reactor. A pertubative formulation, based on flux difference technique, which gives directy the reactivity and neutron flux up to the aproximation order required, is presented. As an application of the method, global pertubations represented by fuel temperature variations, are used. Tests were done aiming to verify the relevancy of the approximation order for several intensities of the pertubations considered. (E.G.) [pt

Estimating the four-factor product (ε p Pfnl Ptnl) for the accurate calculation of xenon and samarium reactivities in the Syrian Miniature Neutron Source Reactor

International Nuclear Information System (INIS)

Khattab, K.

2007-01-01

The modified 135 Xe equilibrium reactivity in the Syrian Miniature Neutron Source Reactor (MNSR) was calculated first by using the WIMSD4 and CITATION codes to estimate the four-factor product (ε p P f nl P t nl). Then, precise calculations of 135 Xe and 149 Sm concentrations and reactivities were carried out and compared during the reactor operation time and after shutdown. It was found that the 135 Xe and 149 Sm reactivities did not reach their equilibrium reactivities during the daily operating time of the reactor. The 149 Sm reactivities could be neglected compared to 135 Xe reactivities during the reactor operating time and after shutdown. (author)

Estimating the four-factor product (ε p Pfnl Ptnl) for the accurate calculation of xenon and samarium reactivities in the Syrian Miniature Neutron Source Reactor

International Nuclear Information System (INIS)

Khattab, K.

2007-01-01

The modified 135 Xe equilibrium reactivity in the Syrian Miniature Neutron Source Reactor (MNSR) was calculated first by using the WIMSD4 and CITATION codes to estimate the four-factor product (ε p P fnl P tnl ). Then, precise calculations of 135 Xe and 149 Sm concentrations and reactivities were carried out and compared during the reactor operation time and after shutdown. It was found that the 135 Xe and 149 Sm reactivities did not reach their equilibrium reactivities during the daily operating time of the reactor. The 149 Sm reactivities could be neglected compared to 135 Xe reactivities during the reactor operating time and after shutdown. (author)

Probing the reactivation process of sarin-inhibited acetylcholinesterase with α-nucleophiles: hydroxylamine anion is predicted to be a better antidote with DFT calculations.

Science.gov (United States)

Khan, Md Abdul Shafeeuulla; Lo, Rabindranath; Bandyopadhyay, Tusar; Ganguly, Bishwajit

2011-08-01

Inactivation of acetylcholinesterase (AChE) due to inhibition by organophosphorus (OP) compounds is a major threat to human since AChE is a key enzyme in neurotransmission process. Oximes are used as potential reactivators of OP-inhibited AChE due to their α-effect nucleophilic reactivity. In search of more effective reactivating agents, model studies have shown that α-effect is not so important for dephosphylation reactions. We report the importance of α-effect of nucleophilic reactivity towards the reactivation of OP-inhibited AChE with hydroxylamine anion. We have demonstrated with DFT [B3LYP/6-311G(d,p)] calculations that the reactivation process of sarin-serine adduct 2 with hydroxylamine anion is more efficient than the other nucleophiles reported. The superiority of hydroxylamine anion to reactivate the sarin-inhibited AChE with sarin-serine adducts 3 and 4 compared to formoximate anion was observed in the presence and absence of hydrogen bonding interactions of Gly121 and Gly122. The calculated results show that the rates of reactivation process of adduct 4 with hydroxylamine anion are 261 and 223 times faster than the formoximate anion in the absence and presence of such hydrogen bonding interactions. The DFT calculated results shed light on the importance of the adjacent carbonyl group of Glu202 for the reactivation of sarin-serine adduct, in particular with formoximate anion. The reverse reactivation reaction between hydroxylamine anion and sarin-serine adduct was found to be higher in energy compared to the other nucleophiles, which suggests that this α-nucleophile can be a good antidote agent for the reactivation process. Copyright © 2011 Elsevier Inc. All rights reserved.

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.…

Comparison calculation of a large sodium-cooled fast breeder reactor using the cell code MICROX-2 in connection with ENDF/B-VI and JEF-1.1 neutron data

International Nuclear Information System (INIS)

Pelloni, S.

1992-02-01

We have obtained results for a large sodium-cooled fast breeder reactor benchmark using data from the ENDF/B-VI and from Revision 1 of the JEF-1 (JEF-1.1) evaluation. The required cross sections were processed with the NJOY code system (Version 89.62) and homogenized with the spectrum cell code MICROX-2. Multigroup transport-theory calculations in 33 neutron groups (forward and adjoint) were performed using the two-dimensional code TWODANT and kinetic parameters were determined using the first-order perturbation-theory code PERT-V. We calculated eigenvalues, neutron balance data, global and regional breeding and conversion ratios, central rate ratios and reactivity worths with and without sodium, effective delayed neutron fraction and inhour reactivity, regional sodium void reactivity, and isothermal core fuel Doppler-reactivities. In particular, it is shown that good agreement (generally within one standard deviation) is achieved between these results and the average values over sixteen benchmark solutions obtained in the past. The eigenvalues predicted with ENDF/B-VI are up to 0.7% larger than those calculated with JEF-1.1 cross sections. This discrepancy is mainly due to different inelastic scattering cross sections for 23 Na and 238 U, and to different fast fission and nubar data for 239 Pu. (author) 5 figs., 30 tabs., 24 refs

Reactivity worth of gas expansion modules (GEMs) in the fast flux test facility

International Nuclear Information System (INIS)

Campbell, L.R.; Nelson, J.V.; Burke, T.M.; Rawlins, J.A.; Daughtry, J.W.; Bennett, R.A.

1986-01-01

A new passive shutdown device called a gas expansion module (GEM) has been developed at Hanford Engineering Development Laboratory to insert negative reactivity during a primary system loss of flow in a liquid-metal reactor (LMR). A GEM is a hollow removable core component which is sealed at the top and open at the bottom. An argon gas bubble trapped inside the assembly expands when core inlet pressure decreases (caused by a flow reduction) and expels sodium from the assembly. The GEMs are designed so that the level of the liquid-sodium primary system coolant within a GEM is above the top of the core when the primary pumps are operating at full flow and is below the bottom of the core when the primary pumps are off. When a GEM is placed at the boundary of the core and radial reflector, the drop in sodium level increases core neutron leakage and inserts negative reactivity. The results of these measurements confirm the effectiveness of GEMs in adding negative reactivity in loss-of-flow situations. It follows, therefore, that the inherent safety of LMRs, comparable in size to the FFTF, can be enhanced by the use of GEMs

The development and validation of control rod calculation methods

International Nuclear Information System (INIS)

Rowlands, J.L.; Sweet, D.W.; Franklin, B.M.

1979-01-01

Fission rate distributions have been measured in the zero power critical facility, ZEBRA, for a series of eight different arrays of boron carbide control rods. Diffusion theory calculations have been compared with these measurements. The normalised fission rates differ by up to about 30% in some regions, between the different arrays, and these differences are well predicted by the calculations. A development has been made to a method used to produce homogenised cross sections for lattice regions containing control rods. Calculations show that the method also reproduces the reaction rate within the rod and the fission rate dip at the surface of the rod in satisfactory agreement with the more accurate calculations which represent the fine structure of the rod. A comparison between diffusion theory and transport theory calculations of control rod reactivity worths in the CDFR shows that for the standard design method the finite mesh approximation and the difference between diffusion theory and transport theory (the transport correction) tend to cancel and result in corrections to be applied to the standard mesh diffusion theory calculations of about +- 2% or less. This result applies for mesh centred finite difference diffusion theory codes and for the arrays of natural boron carbide control rods for which the calculations were made. Improvements have also been made to the effective diffusion coefficients used in diffusion theory calculations for control rod followers and these give satisfactory agreement with transport theory calculations. (U.K.)

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...

Blackness coefficients, effective diffusion parameters, and control rod worths for thermal reactors - Methods

Energy Technology Data Exchange (ETDEWEB)

Bretscher, M M [Argonne National Laboratory, Argonne, IL 60439 (United States)

1985-07-01

Simple diffusion theory cannot be used to evaluate control rod worths in thermal neutron reactors because of the strongly absorbing character of the control material. However, reliable control rod worths can be obtained within the framework of diffusion theory if the control material is characterized by a set of mesh-dependent effective diffusion parameters. For thin slab absorbers the effective diffusion parameters can be expressed as functions of a suitably-defined pair of 'blackness coefficients'. Methods for calculating these blackness coefficients in the P1, P3, and P5 approximations, with and without scattering, are presented. For control elements whose geometry does not permit a thin slab treatment, other methods are needed for determining the effective diffusion parameters. One such method, based on reaction rate ratios, is discussed. (author)

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

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.

Reactivity costs in MARIA reactor

International Nuclear Information System (INIS)

Marcinkowska, Zuzanna E.; Pytel, Krzysztof M.; Frydrysiak, Andrzej

2017-01-01

Highlights: • The methodology for calculating consumed fuel cost of excess reactivity is proposed. • Correlation between time integral of the core excess reactivity and released energy. • Reactivity price gives number of fuel elements required for given excess reactivity. - Abstract: For the reactor operation at high power level and carrying out experiments and irradiations the major cost of reactor operation is the expense of nuclear fuel. In this paper the methodology for calculating consumed fuel cost-relatedness of excess reactivity is proposed. Reactivity costs have been determined on the basis of operating data. A number of examples of calculating the reactivity costs for processes such as: strong absorbing material irradiation, molybdenium-99 production, beryllium matrix poisoning and increased moderator temperature illustrates proposed method.

Molecular dynamics calculation of thermophysical properties for a highly reactive liquid.

Science.gov (United States)

Wang, H P; Luo, B C; Wei, B

2008-10-01

In order to further understand the physical characteristics of liquid silicon, the thermophysical properties are required over a broad temperature range. However, its high reactivity brings about great difficulties in the experimental measurement. Here we report the thermophysical properties by molecular dynamics calculation, including density, specific heat, diffusion coefficient, and surface tension. The calculation is performed with a system consisting of 64,000 atoms, and employing the Stillinger-Weber (SW) potential model and the modified embedded atom method (MEAM) potential model. The results show that the density increases as a quadratic function of undercooling, and the value calculated by SW potential model is only 2-4 % smaller than the reported experimental data. The specific heat is obtained to be 30.95 J mol;{-1}K;{-1} by SW potential model and 32.50 J mol;{-1}K;{-1} by MEAM potential model, both of which are constants in the corresponding ranges of temperature. The self-diffusion coefficient is exponentially dependent on the temperature and consistent with the Arrhenius equation. The surface tension increases linearly with the rise of undercooling and agrees well with the reported experimental results. This work provides reasonable data in much wider temperature range, especially for the undercooled metastable state.

Regarding KUR Reactivity Measurement System

International Nuclear Information System (INIS)

Nakamori, Akira; Hasegawa, Kei; Tsuchiyama, Tatsuo; Yamamoto, Toshihiro; Okumura, Ryo; Sano, Tadafumi

2012-01-01

This article reported: (1) the outline of the reactivity measurement system of Kyoto University Research Reactor (KUR), (2) the calibration data of control rod, (3) the problems and the countermeasures for range switching of linear output meter. For the laptop PC for the reactivity measurement system, there are four input signals: (1) linear output meter, (2) logarithmic output meter, (3) core temperature gauge, and (4) control rod position. The hardware of reactivity measurement system is controlled with Labview installed on the laptop. Output, reactivity, reactor period, and the change in reactivity due to temperature effect or Xenon effect are internally calculated and displayed in real-time with Labview based on the four signals above. Calculation results are recorded in the form of a spreadsheet. At KUR, the reactor core arrangement was changed, so the control rod was re-calibrated. At this time, calculated and experimental values of reactivity based on the reactivity measurement system were compared, and it was confirmed that the reactivity calculation by Labview was accurate. The range switching of linear output meter in the nuclear instrumentation should automatically change within the laptop, however sometimes this did not function properly in the early stage. It was speculated that undefined percent values during the transition of percent value were included in the calculation and caused calculation errors. The range switching started working properly after fixing this issue. (S.K.)

Qualification of JEFF3.1.1 library for high conversion reactor calculations using the ERASME/R experiment

Energy Technology Data Exchange (ETDEWEB)

Vidal, J. F.; Noguere, G.; Peneliau, Y.; Santamarina, A. [CEA, DEN, DER/SPRC/LEPh, Cadarache, F-13108 Saint-Paul-lez-Durance (France)

2012-07-01

With its low CO{sub 2} production, Nuclear Energy appears to be an efficient solution to the global warming due to green-house effect. However, current LWR reactors are poor uranium users and, pending the development of Fast Neutron Reactors, alternative concepts of PWR with higher conversion ratio (HCPWR) are being studied again at CEA, first studies dating from the middle 80's. In these French designs, low moderation ratio has been performed by tightening the lattice pitch, achieving a conversion ratio of 0.8-0.9 with a MOX fuel coming from PWR UOX recycling. Theses HCPWRs are characterized by a harder neutron spectrum and the calculation uncertainties on the fundamental neutronics parameters are increased by a factor 3 regarding a standard PWR lattice, due to the major contribution of the Plutonium isotopes and of the epithermal energy range to the reaction rates. In order to reduce these uncertainties, a 3-year experimental validation program called ERASME has been performed by CEA from 1984 to 1986 in the EOLE reactor. Monte Carlo analysis of the ERASME/R experiments with the Monte Carlo code TRIPOLI4 allowed the qualification of the recommended JEFF.3.1.1 library for major neutronics parameters. K{sub eff} of the MOX under-moderated lattice is over-predicted by 440 {+-} 830 pcm (2{sigma}); the conversion ratio, indicator of the good use of uranium, is also slightly over-predicted: 2 % {+-} 4 % (2{sigma}) and the same for B4C absorber rods worth and soluble boron worth, over-predicted by 2 %, both in the 2 standard deviations range. The radial fission maps of heterogeneities (water-holes, B4C and fertile rods) are well reproduced: maximal (C-E)/E dispersion is 1.3 %, maximal power peak error is 2.7 %. The void reactivity worth is the only parameter poorly calculated with an overprediction of +12.4% {+-} 1.5%. ERASME/R analysis of MOX reactivity, void effect and spectral indexes will contribute to the reevaluation of {sup 241}Am and Plutonium isotopes

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.

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...

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)

On the difference between DRAGON and WIMS-AECL calculations of the coolant void reactivity

International Nuclear Information System (INIS)

Altiparmakov, D.; Roubtsov, D.; Irish, J.D.

2009-01-01

A difference in the shape of the burnup dependence of the coolant void reactivity (CVR) has been observed between DRAGON and WIMS-AECL calculations. This paper discusses the root cause of the difference and assesses the impact on burnup and full-core reactor calculations. A Fortran procedure has been developed to run WIMS-AECL as necessary in order to mimic DRAGON burnup calculations with leakage effects included. The comparison of standard WIMS-AECL results and simulated DRAGON results demonstrated that the difference is due to different definitions of CVR. If the same CVR definition is used, then the results of both WIMS-AECL and DRAGON analyses are essentially indistinguishable. The discrepancies in the fuel composition and cell-averaged two-group cross sections that are due to differences in WIMS-AECL and DRAGON leakage treatments are insignificant. (author)

Analyses of criticality and reactivity for TRACY experiments based on JENDL-3.3 data library

International Nuclear Information System (INIS)

Sono, Hiroki; Miyoshi, Yoshinori; Nakajima, Ken

2003-01-01

The parameters on criticality and reactivity employed for computational simulations of the TRACY supercritical experiments were analyzed using a recently revised nuclear data library, JENDL-3.3. The parameters based on the JENDL-3.3 library were compared to those based on two former-used libraries, JENDL-3.2 and ENDF/B-VI. In the analyses computational codes, MVP, MCNP version 4C and TWOTRAN, were used. The following conclusions were obtained from the analyses: (1) The computational biases of the effective neutron multiplication factor attributable to the nuclear data libraries and to the computational codes do not depend the TRACY experimental conditions such as fuel conditions. (2) The fractional discrepancies in the kinetic parameters and coefficients of reactivity are within ∼5% between the three libraries. By comparison between calculations and measurements of the parameters, the JENDL-3.3 library is expected to give closer values to the measurements than the JENDL-3.2 and ENDF/B-VI libraries. (3) While the reactivity worth of transient rods expressed in the $ unit shows ∼5% discrepancy between the three libraries according to their respective β eff values, there is little discrepancy in that expressed in the Δk/k unit. (author)

Methods for reactor physics calculations for control rods in fast reactors

International Nuclear Information System (INIS)

Grimstone, M.J.; Rowlands, J.L.

1988-12-01

The IAEA Specialists' Meeting on ''Methods for Reactor Physics Calculations for Control Rods in Fast Reactors'' was held in Winfrith, United Kingdom, on 6-8 December, 1988. The meeting was attended by 23 participants from nine countries. The purpose of the meeting was to review the current calculational methods and their accuracy as assessed by theoretical studies and comparisons with measurements, and then to identify the requirements for improved methods or additional studies and comparisons. The control rod properties or effects to be considered were their reactivity worths, their effect on the power distribution through the core, and the reaction rates and energy deposition both within and adjacent to the rods. The meeting was divided into five sessions, in the first of which each national delegation presented a brief overview of their programme of work on calculational methods for fast reactor control rods. In the next three sessions a total of seventeen papers were presented describing calculational methods and assessments of their accuracy. The final session was a discussion to draw conclusions regarding the current status of methods and the further developments and validation work required. A separate abstract was prepared for each of the 23 papers presented at the meeting. Refs, figs and tabs

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

Calculation of nuclear reactivity using the generalised Adams-Bashforth-Moulton predictor corrector method

Energy Technology Data Exchange (ETDEWEB)

Suescun-Diaz, Daniel [Surcolombiana Univ., Neiva (Colombia). Groupo de Fisica Teorica; Narvaez-Paredes, Mauricio [Javeriana Univ., Cali (Colombia). Groupo de Matematica y Estadistica Aplicada Pontificia; Lozano-Parada, Jamie H. [Univ. del Valle, Cali (Colombia). Dept. de Ingenieria

2016-03-15

In this paper, the generalisation of the 4th-order Adams-Bashforth-Moulton predictor-corrector method is proposed to numerically solve the point kinetic equations of the nuclear reactivity calculations without using the nuclear power history. Due to the nature of the point kinetic equations, different predictor modifiers are used in order improve the precision of the approximations obtained. The results obtained with the prediction formulas and generalised corrections improve the precision when compared with previous methods and are valid for various forms of nuclear power and different time steps.

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.)

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.

EBRPOCO - a program to calculate detailed contributions of power reactivity components of EBR-II

International Nuclear Information System (INIS)

Meneghetti, D.; Kucera, D.A.

1981-01-01

The EBRPOCO program has been developed to facilitate the calculations of the power coefficients of reactivity of EBR-II loadings. The program enables contributions of various components of the power coefficient to be delineated axially for every subassembly. The program computes the reactivity contributions of the power coefficients resulting from: density reduction of sodium coolant due to temperature; displacement of sodium coolant by thermal expansions of cladding, structural rods, subassembly cans, and lower and upper axial reflectors; density reductions of these steel components due to temperature; displacement of bond-sodium (if present) in gaps by differential thermal expansions of fuel and cladding; density reduction of bond-sodium (if present) in gaps due to temperature; free axial expansion of fuel if unrestricted by cladding or restricted axial expansion of fuel determined by axial expansion of cladding. Isotopic spatial contributions to the Doppler component my also be obtained. (orig.) [de

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…

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)

Criticality coefficient calculation for a small PWR using Monte Carlo Transport Code

Energy Technology Data Exchange (ETDEWEB)

Trombetta, Debora M.; Su, Jian, E-mail: dtrombetta@nuclear.ufrj.br, E-mail: sujian@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil); Chirayath, Sunil S., E-mail: sunilsc@tamu.edu [Department of Nuclear Engineering and Nuclear Security Science and Policy Institute, Texas A and M University, TX (United States)

2015-07-01

Computational models of reactors are increasingly used to predict nuclear reactor physics parameters responsible for reactivity changes which could lead to accidents and losses. In this work, preliminary results for criticality coefficient calculation using the Monte Carlo transport code MCNPX were presented for a small PWR. The computational modeling developed consists of the core with fuel elements, radial reflectors, and control rods inside a pressure vessel. Three different geometries were simulated, a single fuel pin, a fuel assembly and the core, with the aim to compare the criticality coefficients among themselves.The criticality coefficients calculated were: Doppler Temperature Coefficient, Coolant Temperature Coefficient, Coolant Void Coefficient, Power Coefficient, and Control Rod Worth. The coefficient values calculated by the MCNP code were compared with literature results, showing good agreement with reference data, which validate the computational model developed and allow it to be used to perform more complex studies. Criticality Coefficient values for the three simulations done had little discrepancy for almost all coefficients investigated, the only exception was the Power Coefficient. Preliminary results presented show that simple modelling as a fuel assembly can describe changes at almost all the criticality coefficients, avoiding the need of a complex core simulation. (author)

Advances in neutronics calculation of fast neutron reactors - Demonstration on Super-Phenix reactor

International Nuclear Information System (INIS)

Czernecki, Sebastien

1998-01-01

The fast reactor european neutronics calculations system, ERANOS, has integrated recent improvements both in nuclear data, with the use of the adjusted nuclear library ERALIB 1 from the JEF2.2 library, and calculation methods, with the use of the new european cell code, ECCO, and the deterministic code, TGV/VARIANT. This code performs full 3-D reactor calculation in the transport theory with variational method. The aim of this work is to create and validate a new calculational scheme for fast spectrum systems offering good compromise between accuracy and running time. The new scheme is based on these improvements plus a special procedure accounting for control rod heterogeneity, which uses a reactivity equivalence homogenization. The new scheme has been validated by means of experiment/calculation comparisons, using the extensive start-up program measurements performed in Super-Phenix reactor. The validation uses also recent measurements performed in the Phenix reactor. The results are very satisfactory and show a significant improvement for almost all core parameters, especially for critical mass, control rod worth and radial subassembly power distribution. A detailed analysis of the discrepancies between the old scheme and the new one for this parameter allows to understand the separate effects of methods and nuclear data on the radial power distribution shape. (author) [fr

Calculation of reactivity by digital processing; Calcul de la reactivite par traitement numerique

Energy Technology Data Exchange (ETDEWEB)

Hedde, J. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1968-12-01

With a view to exploring the new possibilities offered by digital techniques, a description is given of the optimum theoretical conditions of a computation of the realtime reactivity using counting samples (obtained from a nuclear reactor). The degree to which these optimum conditions can be attained depends on the complexity of the processing which can be accepted. A compromise thus has to be made between the accuracy required and the simplicity of the equipment carrying out the processing. An example is given, using a relatively simple structure, which gives an idea of the accuracy of the results obtained over a wide range of reactor power. (author) [French] Dans le but d'explorer les possibilites nouvelles des techniques numeriques, on decrit les conditions theoriques optimales d'un calcul de la reactivite en temps reel a partir d'echantillons de comptage (en provenance d'un reacteur nucleaire). Ces conditions optimales peuvent etre approchees d'autant mieux que l'on accepte un traitement plus complexe. Un compromis est donc a faire entre la precision desiree et la simplicite du materiel assurant le traitement. Un exemple adoptant une structure de complexite reduite permet de juger de la precision des resultats obtenus sur une importante plage d'evolution de la puissance. (auteur)

Calculation of reactivity by digital processing; Calcul de la reactivite par traitement numerique

Energy Technology Data Exchange (ETDEWEB)

Hedde, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1968-12-01

With a view to exploring the new possibilities offered by digital techniques, a description is given of the optimum theoretical conditions of a computation of the realtime reactivity using counting samples (obtained from a nuclear reactor). The degree to which these optimum conditions can be attained depends on the complexity of the processing which can be accepted. A compromise thus has to be made between the accuracy required and the simplicity of the equipment carrying out the processing. An example is given, using a relatively simple structure, which gives an idea of the accuracy of the results obtained over a wide range of reactor power. (author) [French] Dans le but d'explorer les possibilites nouvelles des techniques numeriques, on decrit les conditions theoriques optimales d'un calcul de la reactivite en temps reel a partir d'echantillons de comptage (en provenance d'un reacteur nucleaire). Ces conditions optimales peuvent etre approchees d'autant mieux que l'on accepte un traitement plus complexe. Un compromis est donc a faire entre la precision desiree et la simplicite du materiel assurant le traitement. Un exemple adoptant une structure de complexite reduite permet de juger de la precision des resultats obtenus sur une importante plage d'evolution de la puissance. (auteur)

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...

GAPER-1D, 1-D Multigroup 1. Order Perturbation Transport Theory for Reactivity Coefficient

International Nuclear Information System (INIS)

Koch, P.K.

1976-01-01

1 - Description of problem or function: Reactivity coefficients are computed using first-order transport perturbation theory for one- dimensional multi-region reactor assemblies. The number of spatial mesh-points and energy groups is arbitrary. An elementary synthesis scheme is employed for treatment of two- and three-dimensional problems. The contributions to the change in inverse multiplication factor, delta(1/k), from perturbations in the individual capture, net fission, total scattering, (n,2n), inelastic scattering, and leakage cross sections are computed. A multi-dimensional prompt neutron lifetime calculation is also available. 2 - Method of solution: Broad group cross sections for the core and perturbing or sample materials are required as input. Scalar neutron fluxes and currents, as computed by SN transport calculations, are then utilized to solve the first-order transport perturbation theory equations. A synthesis scheme is used, along with independent SN calculations in two or three dimensions, to treat a multi- dimensional assembly. Spherical harmonics expansions of the angular fluxes and scattering source terms are used with leakage and anisotropic scattering treated in a P1 approximation. The angular integrations in the perturbation theory equations are performed analytically. Various reactivity coefficients and material worths are then easily computed at specified positions in the assembly. 3 - Restrictions on the complexity of the problem: The formulation of the synthesis scheme used for two- and three-dimensional problems assumes that the fluxes and currents were computed by the DTF4 code (NESC Abstract 209). Therefore, fluxes and currents from two- or three-dimensional transport or diffusion theory codes cannot be used

Calculation of Reactivity Build up in KANUPP core in Case of Large Break LOCA

International Nuclear Information System (INIS)

Arshad, M. W.

2012-01-01

Loss of Coolant Accident (LOCA) in a Pressurized Heavy Water Reactor (PHWR) leads to coolant expulsion in a primary heat transport system resulting in depressurization and possible core voiding. This results in deterioration of cooling conditions in reactor channels and increase in power before reactor shutdown, leading to higher fuel temperatures.The objective of this thesis is to couple Thermal Hydraulics Data for finding status of 2288 fuel bundles having unique coolant density along with continuous changing state of coolant. WIMCER and CITCER are used for the core calculation in case of LOCA and Thermal Hydraulic Data is obtained from the Thermal Hydraulic code TUF (two unequal flows). These codes are coupled with each other in C programming. Due to degradation of coolant in case of LOCA, the power and reactivity start increasing. Near to 5 mk of reactivity the moderator dump start and reactor goes shut down. The result obtained from these code is followed the same trend as shown in KFSAR. (author)

TP1 - A computer program for the calculation of reactivity and kinetic parameters by one-dimensional neutron transport perturbation theory

International Nuclear Information System (INIS)

Kobayashi, K.

1979-03-01

TP1, a FORTRAN-IV program based on transport theory, has been developed to determine reactivity effects and kinetic parameters such as effective delayed neutron fractions and mean generation time by applying the usual perturbation formalism for one-dimensional geometry. Direct and adjoint angular dependent neutron fluxes are read from an interface file prepared by using the one-dimensional Ssub(n)-code DTK which provides options for slab, cylindrical and spherical geometry. Multigroup cross sections which are equivalent to those of the DTK-calculations are supplied in the SIGM-block which is also read from an interface file. This block which is usually produced by the code GRUCAL should contain the necessary delayed neutron data, which can be added to the original SIGMN-block by using the code SIGMUT. Two perturbation options are included in TP1: a) the usual first oder perturbation theory can be applied to determine probe reactivities, b) assuming that there are available direct fluxes for the unperturbed reactor system and adjoint fluxes for the perturbed system, the exact reactivity effect induced by the perturbation can be determined by an exact perturbation calculation. According to the input specifications, the output lists the reactivity contributions for each neutron reaction process in the desired detailed spatial and energy group resolution. (orig./RW) [de

Verification and uncertainty evaluation of HELIOS/MASTER nuclear design system

Energy Technology Data Exchange (ETDEWEB)

Song, Jae Seung; Kim, J. C.; Cho, B. O. [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-03-01

A nuclear design system HELIOS/MASTER was established and core follow calculations were performed for Yonggwang Unit 1 cycles 1 through 7 and Yonggwang Unit 3 cycles 1 through 2. The accuracy of HELIOS/MASTER system was evaluated by estimations of uncertainties of reactivity and peaking factors and by comparisons of the maximum differences of isothermal temperature coefficient, inverse boron worth and control rod worth with the CASMO-3/MASTER uncertainties. The reactivity uncertainty was estimated by 362 pcm, and the uncertainties of three-dimensional, axially integrated radial, and planar peaking factors were evaluated by 0.048, 0.034, and 0.044 in relative power unit, respectively. The maximum differences of isothermal temperature coefficient, inverse boron worth and control rod worth were within the CASMO-3/MASTER uncertainties. 17 refs., 17 figs., 10 tabs. (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

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.

A group of neutronics calculations in the MNSR using the MCNP-4C code

International Nuclear Information System (INIS)

Khattab, K.; Sulieman, I.

2009-11-01

The MCNP-4C code was used to model the 3-D core configuration for the Syrian Miniature Neutron Source Reactor (MNSR). The continuous energy neutron cross sections were evaluated from ENDF/B-VI library to calculate the thermal and fast neutron fluxes in the MNSR inner and outer irradiation sites. The thermal fluxes in the MNSR inner irradiation sites were measured for the first time using the multiple foil activation method. Good agreements were noticed between the calculated and measured results. This model is used as well to calculate neutron flux spectrum in the reactor inner and outer irradiation sites and the reactor thermal power. Three 3-D neutronic models for the Syrian MNSR reactor using the MCNP-4C code were developed also to assess the possibility of fuel conversion from 89.87 % HEU fuel (UAl 4 -Al) to 19.75 % LEU fuel (UO 2 ). This model is used in this paper to calculate the following reactor core physics parameters: clean cold core excess reactivity, calibration of the control rod worth and calculation its shut down margin, calibration of the top beryllium shim plate reflector, axial neutron flux distributions in the inner and outer irradiation sites and the kinetics parameters ( ι p l and β e ff). (authors)

Calculations of reactivity based in the solution of the Neutron transport equation in X Y geometry and Lineal perturbation theory

International Nuclear Information System (INIS)

Valle G, E. del; Mugica R, C.A.

2005-01-01

In our country, in last congresses, Gomez et al carried out reactivity calculations based on the solution of the diffusion equation for an energy group using nodal methods in one dimension and the TPL approach (Lineal Perturbation Theory). Later on, Mugica extended the application to the case of multigroup so much so much in one as in two dimensions (X Y geometry) with excellent results. Presently work is carried out similar calculations but this time based on the solution of the neutron transport equation in X Y geometry using nodal methods and again the TPL approximation. The idea is to provide a calculation method that allows to obtain in quick form the reactivity solving the direct problem as well as the enclosed problem of the not perturbed problem. A test problem for the one that results are provided for the effective multiplication factor is described and its are offered some conclusions. (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.

JOYO MK-III performance test. Criticality test, excess reactivity measurement and burn-up coefficient measurement

International Nuclear Information System (INIS)

Maeda, Shigetaka; Sekine, Takashi; Kitano, Akihiro; Nagasaki, Hideaki

2005-03-01

The MK-III performance test began in June 2003 to fully characterize the upgraded core and heat transfer system of the experimental fast reactor JOYO. This paper describes the results of the approach to criticality, the excess reactivity evaluation and the burn-up coefficient measurement. In the approach to criticality test, the MK-III core achieved initial criticality at the control rod bank position of 412.8 mm on 14:03 July 2nd, 2003. Because the replacement of the outer two rows of reflector subassemblies with shielding subassemblies reduced the source range monitor signals by a factor of 3 at the same reactor power compared with those in the MK-II core, we measured the change of the monitor's response and determined the count rate 2x10 4 cps.' as an appropriate value judging the zero power criticality. In the excess reactivity evaluation, the zero power excess reactivity at 250degC was 2.99±0.10%Δk/kk' based on the measured critical rod bank position and the measured control rod worths. The predicted value by the JOYO core management code system HESTIA was 3.13±0.16%Δk/kk', showing good agreement with the measured value. The measured excess reactivity was within the safety requirement limit. In the burn-up coefficient measurement, the excess reactivity change versus the reactor burn-up was evaluated. The measurement method adopted was to measure the control rod positions during the rated power operation. A value of -2.12x10 -4 Δk/kk'/MWd was obtained as a measured burn-up coefficient. The value calculated by HESTIA was -2.12x10 -4 Δk/kk'/MWd, and it agreed well with the measured value. All technical safety requirements for MK-III core were satisfied and the calculation accuracy of the core management code system HESTIA was confirmed. (author)

Reactor physics methods, models, and applications used to support the conceptual design of the Advanced Neutron Source

International Nuclear Information System (INIS)

Gehin, J.C.; Worley, B.A.; Renier, J.P.; Wemple, C.A.; Jahshan, S.N.; Ryskammp, J.M.

1995-08-01

This report summarizes the neutronics analysis performed during 1991 and 1992 in support of characterization of the conceptual design of the Advanced Neutron Source (ANS). The methods used in the analysis, parametric studies, and key results supporting the design and safety evaluations of the conceptual design are presented. The analysis approach used during the conceptual design phase followed the same approach used in early ANS evaluations: (1) a strong reliance on Monte Carlo theory for beginning-of-cycle reactor performance calculations and (2) a reliance on few-group diffusion theory for reactor fuel cycle analysis and for evaluation of reactor performance at specific time steps over the fuel cycle. The Monte Carlo analysis was carried out using the MCNP continuous-energy code, and the few- group diffusion theory calculations were performed using the VENTURE and PDQ code systems. The MCNP code was used primarily for its capability to model the reflector components in realistic geometries as well as the inherent circumvention of cross-section processing requirements and use of energy-collapsed cross sections. The MCNP code was used for evaluations of reflector component reactivity effects and of heat loads in these components. The code was also used as a benchmark comparison against the diffusion-theory estimates of key reactor parameters such as region fluxes, control rod worths, reactivity coefficients, and material worths. The VENTURE and PDQ codes were used to provide independent evaluations of burnup effects, power distributions, and small perturbation worths. The performance and safety calculations performed over the subject time period are summarized, and key results are provided. The key results include flux and power distributions over the fuel cycle, silicon production rates, fuel burnup rates, component reactivities, control rod worths, component heat loads, shutdown reactivity margins, reactivity coefficients, and isotope production rates

Dependence of calculated void reactivity on film-boiling representation

International Nuclear Information System (INIS)

Whitlock, J.; Garland, W.

1992-01-01

Partial voiding of a fuel channel can lead to complicated neutronic analysis, because of highly nonuniform spatial distributions. An investigation of the distribution dependence of void reactivity in a Canada deuterium uranium (CANDU) lattice, specifically in the regime of film boiling, was done. Although the core is not expected to be critical at the time of sheath dryout, this study augments current knowledge of void reactivity in this type of lattice

Analysis of reactivity characteristics of the MONJU initial core using JENDL-3.2

Energy Technology Data Exchange (ETDEWEB)

Sasaki, Kenji; Suzuki, Takayuki; Suzuki, Norimichi [Power Reactor and Nuclear Fuel Development Corp., Tsuruga, Fukui (Japan). Monju Construction Office; Itagaki, Yoshihiko

1998-03-01

This paper describes the evaluated results of criticality, absorber rod worth and coolant worth in the MONJU initial cores based on the JENDL-3.2 library compared with those of the JENDL-2 library. We confirm that the ratios of calculated and experimental (C/E) values using the JENDL-3.2 library are slightly better than those based on the JENDL-2 library. (author)

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.

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

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.

Reactive sites influence in PMMA oligomers reactivity: a DFT study

Science.gov (United States)

Paz, C. V.; Vásquez, S. R.; Flores, N.; García, L.; Rico, J. L.

2018-01-01

In this work, we present a theoretical study of methyl methacrylate (MMA) living anionic polymerization. The study was addressed to understanding two important experimental observations made for Michael Szwarc in 1956. The unexpected effect of reactive sites concentration in the propagation rate, and the self-killer behavior of MMA (deactivating of living anionic polymerization). The theoretical calculations were performed by density functional theory (DFT) to obtain the frontier molecular orbitals values. These values were used to calculate and analyze the chemical interaction descriptors in DFT-Koopmans’ theorem. As a result, it was observed that the longest chain-length species (related with low concentration of reactive sites) exhibit the highest reactivity (behavior associated with the increase of the propagation rate). The improvement in this reactivity was attributed to the crosslinking produced in the polymethyl methacrylate chains. Meanwhile, the self-killer behavior was associated with the intermolecular forces present in the reactive sites. This behavior was associated to an obstruction in solvation, since the active sites remained active through all propagation species. The theoretical results were in good agreement with the Szwarc experiments.

Calculation of the negative reactivity inserted by the shutdown system number two (SDS2) of a CANDU reactor

Energy Technology Data Exchange (ETDEWEB)

Arsenault, B [Ecole Polytechnique, Montreal, PQ (Canada)

1994-12-31

The secondary shutdown system (SDS2) of a CANDU reactor consists of liquid poison injection through nozzles disposed horizontally across the core. The nominal concentration of gadolinium nitrate poison is 8000 ppm. With the methods available to the nuclear industry for calculating the negative reactivity inserted by the SDS2, some approximations are needed, and a simplified model of poison propagation has to be used to calculate the differential cross sections. The objective of this paper is to evaluate the errors introduced by the approximations in the supercell and core calculations. The MULTICELL and EXCELL codes gave different power distributions, and further work was recommended. 9 refs., 2 tabs., 4 figs.

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.

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.

Phase rule calculations and the thermodynamics of reactive systems under chemical equilibrium

Directory of Open Access Journals (Sweden)

PLATT G. M.

1999-01-01

Full Text Available In this paper, we examine the resolution of some phase rule problems within the context of multiple chemical equilibrium reactions, using cubic equations of state and an activity coefficient model. Bubble and dew reactive surfaces, reactive azeotropic loci and reactive critical loci are generated and presented in graphical form. Also isobaric bubble and dew reactive enthalpy loci, which may be useful in the modeling of reactive distillation operations, are depicted. All the formalism here employed is developed within the coordinate transformation of Ung and Doherty, which is appropriate for equilibrium reactive or multireactive systems. The major contribution of this work is the determination of critical loci for reactive or multireactive equilibrium systems. Since it is known that for some class of chemical reactions the kinetics and product distribution exhibit high sensitivity to pressure near criticality, the present study may be useful as a predicting tool in these cases if the chemical equilibrium condition is not too far from the real phenomenon.

A DRAGON-MCNP comparison of void reactivity calculations

Energy Technology Data Exchange (ETDEWEB)

Marleau, G [Ecole Polytechnique, Montreal, PQ (Canada). Inst. de Genie Nucleaire; Milgram, M S [Atomic Energy of Canada Ltd., Chalk River, ON (Canada)

1996-12-31

The determination of the reactivity coefficients associated with coolant voiding in a CANDU reactor is a subject which has attracted a large amount of interest in the last few years both from the theoretical and experimental point of view. One expects that deterministic codes such as DRAGON and WIMS-AECL or the MCNP4 Monte Carlo code should be able to adequately simulate the cell behaviour upon coolant voiding. However, the absence of an experimental database at equilibrium and discharge burnups has not permitted the full validation of any of these lattice codes, although a partial validation through comparison of two different computer codes has been considered. Here we present a comparison between DRAGON and MCNP4 of the void reactivity evaluation for fresh fuel. (author). 16 refs., 5 tabs.

A DRAGON-MCNP comparison of void reactivity calculations

International Nuclear Information System (INIS)

Marleau, G.

1995-01-01

The determination of the reactivity coefficients associated with coolant voiding in a CANDU reactor is a subject which has attracted a large amount of interest in the last few years both from the theoretical and experimental point of view. One expects that deterministic codes such as DRAGON and WIMS-AECL or the MCNP4 Monte Carlo code should be able to adequately simulate the cell behaviour upon coolant voiding. However, the absence of an experimental database at equilibrium and discharge burnups has not permitted the full validation of any of these lattice codes, although a partial validation through comparison of two different computer codes has been considered. Here we present a comparison between DRAGON and MCNP4 of the void reactivity evaluation for fresh fuel. (author). 16 refs., 5 tabs

Calculation methods of reactivity using derivatives of nuclear power and Filter fir; Metodos para o calculo da reatividade usando derivadas da potencia nuclear e o filtro FIR

Energy Technology Data Exchange (ETDEWEB)

Diaz, Daniel Suescun

2007-07-01

This work presents two new methods for the solution of the inverse point kinetics equation. The first method is based on the integration by parts of the integral of the inverse point kinetics equation, which results in a power series in terms of the nuclear power in time dependence. Applying some conditions to the nuclear power, the reactivity is represented as first and second derivatives of this nuclear power. This new calculation method for reactivity has special characteristics, amongst which the possibility of using different sampling periods, and the possibility of restarting the calculation, after its interruption associated it with a possible equipment malfunction, allowing the calculation of reactivity in a non-continuous way. Apart from this reactivity can be obtained with or without dependency on the nuclear power memory. The second method is based on the Laplace transform of the point kinetics equations, resulting in an expression equivalent to the inverse kinetics equation as a function of the power history. The reactivity can be written in terms of the summation of convolution with response to impulse, characteristic of a linear system. For its digital form the Z-transform is used, which is the discrete version of the Laplace transform. In this method it can be pointed out that the linear part is equivalent to a filter named Finite Impulse Response (Fir). The Fir filter will always be, stable and non-varying in time, and, apart from this, it can be implemented in the non-recursive way. This type of implementation does not require feedback, allowing the calculation of reactivity in a continuous way. The proposed methods were validated using signals with random noise and showing the relationship between the reactivity difference and the degree of the random noise. (author)

What is a free customer worth? Armchair calculations of nonpaying customers' value can lead to flawed strategies.

Science.gov (United States)

Gupta, Sunil; Mela, Carl F

2008-11-01

Free customers who are subsidized by paying customers are essential to a vast array of businesses, such as media companies, employment services, and even IT providers. But because they generate revenue only indirectly, figuring out the true value of those customers--and how much attention to devote them--has always been a challenge. Traditional customer-valuation models don't help; they focus exclusively on paying customers and largely ignore network effects, or how customers help draw other customers to a business. Now a new model, devised by professors Gupta, of Harvard Business School, and Mela, of Fuqua School of Business, takes into account not only direct network effects (where buyers attract more buyers or sellers more sellers) but also indirect network effects (where buyers attract more sellers or vice versa) . The model calculates the precise long-term impact of each additional free customer on a company's profits, factoring in the degree to which he or she brings in other customers--whether free or paying--and the ripple effect of those customers. The model helped an online auction house make several critical decisions. The business made its money on fees charged to sellers but recognized that its free customers--its buyers--were valuable, too. As competition heated up, the company worried that it wasn't wooing enough buyers. Using the model, the business discovered that the network effects of buyers were indeed large and that those customers were worth over $1,000 each--much more than had been assumed. Armed with that information, the firm increased its research on buyers, invested more in targeting them with ads, and improved their experience. The model also helped the company identify the effects of various pricing strategies on sellers, showing that they became less price-sensitive over time. As a result, the company raised the fees it charged them as well.

Validation of DRAGON4/DONJON4 simulation methodology for a typical MNSR by calculating reactivity feedback coefficient and neutron flux

Science.gov (United States)

Al Zain, Jamal; El Hajjaji, O.; El Bardouni, T.; Boukhal, H.; Jaï, Otman

2018-06-01

The MNSR is a pool type research reactor, which is difficult to model because of the importance of neutron leakage. The aim of this study is to evaluate a 2-D transport model for the reactor compatible with the latest release of the DRAGON code and 3-D diffusion of the DONJON code. DRAGON code is then used to generate the group macroscopic cross sections needed for full core diffusion calculations. The diffusion DONJON code, is then used to compute the effective multiplication factor (keff), the feedback reactivity coefficients and neutron flux which account for variation in fuel and moderator temperatures as well as the void coefficient have been calculated using the DRAGON and DONJON codes for the MNSR research reactor. The cross sections of all the reactor components at different temperatures were generated using the DRAGON code. These group constants were used then in the DONJON code to calculate the multiplication factor and the neutron spectrum at different water and fuel temperatures using 69 energy groups. Only one parameter was changed where all other parameters were kept constant. Finally, Good agreements between the calculated and measured have been obtained for every of the feedback reactivity coefficients and neutron flux.

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)

Reactivity worth measurements on fast burst reactor Caliban - description and interpretation of integral experiments for the validation of nuclear data

Energy Technology Data Exchange (ETDEWEB)

Richard, B. [Commissariat a l' Energie Atomique et Aux Energies Alternatives CEA, DAM, VALDUC, F-21120 Is-sur-Tille (France)

2012-07-01

Reactivity perturbation experiments using various materials are being performed on the HEU fast core CALIBAN, an experimental device operated by the CEA VALDUC Criticality and Neutron Transport Research Laboratory. These experiments provide valuable information to contribute to the validation of nuclear data for the materials used in such measurements. This paper presents the results obtained in a first series of measurements performed with Au-197 samples. Experiments which have been conducted in order to improve the characterization of the core are also described and discussed. The experimental results have been compared to numerical calculation using both deterministic and Monte Carlo neutron transport codes with a simplified model of the reactor. This early work led to a methodology which will be applied to the future experiments which will concern other materials of interest. (authors)

Development and qualification of reference calculation schemes for absorbers in pressured water reactor

International Nuclear Information System (INIS)

Blanc-Tranchant, P.

2001-01-01

The general field in which this work takes place is the field of the accuracy improvement of neutronic calculations, required to operate Pressurized Water Reactors (PWR) with a better precision and a lower cost. More specifically, this thesis deals with the calculation of the absorber clusters used to control these reactors. The first aim of that work was to define and validate a reference calculation route of such an absorber cluster, based on the deterministic code APOLLO2. This calculation scheme was then to be checked against experimental data. This study of the complex situation of absorber clusters required several intermediate studies, of simpler problems, such as the study of fuel rods lattices and the study of single absorber rods (B4C, AIC, Hafnium) isolated in such lattices. Each one of these different studies led to a particular reference calculation route. All these calculation routes were developed against reference continuous energy Monte-Carlo calculations, carried out with the stochastic code TRIPOLI4. They were then checked against experimental data measured during French experimental programs, undertaken within the EOLE experimental reactor, at the Nuclear Research Center of Cadarache: the MISTRAL experiments for the study of isolated absorber rods and the EPICURE experiments for the study of absorber clusters. This work led to important improvements in the calculation of isolated absorbers and absorber clusters. The reactivity worth of these clusters in particular, can now be obtained with a great accuracy: the discrepancy observed between the calculated and the experimental values is less than 2.5 %, and then slightly lower than the experimental uncertainty. (author)

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 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

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)

Development and qualification of reference calculation schemes for absorbers in pressured water reactor; Elaboration et qualification de schemas de calcul de reference pour les absorbants dans les reacteurs a eau pressurisee

Energy Technology Data Exchange (ETDEWEB)

Blanc-Tranchant, P

2001-07-01

The general field in which this work takes place is the field of the accuracy improvement of neutronic calculations, required to operate Pressurized Water Reactors (PWR) with a better precision and a lower cost. More specifically, this thesis deals with the calculation of the absorber clusters used to control these reactors. The first aim of that work was to define and validate a reference calculation route of such an absorber cluster, based on the deterministic code APOLLO2. This calculation scheme was then to be checked against experimental data. This study of the complex situation of absorber clusters required several intermediate studies, of simpler problems, such as the study of fuel rods lattices and the study of single absorber rods (B4C, AIC, Hafnium) isolated in such lattices. Each one of these different studies led to a particular reference calculation route. All these calculation routes were developed against reference continuous energy Monte-Carlo calculations, carried out with the stochastic code TRIPOLI4. They were then checked against experimental data measured during French experimental programs, undertaken within the EOLE experimental reactor, at the Nuclear Research Center of Cadarache: the MISTRAL experiments for the study of isolated absorber rods and the EPICURE experiments for the study of absorber clusters. This work led to important improvements in the calculation of isolated absorbers and absorber clusters. The reactivity worth of these clusters in particular, can now be obtained with a great accuracy: the discrepancy observed between the calculated and the experimental values is less than 2.5 %, and then slightly lower than the experimental uncertainty. (author)

ACCOUNTING OF REACTIVE POWER COMPENSATION LEVEL AT PAYMENT CALCULATION OF TECHNOLOGICAL CONSUMPTION (LOSSES OF ELECTRIC POWER FOR ITS TRANSMISSION IN POWER NETWORK

Directory of Open Access Journals (Sweden)

E. P. Zabello

2005-01-01

Full Text Available The method is proposed to make a correction in payment for consumption of reactive energy and power which is attributed to deviation of actual activation energy losses for reactive power compensation from their standard value. It is recommended to calculate standard loss values for every voltage level and actual loss values are to be determined with the help of application of remote electronic accounting means in the current mode of power consumption.

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…

Reactivity studies on the advanced neutron source

International Nuclear Information System (INIS)

Ryskamp, J.M.; Redmond, E.L. II; Fletcher, C.D.

1990-01-01

An Advanced Neutron Source (ANS) with a peak thermal neutron flux of about 8.5 x 10 19 m -2 s -1 is being designed for condensed matter physics, materials science, isotope production, and fundamental physics research. The ANS is a new reactor-based research facility being planned by Oak Ridge National Laboratory (ORNL) to meet the need for an intense steady-state source of neutrons. The design effort is currently in the conceptual phase. A reference reactor design has been selected in order to examine the safety, performance, and costs associated with this one design. The ANS Project has an established, documented safety philosophy, and safety-related design criteria are currently being established. The purpose of this paper is to present analyses of safety aspects of the reference reactor design that are related to core reactivity events. These analyses include control rod worth, shutdown rod worth, heavy water voiding, neutron beam tube flooding, light water ingress, and single fuel element criticality. Understanding these safety aspects will allow us to make design modifications that improve the reactor safety and achieve the safety related design criteria. 8 refs., 3 tabs

Contribution to the qualification of calculation methods of reactivity and of flux and power distributions in nuclear pressurized water reactor cores

International Nuclear Information System (INIS)

Abit, K.

1984-01-01

The last stage of the creation computer methods and calculations consists of verifying the running and qualifying the results obtained. The work of the present thesis consisted of improving a coupling method between radial and axial phenomena in a PWR core, refering to three-dimensional calculations, while ensuring a perfect coherence between the programmed physical models. The calculation results have been compared to measurements of reactivity and of flux distributions realized during start-up tests. Thus, the methods have been applied to the calculation of the evolution of a burnable poison (gadolinium) in view of operation in long campaign. 13 refs [fr

Cue Reactivity in Nicotine and Alcohol Addiction: A Cross-Cultural View

Science.gov (United States)

Lv, Wanwan; Wu, Qichao; Liu, Xiaoming; Chen, Ying; Song, Hongwen; Yang, Lizhuang; Zhang, Xiaochu

2016-01-01

A wealth of research indicates that cue reactivity is critical to understanding the neurobiology of nicotine and alcohol addiction and developing treatments. Functional magnetic resonance imaging (fMRI) and electroencephalograph (EEG) studies have shown abnormal cue reactivity in various conditions between nicotine or alcohol addicts and the healthy. Although the causes of these abnormalities are still unclear, cultural effect can not be ignored. We conduct an review of fMRI and EEG studies about the cue reactivity in nicotine and alcohol addiction and highlight the cultural perspective. We suggest that cultural cue reactivity is a field worth of exploring which may has an effect on addictive behavior through emotion and attention. The cultural role of nicotine and alcohol addiction would provide new insight into understanding the mechanisms of nicotine and alcohol addiction and developing culture-specific therapies. We consider that culture as a context may be a factor that causes confusing outcomes in exploring nicotine and alcohol addiction which makes it possible to control the cultural influences and further contribute to the more consistent results. PMID:27635123

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...

Development of a standard for calculation and measurement of the moderator temperature coefficient of reactivity in water-moderated power reactors

International Nuclear Information System (INIS)

Mosteller, R.D.; Hall, R.A.; Lancaster, D.B.; Young, E.H.; Gavin, P.H.; Robertson, S.T.

1998-01-01

The contents of ANS 19.11, the standard for ''Calculation and Measurement of the Moderator Temperature Coefficient of Reactivity in Water-Moderated Power Reactors,'' are described. The standard addresses the calculation of the moderator temperature coefficient (MTC) both at standby conditions and at power. In addition, it describes several methods for the measurement of the at-power MTC and assesses their relative advantages and disadvantages. Finally, it specifies a minimum set of documentation requirements for compliance with the standard

The treatment of absorber rod heterogeneity effects using homogeneous equivalent cross-sections and their application in large fast reactors

International Nuclear Information System (INIS)

Newton, T.D.

1988-01-01

This paper examines the application of homogeneous equivalent absorber rod cross-sections to the calculation of control rod anti-reactivities in large fast reactors. The method used to obtain the equivalent cross-sections is described and their validity in simple whole core geometry calculations is verified. Finally, they are employed in the calculation of control rod anti-reactivity worths in the Super Phenix 1 fast reactor and the results are compared with measured values. (author). 5 refs, 5 figs, 9 tabs

EXPERIMENTAL EVALUATION OF THE FULLY LOADED ELK RIVER REACTOR

Energy Technology Data Exchange (ETDEWEB)

Fisher, J. R.; Diaz, A.

1963-06-15

The loading and testing program of the Elk River Reactor confirmed the predicted values. The measured cold, clean excess reactivity agrees to 2% and the control rod worths to 1% of the calculated values. The reactivity for various core loadings and rod positions is tabulated. The effects of spiked elements on the reactivity and radial peak-toaverage power ratio were studied. (D.L.C.)

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.

Reactive Strength Index: A Poor Indicator of Reactive Strength?

Science.gov (United States)

Healy, Robin; Kenny, Ian; Harrison, Drew

2017-11-28

The primary aim was to assess the relationships between reactive strength measures and associated kinematic and kinetic performance variables achieved during drop jumps. A secondary aim was to highlight issues with the use of reactive strength measures as performance indicators. Twenty eight national and international level sprinters, consisting of fourteen men and women, participated in this cross-sectional analysis. Athletes performed drop jumps from a 0.3 m box onto a force platform with dependent variables contact time (CT), landing time (TLand), push-off time (TPush), flight time (FT), jump height (JH), reactive strength index (RSI, calculated as JH / CT), reactive strength ratio (RSR, calculated as FT / CT) and vertical leg spring stiffness (Kvert) recorded. Pearson's correlation test found very high to near perfect relationships between RSI and RSR (r = 0.91 to 0.97), with mixed relationships found between RSI, RSR and the key performance variables, (Men: r = -0.86 to -0.71 between RSI/RSR and CT, r = 0.80 to 0.92 between RSI/RSR and JH; Women: r = -0.85 to -0.56 between RSR and CT, r = 0.71 between RSI and JH). This study demonstrates that the method of assessing reactive strength (RSI versus RSR) may be influenced by the performance strategies adopted i.e. whether an athlete achieves their best reactive strength scores via low CTs, high JHs or a combination. Coaches are advised to limit the variability in performance strategies by implementing upper and / or lower CT thresholds to accurately compare performances between individuals.

Validation of DRAGON side-step method for Bruce-A restart Phase-B physics tests

International Nuclear Information System (INIS)

Shen, W.; Ngo-Trong, C.; Davis, R.S.

2004-01-01

The DRAGON side-step method, developed at AECL, has a number of advantages over the all-DRAGON method that was used before. It is now the qualified method for reactivity-device calculations. Although the side-step-method-generated incremental cross sections have been validated against those previously calculated with the all-DRAGON method, it is highly desirable to validate the side-step method against device-worth measurements in power reactors directly. In this paper, the DRAGON side-step method was validated by comparison with the device-calibration measurements made in Bruce-A NGS Unit 4 restart Phase-B commissioning in 2003. The validation exercise showed excellent results, with the DRAGON code overestimating the measured ZCR worth by ∼5%. A sensitivity study was also performed in this paper to assess the effect of various DRAGON modelling techniques on the incremental cross sections. The assessment shows that the refinement of meshes in 3-D and the use of the side-step method are two major reasons contributing to the improved agreement between the calculated ZCR worths and the measurements. Use of different DRAGON versions, DRAGON libraries, local-parameter core conditions, and weighting techniques for the homogenization of tube clusters inside the ZCR have a very small effect on the ZCR incremental thermal absorption cross section and ZCR reactivity worth. (author)

Sensitivity of BWR shutdown margin tests to local reactivity anomalies

International Nuclear Information System (INIS)

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

1987-01-01

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

Experimental evaluation of reactivity constraints for the closed-loop control of reactor power

International Nuclear Information System (INIS)

Bernard, J.A.; Lanning, D.D.; Ray, A.

1984-01-01

General principles for the closed-loop, digital control of reactor power have been identified, quantitatively enumerated, and experimentally demonstrated on the 5 MWt Research Reactor, MITR-II. The basic concept is to restrict the net reactivity so that it is always possible to make the reactor period infinite at the desired termination point of a transient by reversing the direction of motion of whatever control mechanism is associated with the controller. This capability is formally referred to as ''feasibility of control''. A series of ten experiments have been conducted over a period of eighteen months to demonstrate the efficacy of this property for the automatic control of reactor power. It has been shown that a controller which possesses this property is capable of both raising and lowering power in a safe, efficient manner while using a control rod of varying differential worth, that the reactivity constraints are a sufficient condition for the automatic control of reactor power, and that the use of a controller based on reactivity constraints can prevent overshoots either due to attempts to control a transient with a control rod of insufficient differential worth or due to failure to properly estimate when to commence rod insertion. Details of several of the more significant tests are presented together with a discussion of the rationale for the development of closed-loop control in large commercial power systems. Specific consideration is given to the motivation for designing a controller based on feasibility of control and the associated licensing issues

Development of a standard data base for FBR core nuclear design. 9. Analysis of FCA XVII-1 experiments

International Nuclear Information System (INIS)

Yokoyama, Kenji; Ishikawa, Makoto; Oigawa, Hiroyuki; Iijima, Susumu

1998-10-01

Pnc had developed the adjusted nuclear cross-section library in which the results of the Jupiter experiments were reflected. Using this adjusted library, the distinct improvement of the accuracy in nuclear design of Fbr cores had been achieved. As a recent research, JNC develops a database of other 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 FCA XVII-1 assembly. FCA XVII-1 is a representative mock-up of a MOX fuel sodium cooling FBR core. The criticality, reaction rate ratio, sodium void reactivity worth and 238 U Doppler reactivity worth of FCA XVII-1 were analyzed. The results of C/E values calculated by the standard analytical method for JUPITER experiments are similar to those calculated by the method of JAERI, except for the sodium void reactivity. So, further investigation for sodium void reactivity is necessary. Furthermore, sensitivity analysis shows the characteristics of FCA XVII-1 in comparison with ZPPR-9. (author)

Proceedings of the first analysis meeting on JUPITER Program. Report SN-241-80-14

Energy Technology Data Exchange (ETDEWEB)

NONE

1980-09-01

These proceedings provide an analysis of the JUPITER Experiment in ZPPR-9. A general description of data and methods is provided along with discussions of criticality, reaction rate ratio and reaction rate, Doppler and sample reactivity worth, sodium void worth, and control rod worth. Topics provided are: Considerations on a Critical Experiment Program for a Large Fast Breeder Reactor; Analysis of Sodium Void Worths in ZPPR-3 Modified Phase 3 Core; Fuel Pin and Subassembly Heterogeneity Effect on Neutronics Properties of a Fast Power Reactor; and An Evaluation Method for Mesh Size Effects in Neutron Diffusion Calculations. The individual topics have been cataloged separately.

Elaboration and qualification of a reference calculation routes for the absorbers in the PWR reactors

International Nuclear Information System (INIS)

Blanc-Tranchant, P.

1999-11-01

The general field in which this work takes place is the field of the accuracy improvement of neutronic calculations, required to operate Pressurized Water Reactors (PWR) with a better precision and a lower cost. More specifically, this thesis deals with the calculation of the absorber clusters used to control these reactors. The first aim of that work was to define and validate a reference calculation route of such an absorber cluster, based on the deterministic code Apollo 2. This calculation scheme was then to be checked against experimental data. This study of the complex situation of absorber clusters required several intermediate studies, of simpler problems, such as the study of fuel rods lattices and the study of single absorber rods (B 4 C, AIC, Hafnium) isolated in such lattices. Each one of these different studies led to a particular reference calculation route. All these calculation routes were developed against reference continuous energy Monte-Carlo calculations, carried out with the stochastic code TRIPOLI14. They were then checked against experimental data measured during french experimental programs, undertaken within the EOLE experimental reactor, at the Nuclear Research Center of Cadarache: the MISTRAL experiments for the study of isolated absorber rods and the EPICURE experiments for the study of absorber clusters. This work led to important improvements in the calculation of isolated absorbers and absorber clusters. The reactivity worth of these clusters in particular, can now be obtained with a great accuracy: the discrepancy observed between the calculated and the experimental values is less than 2.5 %, and then slightly lower than the experimental uncertainty. (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.

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

Analysis of measurements for a uranium-free core experiment at the BFS-2 critical assembly

Energy Technology Data Exchange (ETDEWEB)

Hunter, Stuart [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

1999-04-01

This document describes a series of calculations that were carried out to model various measurements from the BFS-58-1-I1 experiment. BFS-58-1-I1 was a mock-up of a uranium-free, Pu burning core at BFS-2, a Russian critical assembly operated by IPPE. The experiment measured values of Keff, Na void reactivity worth, material sample reactivity worths and reaction rate ratios. The experiments were modelled using a number of different methods. Basic nuclear data was taken from JENDL-3.2, in either 70 or 18 groups. Cross-section data for the various material regions of the assembly were calculated by either SLAROM or CASUP; the heterogeneous structure of the core regions was modelled in these calculations, with 3 different options considered for representing the (essentially 2D) geometry of the assembly components in a 1D cell model. Whole reactor calculations of flux and Keff were done using both a diffusion model (CITATION) and a transport model (TWOTRAN2), both using an RZ geometry. Reactivity worths were calculated both directly from differences in Keff values and by using the exact perturbation calculations of PERKY and SN-PERT (for CITATION and TWOTRAN2, respectively). Initial calculations included a number of inaccuracies in the assembly representation, a result of communication difficulties between JNC and IPPE; these errors were removed for the final calculations that are presented. Calculations for the experiments have also been carried out in Russia (IPPE) and France (CEA) as part of an international comparison exercise, some of those results are also presented here. The calculated value of Keff was 1.1%{delta}k/k higher than the measured value, Na void worth C/E values were {approx}1.06; these results were considered to be reasonable. (Discrepancies in certain Na void values were probably due to experimental causes , though the effect should be investigated in any future experiments.) Several sample worth values were small compared with calculational

Development of reactivity feedback effect measurement techniques under sub-critical condition in fast reactors

International Nuclear Information System (INIS)

Kitano, A.; Nishi, H.; Suzuki, T.; Okajima, S.; Kanemoto, S.

2012-01-01

The first-of-a-kind reactor has been licensed by a safety examination of the plant design based on the measured data in precedent mock-up experiments. The validity of the safety design can be confirmed without a mock-up experiment, if the reactor feed-back characteristics can be measured before operation, with the constructed reactor itself. The 'Synthesis Method', a systematic and sophisticated method of sub-criticality measurement, is proposed in this work to ensure the safety margin before operation. The 'Synthesis Method' is based on the modified source multiplication method (MSM) combined with the noise analysis method to measure the reference sub-criticality level for MSM. A numerical simulation for the control-rod reactivity worth and the isothermal feed-back reactivity was conducted for typical fast reactors of 100 MWe-size, 300 MWe-size, 750 MWe-size, and 1500 MWe-size to investigate the applicability of Synthesis Method. The number of neutron detectors and their positions necessary for the measurement were investigated for both methods of MSM and the noise analysis by a series of parametric survey calculations. As a result, it was suggested that a neutron detector located above the core center and three or more neutron detectors located above the radial blanket region enable the measurement of sub-criticality within 10% uncertainty from -$0.5 to -$2 and within 15% uncertainty for the deeper sub-criticality. (authors)

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

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.

ZnO/spiral-shaped glass for solar photocatalytic oxidation of Reactive Red 120

Directory of Open Access Journals (Sweden)

Montaser Y. Ghaly

2017-05-01

Full Text Available ZnO/glass spiral (GS was prepared by immobilization of ZnO on GS with facile method, and was characterized by X-ray diffraction analysis (XRD, scanning electron microscope (SEM and the crystallite size of ZnO on GS surface was calculated. SEM showed rod-like shape of ZnO particles on GS surface. Photocatalytic activity of prepared immobilized photocatalyst was investigated for decolourization and degradation of C.I. Reactive Red 120 (RR-120 dye under sunlight. The kinetics of decolourization and degradation removal has been investigated. The effect of pH on decolourization and degradation of dye was studied. The decolourization and degradation of dye were followed by pseudo-first order reaction. The decolourization and degradation of RR-120 dye were enhanced by H2O2 addition to definite dosage beyond that the effect is diminished. Also, the reusability of immobilized ZnO on GS was tested for photocatalytic degradation of dye and it was worth noting that it has high efficiency with slight decrease (5% after five successive runs.

Modeling of Toxicity-Relevant Electrophilic Reactivity for Guanine with Epoxides: Estimating the Hard and Soft Acids and Bases (HSAB) Parameter as a Predictor.

Science.gov (United States)

Zhang, Jing; Wang, Chenchen; Ji, Li; Liu, Weiping

2016-05-16

According to the electrophilic theory in toxicology, many chemical carcinogens in the environment and/or their active metabolites are electrophiles that exert their effects by forming covalent bonds with nucleophilic DNA centers. The theory of hard and soft acids and bases (HSAB), which states that a toxic electrophile reacts preferentially with a biological macromolecule that has a similar hardness or softness, clarifies the underlying chemistry involved in this critical event. Epoxides are hard electrophiles that are produced endogenously by the enzymatic oxidation of parent chemicals (e.g., alkenes and PAHs). Epoxide ring opening proceeds through a SN2-type mechanism with hard nucleophile DNA sites as the major facilitators of toxic effects. Thus, the quantitative prediction of chemical reactivity would enable a predictive assessment of the molecular potential to exert electrophile-mediated toxicity. In this study, we calculated the activation energies for reactions between epoxides and the guanine N7 site for a diverse set of epoxides, including aliphatic epoxides, substituted styrene oxides, and PAH epoxides, using a state-of-the-art density functional theory (DFT) method. It is worth noting that these activation energies for diverse epoxides can be further predicted by quantum chemically calculated nucleophilic indices from HSAB theory, which is a less computationally demanding method than the exacting procedure for locating the transition state. More importantly, the good qualitative/quantitative correlations between the chemical reactivity of epoxides and their bioactivity suggest that the developed model based on HSAB theory may aid in the predictive hazard evaluation of epoxides, enabling the early identification of mutagenicity/carcinogenicity-relevant SN2 reactivity.

A review of reactor physics uncertainties and validation requirements for the modular high-temperature gas-cooled reactor

International Nuclear Information System (INIS)

Baxter, A.M.; Lane, R.K.; Hettergott, E.; Lefler, W.

1991-01-01

The important, safety-related, physics parameters for the low-enriched Modular High-Temperature gas-Cooled Reactor (MHTGR) such as control rod worth, shutdown margins, temperature coefficients, and reactivity worths, are considered, and estimates are presented of the uncertainties in the calculated values of these parameters. The basis for the uncertainty estimate in several of the important calculated parameters is reviewed, including the available experimental data used in obtaining these estimates. Based on this review, the additional experimental data needed to complete the validation of the methods used to calculate these parameters is presented. The role of benchmark calculations in validating MHTGR reactor physics data is also considered. (author). 10 refs, 5 figs, 3 tabs

WIMS-AECL/RFSP code validation of reactivity calculations following a long shutdown using the simple-cell history-based method

International Nuclear Information System (INIS)

Ardeshiri, F.; Donnelly, J.V.; Arsenault, B.

1998-01-01

The purpose of this analysis is to validate the Reactor Fuelling Simulation Program (RFSP) using the simple-cell model (SCM) history-based method in a startup simulation following a reactor shutdown period. This study is part of the validation work for history-based calculations, using the WIMS-AECL code with the ENDF/B-V library, and the SCM linked to the RFSP code. In this work, the RFSP code with the SCM history-based method was used to track a 1-year period of the Point Lepreau reactor operating history, that included a 12-day reactor shutdown and subsequent startup. Measured boron and gadolinium concentrations were used in the RFSP simulations, and the predicted values of core reactivity were compared to the reference (pre-shutdown) value. The discrepancies in core reactivity are shown to be better than ±2 milli-k at any time, and better than about ±0.5 milli-k towards the end of the startup transient. The results of this analysis also show that the calculated maximum channel and bundle powers are within an acceptable range during both the core-follow and the reactor startup simulations. (author)

Development of a perturbation code, PERT-K, for hexagonal core geometry

Energy Technology Data Exchange (ETDEWEB)

Kim, Taek Kyum; Kim, Sang Ji; Song, Hoon; Kim, Young Il; Kim, Young Jin [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-01-01

A perturbation code for hexagonal core geometry has been developed based on Nodal Expansion Method. By using relevant output files of DIF3D code, it can calculate the reactivity changes caused by perturbation in composition or/and neutron cross section libraries. The accuracy of PERT-K code has been validated by calculating the reactivity changes due to fuel composition change, the sodium void coefficients, and the sample reactivity worths of BFS-73-1 critical experiments. In the case of 10% reduction in all fuel isotopics at a assembly located in the outer core, PERT-K computation agrees with the direct computation by DIF3D within 60 pcm. The sample reactivity worths of BFS-73-1 critical experiments are predicted with PERT-K code within the experimental error bounds. For 100% sodium void occurrence at the inner core, the maximum difference of reactivity changes between PERT-K and direct DIF3D computations is less than 40 pcm. On the other hand, the same sodium void condition at the outer core leads to a difference of reactivity change greater than 400 pcm. However, as sodium voiding becomes near zero value, the difference becomes less and rapidly falls within the acceptable bound, i.e. 40 pcm. (author). 11 refs., 9 figs., 6 tabs.

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)

Burn-up TRIGA Mark II benchmark experiment

International Nuclear Information System (INIS)

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

1998-01-01

Different reactor codes are used for calculations of reactor parameters. The accuracy of the programs is tested through comparison of the calculated values with the experimental results. Well-defined and accurately measured benchmarks are required. The experimental results of reactivity measurements, fuel element reactivity worth distribution and fuel-up measurements are presented in this paper. The experiments were performed with partly burnt reactor core. The experimental conditions were well defined, so that the results can be used as a burn-up benchmark test case for a TRIGA Mark II reactor calculations.(author)

Elaboration and qualification of a reference calculation routes for the absorbers in the PWR reactors; Elaboration et qualification des schemas de calcul de reference pour les absorbants dans les reacteurs a eau pressurisee

Energy Technology Data Exchange (ETDEWEB)

Blanc-Tranchant, P

1999-11-01

The general field in which this work takes place is the field of the accuracy improvement of neutronic calculations, required to operate Pressurized Water Reactors (PWR) with a better precision and a lower cost. More specifically, this thesis deals with the calculation of the absorber clusters used to control these reactors. The first aim of that work was to define and validate a reference calculation route of such an absorber cluster, based on the deterministic code Apollo 2. This calculation scheme was then to be checked against experimental data. This study of the complex situation of absorber clusters required several intermediate studies, of simpler problems, such as the study of fuel rods lattices and the study of single absorber rods (B{sub 4}C, AIC, Hafnium) isolated in such lattices. Each one of these different studies led to a particular reference calculation route. All these calculation routes were developed against reference continuous energy Monte-Carlo calculations, carried out with the stochastic code TRIPOLI14. They were then checked against experimental data measured during french experimental programs, undertaken within the EOLE experimental reactor, at the Nuclear Research Center of Cadarache: the MISTRAL experiments for the study of isolated absorber rods and the EPICURE experiments for the study of absorber clusters. This work led to important improvements in the calculation of isolated absorbers and absorber clusters. The reactivity worth of these clusters in particular, can now be obtained with a great accuracy: the discrepancy observed between the calculated and the experimental values is less than 2.5 %, and then slightly lower than the experimental uncertainty. (author)

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…

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…

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.

RA reactor reactivity changes before refurbishment - Task 3.08/02; Zadatak 3.08/02 - Promene reaktivnosti reaktora RA do remonta

Energy Technology Data Exchange (ETDEWEB)

Dobrosavljevic, N; Strugar, P; Stamenkovic, S [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Serbia and Montenegro)

1963-12-15

From the the end of 1959, when the RA reactor started operation until January 1963 reactor was operated with the initial fuel batch of 56 fuel channels. After 310 MWd 68 fuel channels were added to the reactor core, and after further 357 MWd the core was filled up to the maximum of 88 fuel channels. Basic reactor parameters were systematically measured during two years of operation. This report covers the measurements concerned directly with the reactor operation: calibration of the control rods and their reactivity worths during operation, determining the total built-in reactivity excess and its change during burnup, determination of reactivity dependence on the temperature, xenon effect in the core.

Digital reactivity meter

International Nuclear Information System (INIS)

Jiang Zongbing

1996-02-01

The importance and the usual methods of reactivity measurement in a nuclear reactor are presented. Emphasis is put upon the calculation principle, software and hardware components, main specifications, application, as well as the features of the digital reactivity meter. The test results of operation in various reactors shown that the meter possess the following features: high accuracy, short response time, low output noise, high resolution, wide measuring range, simple and flexible to operate, high stability and reliability. In addition, the reactivity meter can save the measuring data automatically and have a perfect capability of self-verifying. It not only meet the requirement of the reactivity measurement in nuclear power plant, but also can be applied to various types of reactors. (1 tab.)

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…

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

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…

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

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.

Fast critical experiments in FCA and their analysis

International Nuclear Information System (INIS)

Hirota, Jitsuya

1984-02-01

JAERI Fast Critical Facility FCA went critical for the first time in April, 1967. Since then, critical experiments and their analysis were carried out on thirty-five assemblies until march, 1982. This report summarizes many achievements obtained in these fifteen years and points out disagreements observed between the calculation and experiment for further studies. A series of mock-up experiments for Experimental Fast Reactor JOYO, a theoretical and numerical study of adjustment of group constants by using integral data and a development of proton-recoil counter system for fast neutron spectrum measurement won high praise. Studies of Doppler effect of structural materials, effect of fission product accumulation on sodium-void worth, axially heterogeneous core and actinide cross sections attracted world-side attention. Significant contributions were also made to Prototype Fast Breeder Reactor MONJU through the partial mock-up experiments. Disagreements between the calculation and experiment were observed in the following items; reaction rate distribution and reactivity worth of B 4 C absorber in radial blanket, central reactivity worth in core with reflector, plate/pin fuel heterogeneity effect on criticality, sodium-void effect in central core region, Doppler effect of structural materials, core neutron spectrum near large resonances of iron and oxygen, effect of fission product accumulation on sodium-void worth, physics property of heterogeneous core, reactivity change resulted from fuel slumping and so on. Further efforts should be made to solve these disagreements through recalculating the experimental results with newly developed data and methods and carrying out the experiments intended to identify the cause of disagreement. (author)

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…

BN600 reactivity definition

International Nuclear Information System (INIS)

Zheltyshev, V.; Ivanov, A.

2000-01-01

Since 1980, the fast BN600 reactor with sodium coolant has been operated at Beloyarsk Nuclear Power Plant. The periodic monitoring of the reactivity modifications should be implemented in compliance with the standards and regulations applied in nuclear power engineering. The reactivity measurements are carried out in order to confirm the basic neutronic features of a BN600 reactor. The reactivity measurements are aimed to justify that nuclear safety is provided in course of the in-reactor installation of the experimental core components. Two reactivity meters are to be used on BN600 operation: 1. Digital on-line reactivity calculated under stationary reactor operation on power (approximation of the point-wise kinetics is applied). 2. Second reactivity meter used to define the reactor control rod operating components efficiency under reactor startup and take account of the changing efficiency of the sensor, however, this is more time-consumptive than the on-line reactivity meter. The application of two reactivity meters allows for the monitoring of the reactor reactivity under every operating mode. (authors)

Relationship between the optimum cut off frequency for Butter worth filter and lung-heart ratio in 99mTc myocardial SPECT

International Nuclear Information System (INIS)

Salihin Yusoff, M. N.; Zakaria, A.

2010-01-01

We investigated whether the lung-heart ratio parameter can be used to identify the optimum cut off frequency for Butter worth filter in 99m Tc myocardial SPECT imaging. Materials and Methods: This study involved a cardiac phantom system consisting of cardiac insert in which 1.10 cm cold defect was inserted into its myocardium wall and filled with 4.0 μCi/ml (0.148 MBq/ml) 99m Tc concentration. The cardiac insert was then put into a cylindrical tank which filled with six different 99m Tc concentrations as background. Thus, six target background concentrations ratios (T/B) were carried out. The lung-heart ratio was determined for every SPECT raw image obtained corresponding to each T/B. Then, 130 different combinations of filter parameters from Butter worth filter were utilized to reconstruct each SPECT raw image. The determination of count in myocardium, background, and defect regions of interest were performed for every reconstructed image. All the count values were then used to calculate contrast, signal-to-noise ratio, and defect size. Each criterion was graded (1 to 100) and then summed together to obtain total grade. The optimum cut off frequency for each lung-heart ratio was determined from the total grade. The relation between optimum cut off frequency for Butter worth filter and lung-heart ratio was established using linear regression. Results: There were good relationship between the optimum Butter worth cut off frequency and lung-heart ratio (R 2 = 0.864, p<0.01). The optimal cut off frequency correspond to the change in lung-heart ratio can be expressed by the equation: Optimum cut off frequency=0.715*lung-heart ratio + 0.227. Conclusion: This study suggests that the optimum cut off frequency for Butter worth filter should be determined by referring to lung-heart ratio in each patient study.

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.

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.

Reactivity-induced time-dependencies of EBR-II linear and non-linear feedbacks

International Nuclear Information System (INIS)

Grimm, K.N.; Meneghetti, D.

1988-01-01

Time-dependent linear feedback reactivities are calculated for stereotypical subassemblies in the EBR-II reactor. These quantities are calculated from nodal reactivities obtained from a kinetic code analysis of an experiment in which the change in power resulted from the dropping of a control rod. Shown with these linear reactivities are the reactivity associated with the control-rod shaft contraction and also time-dependent non-linear (mainly bowing) component deduced from the inverse kinetics of the experimentally measured fission power and the calculated linear reactivities. (author)

Use of the 'DRAGON' program for the calculation of reactivity devices; Utilizacion del programa DRAGON para el calculo de mecanismos de reactividad

Energy Technology Data Exchange (ETDEWEB)

Mollerach, Ricardo; Fink, Jose [Nucleoelectrica Argentina SA (NASA), Buenos Aires (Argentina)

2003-07-01

DRAGON is a computer program developed at the Ecole Polytechnique of the University of Montreal and adopted by AECL for the transport calculations associated to reactivity devices. This report presents aspects of the implementation in NASA of the DRAGON program. Some cases of interest were evaluated. Comparisons with results of known programs as WIMS D5, and with experiments were done. a) Embalse (CANDU 6) cell without burnup and leakage. Calculations of macroscopic cross sections with WIMS and DRAGON show very good agreement with smaller differences in the thermal constants. b) Embalse fresh cell with different leakage options. c) Embalse cell with leakage and burnup. A comparison of k-infinity and k-effective with WIMS and DRAGON as a function of burnup shows that the differences ((D-W)/D) for fresh fuel are -0.17% roughly constant up to about 2500 MWd/tU, and then decrease to -0.06 % for 8500 MWd/tU. Experiments made in 1977 in ZED-2 critical facility, reported in [3], were used as a benchmark for the cell and supercell DRAGON calculations. Calculated fluxes were compared with experimental values and the agreement is so good. d) ZED-2 cell calculation. The measured buckling was used as geometric buckling. This case can be considered an experimental verification. The calculated reactivity with DRAGON is about 2 mk, and can be considered satisfactory. WIMS k-effective value is about one mk higher. e) Supercell calculations for ZED-2 vertical and horizontal tube and rod adjuster using 2D and 3D models were done. Comparisons between measured and calculated fluxes in the vicinity of the adjuster rods. Incremental cross sections for these adjusters were calculated using different options. f) ZED-2 reactor calculations with PUMA reveal a good concordance with critical heights measured in experiments. The report describes also particular features of the code and recommendations regarding its use that may be useful for new users. (author)

Reactivity of polychlorinated biphenyls in nucleophilic and electrophilic substitutions

Energy Technology Data Exchange (ETDEWEB)

Gorbunova, Tatyana I., E-mail: gorbunova@ios.uran.ru [I. Ya. Postovskii Institute of Organic Synthesis, Ural Branch, Russian Academy of Sciences, Kovalevskoy St., 22, Ekaterinburg 620990 (Russian Federation); Subbotina, Julia O. [Ural Federal University named after the first President of Russia B.N. Yeltsin, Mira St., 19, Ekaterinburg 620002 (Russian Federation); Saloutin, Viktor I.; Chupakhin, Oleg N. [I. Ya. Postovskii Institute of Organic Synthesis, Ural Branch, Russian Academy of Sciences, Kovalevskoy St., 22, Ekaterinburg 620990 (Russian Federation)

2014-08-15

Graphical abstract: - Highlights: • Quantum chemical calculations were carried out for PCBs congeners. • Calculated descriptors were used to explain the PCBs reactivity in S{sub N} and S{sub E} substitutions. • Obtained data were used to estimate the PCBs reactivity in the S{sub N} reactions. • Calculated descriptors were insufficient to explain the PCBs reactivity in the S{sub E} reactions. • New neutralization methods of the large-capacity PCBs were discussed. - Abstract: To explain the chemical reactivity of polychlorinated biphenyls in nucleophilic (S{sub N}) and electrophilic (S{sub E}) substitutions, quantum chemical calculations were carried out at the B3LYP/6-31G(d) level of the Density Functional Theory in gas phase. Carbon atomic charges in biphenyl structure were calculated by the Atoms-in-Molecules method. Chemical hardness and global electrophilicity index parameters were determined for congeners. A comparison of calculated descriptors and experimental data for congener reactivity in the S{sub N} and S{sub E} reactions was made. It is shown that interactions in the S{sub N} mechanism are reactions of the hard acid–hard base type, these are the most effective in case of highly chlorinated substrates. To explain the congener reactivity in the S{sub E} reactions, correct descriptors were not established. The obtained results can be used to carry out chemical transformations of the polychlorinated biphenyls in order to prepare them for microbiological destruction or preservation.

Beginning-of-life neutronic analysis of a 3000-MW(t) HTGR

International Nuclear Information System (INIS)

Vigil, J.C.

1975-12-01

The results of a study of safety-related neutronic characteristics for the beginning-of-life core of a 3000-MW(t) High-Temperature Gas-Cooled Reactor are presented. Emphasis was placed on the temperature-dependent reactivity effects of fuel, moderator, control poisons, and fission products. Other neutronic characteristics studied were gross and local power distributions, neutron kinetics parameters, control rod and other material worths and worth distributions, and the reactivity worth of a selected hypothetical perturbation in the core configuration. The study was performed for the most part using discrete-ordinates transport theory codes and neutron cross sections that were interpolated from a four-parameter nine-group library supplied by the HTGR vendor. A few comparison calculations were also performed using nine-group data generated with an independent cross-section processing code system. Results from the study generally agree well with results reported by the HTGR vendor

Measurement and analysis of reactivity temperature coefficient of CEFR

International Nuclear Information System (INIS)

Chen Yiyu; Hu Yun; Yang Xiaoyan; Fan Zhendong; Zhang Qiang; Zhao Jinkun; Li Zehua

2013-01-01

The reactivity temperature coefficient of CEFR was calculated by CITATION program and compared with the results calculated by correlative programs and measured from experiments for temperature effects. It is indicated that the calculation results from CITATION agree well with measured values. The reactivity temperature coefficient of CEFR is about -4 pcm/℃. The deviation of the measured values between the temperature increasing and decreasing processes is about 11%, which satisfies the experiment acceptance criteria. The measured results can validate the calculation ones by program and can provide important reference data for the safety operation of CEFR and the analysis of the reactivity balance in the reactor refueling situation. (authors)

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

Reactivity Accidents in CAREM-25 Core with and Without Safety Systems Actuation

International Nuclear Information System (INIS)

Gimenez, Marcelo; Vertullo, Alicia; Schlamp, Miguel

2000-01-01

A reactivity accident in CAREM core can be provoked by different initiating events, a cold water injection in pressure vessel, a secondary side steam line breakage and a failure in the absorbing rods drive system.The present work analyses inadverted control rod withdraws transients.Maximum worth control rod (2.5 $) at normal velocity (1 cm/s) is adopted for the simulations (Reactivity ramp of 0.018 $/s).Different scenarios considering actuation of first shutdown system (FSS), second shutdown system (SSS) and selflimiting conditions were modeled.Results of the accident with actuation of FSS show that safety margins are well above critical values (DNBR and CPR).In the cases with failure of the FSS and success of SSS or selflimited, safety margins are below critical values, however, the SSS provides a reduction of elapsed time under advised margins

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.

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.

Calculations of Changes in Reactivity during some regular periods of operation of JEN-1 MOD Reactor; Calculo de vairaciones de reactividad en algunos periodos regulares de operacion del reactor JEN-1 Mod.

Energy Technology Data Exchange (ETDEWEB)

Alcala Ruiz, F

1973-07-01

By a Point-Reactor model and Perturbation Theory, changes in reactivity during some regular operating periods of JEN-1 MOD Reactor have been calculated and compared with available measured values. they were in good agreement. Also changes in reactivity have been calculated during operations at higher power levels than the present one, concluding some practical consequences for the case of increasing the present power of this reactor. (Author)

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.

TREAT experimental data base regarding fuel dispersals in LMFBR loss-of-flow accidents

International Nuclear Information System (INIS)

Simms, R.; Fink, C.L.; Stanford, G.S.; Regis, J.P.

1981-01-01

The reactivity feedback from fuel relocation is a central issue in the analysis of loss-of-flow (LOF) accidents in LMFBRs. Fuel relocation has been studied in a number of LOF simulations in the TREAT reactor. In this paper the results of these tests are analyzed, using, as the principal figure of merit, the changes in equivalent fuel worth associated with the fuel motion. The equivalent fuel worth was calculated from the measured axial fuel distributions by weighting the data with a typical LMFBR fuel-worth function. At nominal power, the initial fuel relocation resulted in increases in equivalent fuel worth. Above nominal power the fuel motion was dispersive, but the dispersive driving forces could not unequivocally be identified from the experimental data

Reactivity Impact of Difference of Nuclear Data Library for PWR Fuel Assembly Calculation by Using AEGIS Code

International Nuclear Information System (INIS)

Ohoka, Yasunori; Tatsumi, Masahiro; Sugimura, Naoki; Tabuchi, Masato

2011-01-01

In 2010, the latest version of the Japanese Evaluated Nuclear Data Library (JENDL-4.0) has been released by JAEA. JENDL-4.0 is major update from JENDL- 3.3, and confirmed to give good accuracy by integral test for fission reactor systems such as fast neutron system and thermal neutron system. In this study, we evaluated the reactivity impact due to difference between ENDF/B-VII.0 and JENDL-4.0 for PWR fuel assembly burnup calculation using AEGIS code which has been developed by Nuclear Engineering, Ltd. in cooperation with Nuclear Fuel Industries, Ltd. and Nagoya University

Calculation of the fast multiplication factor by the fission matrix method

International Nuclear Information System (INIS)

Naumov, V.A.; Rozin, S.G.; Ehl'perin, T.I.

1976-01-01

A variation of the Monte Carlo method to calculate an effective breeding factor of a nuclear reactor is described. The evaluation procedure of reactivity perturbations by the Monte Carlo method in the first order perturbation theory is considered. The method consists in reducing an integral neutron transport equation to a set of linear algebraic equations. The coefficients of this set are elements of a fission matrix. The fission matrix being a Grin function of the neutron transport equation, is evaluated by the Monte Carlo method. In the program realizing the suggested algorithm, the game for initial neutron energy of a fission spectrum and then for the region of neutron birth, ΔVsub(f)sup(i)has been played in proportion to the product of Σsub(f)sup(i)ΔVsub(f)sup(i), where Σsub(f)sup(i) is a macroscopic cross section in the region numbered at the birth energy. Further iterations of a space distribution of neutrons in the system are performed by the generation method. In the adopted scheme of simulation of neutron histories the emission of secondary neutrons is controlled by weights; it occurs at every collision and not only in the end on the history. The breeding factor is calculated simultaneously with the space distribution of neutron worth in the system relative to the fission process and neutron flux. Efficiency of the described procedure has been tested on the calculation of the breeding factor for the Godiva assembly, simulating a fast reactor with a hard spectrum. A high accuracy of calculations at moderate number of zones in the core and reasonable statistics has been stated

From parallel to distributed computing for reactive scattering calculations

International Nuclear Information System (INIS)

Lagana, A.; Gervasi, O.; Baraglia, R.

1994-01-01

Some reactive scattering codes have been ported on different innovative computer architectures ranging from massively parallel machines to clustered workstations. The porting has required a drastic restructuring of the codes to single out computationally decoupled cpu intensive subsections. The suitability of different theoretical approaches for parallel and distributed computing restructuring is discussed and the efficiency of related algorithms evaluated

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...

Startup physics tests at Temelin NPP, Unit 1

International Nuclear Information System (INIS)

Sedlacek, M.; Minarcin, M.; Toth, L.; Elko, M.; Hascik, R.

2002-01-01

The objective, scope and proceedings of the physics tests of Temelin NPP, Unit 1 physical commissioning are given in this paper. Furthermore, some results of selected physics tests are presented: reactor initial criticality test, determination of reactor power range for physics testing, measurement of control rod cluster assembly group no. 10 reactivity worth in case of limitation system LS(a) actuation, control rod cluster assembly system reactivity worth measurement with single rod cluster assembly of greatest reactivity worth stuck in fully withdrawn position, measurement of differential reactivity worth of control rod cluster assembly group no. 9, boron 'endpoint' determination and measurement of power reactivity coefficient (Authors)

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…

Measurements of negative reactivity in Masurca and Phenix control rods: Prospects for Superphenix

International Nuclear Information System (INIS)

Gauthier, J.C.; Petiot, R.; Coulon, P.; Giese, H.; West, J.P.

1986-01-01

Experimental assessment of the negative reactivity of the control rods in an industrial reactor has recently been the subject of numerous studies conducted in the light of forthcoming startup tests on the core of Superphenix. Representative tests have been carried out both on Phenix and on the Masurca critical mockup, and a test programme for Superphenix has been drawn up. Subcritical measurements (source multiplication technique) have been carried out on Phenix without absolute measurement of a standard. However, a precise relative interpretation using two counters demonstrates good agreement following the correction of spatial effects. The chief value of the rod drop measurements conducted on Masurca was that it provided a means of cross-checking the kinetic method to be validated against a standard source multiplication method. The results demonstrate complete agreement between the two methods. The acceptability of the rod drop method is therefore considered to be established. The programme foreseen for startup of Superphenix and the objectives which have been set are briefly indicated. The calculation methods to be used in respect of the startup tests have been established on the basis of experience gained through interpreting the experiments conducted in the course of the Racine (Masurca) programme. An analysis of these experiments included, among other things, a parametric study that has made it possible to devise a standard calculation method for predicting Superphenix rod worth values. The main feature is a scattering calculation with three energy groups and three dimensions. Two-dimensional scattering and transport calculations are therefore necessary in order to define the corrective factors to be applied to this initial result. The final result of this analysis is thus made equivalent to a 25-energy-group transport calculation with an extremely small spatial mesh

A simple reactivity-meter system

International Nuclear Information System (INIS)

Ferreira, P.S.B.

1992-01-01

This paper describes a new version of a reactivity meter developed at the Institute of Nuclear Energy Research (IPEN) (Brazil). The reactivity meter computes the reactor reactivity utilizing a programmable electrometer that performs the data aquisition. The software commands the main functions of the electrometer, the data acquisition, data transfer, and reactivity calculation. The necessary hardware for this reactivity meter are a programmable electrometer, a microcomputer, and interfaces for the microcomputer to communicate with the electrometer. If it is necessary, it is possible to connect a graphic register to the microcomputer. With this conventional hardware, available in any nuclear reactor facility, one can build a powerful reactivity meter. Adding to these advantages, one can use the microcomputer on-line to analyze the data, store the data on diskettes, or create graphics

Research of three-dimensional transient reactivity feedback in fast reactor

International Nuclear Information System (INIS)

Xu Li; Shi Gong; Ma Dayuan; Yu Hong

2013-01-01

To solve the three-dimensional time-spatial kinetics feedback problems in fast reactor, a mathematical model of the direct reactivity feedback was proposed. Based on the NAS code for fast reactor and the reactivity feedback mechanism, a feedback model which combined the direct reactivity feedback and feedback reflected by the cross section variation was provided for the transient calculation. Furthermore, the fast reactor group collapsing system was added to the code, thus the real time group collapsing calculation could be realized. The isothermal elevated temperature test of CEFR was simulated by using the code. By comparing the calculation result with the test result of the temperature reactivity coefficient, the validity of the model and the code is verified. (authors)

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.

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...

The fourth research co-ordination meeting (RCM) on 'Updated codes and methods to reduce the calculational uncertainties of liquid metal fast reactors reactivity effects'. Working material

International Nuclear Information System (INIS)

2003-01-01

The fourth Research Co-ordination Meeting (RCM) of the Co-ordinated Research Project (CRP) on 'Updated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effect' was held during 19-23 May, 2003 in Obninsk, Russian Federation. The general objective of the CRP is to validate, verify and improve methodologies and computer codes used for the calculation of reactivity coefficients in fast reactors aiming at enhancing the utilization of plutonium and minor actinides. The first RCM took place in Vienna on 24 - 26 November 1999. The meeting was attended by 19 participants from 7 Member States and one from an international organization (France, Germany, India, Japan, Rep. of Korea, Russian Federation, the United Kingdom, and IAEA). The participants from two Member States (China and the U.S.A.) provided their results and presentation materials even though being absent at the meeting. The results for several relevant reactivity parameters obtained by the participants with their own state-of-the-art basic data and codes, were compared in terms of calculational uncertainty, and their effects on the ULOF transient behavior of the hybrid BN- 600 core were evaluated. Contributions of the participants in the benchmark analyses is shown. This report first addresses the benchmark definitions and specifications given for each Phase and briefly introduces the basic data, computer codes, and methodologies applied to the benchmark analyses by various participants. Then, the results obtained by the participants in terms of calculational uncertainty and their effect on the core transient behavior are intercompared. Finally it addresses some conclusions drawn in the benchmarks

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.

Development of a UNIX network compatible reactivity computer

International Nuclear Information System (INIS)

Sanchez, R.F.; Edwards, R.M.

1996-01-01

A state-of-the-art UNIX network compatible controller and UNIX host workstation with MATLAB/SIMULINK software were used to develop, implement, and validate a digital reactivity calculation. An objective of the development was to determine why a Macintosh-based reactivity computer reactivity output drifted intolerably

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…

Study of reactivity of fluidized bed nuclear reactor

International Nuclear Information System (INIS)

Rammsy, J.E.M.

1985-01-01

The reactor physics calculations of a 19 module Fluidized Bed Nuclear Reactor using Leopard and Odog codes are performed. The behaviour of the reactor was studied by calculating the reactivity of the reactor as a function of the parameters governing the operational and accidental conditions of the reactor. The effects of temperature, pressure, and vapor generation in the core on the reactivity are calculated. Also the start up behaviour of the reactor is analyzed. For the purpose of the study of a prototype research reactor, the calculations on a one module reactor have been performed. (Author) [pt

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.

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...

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...

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.

RP-10: commissioning. Reproduction by physical experiences calculation

International Nuclear Information System (INIS)

Higa, Manabu; Madariaga, M.R.

1990-01-01

This work presents the neutronic calculation results, most of which were carried out after such experiences, to verify the calculation methodology developed at the Analysis and Calculation Department of the National Atomic Energy Commission (CNEA). The results obtained were satisfactory, proving that the calculation methodology used is adequate for the design of this type of reactors. The only important disagreement is to evaluate the reactivity excess and cut reactivity, but this responds to a criterion difference and/or that of definition for these parameters. The positions of criticality with errors lower than 100 pcm were predicted. The differential and integral reactivities for the calibration of bars, as well as the flux distribution, are reproduced in a reasonable degree in relation to differences inferior to 10%. (Author) [es

Structure, Reactivity and Dynamics

Indian Academy of Sciences (India)

Understanding structure, reactivity and dynamics is the core issue in chemical ... functional theory (DFT) calculations, molecular dynamics (MD) simulations, light- ... between water and protein oxygen atoms, the superionic conductors which ...

KfK analysis of the SUPER-PHENIX-1 control rod experiments. Pt. 1

International Nuclear Information System (INIS)

Giese, H.

1991-03-01

As proposed by the SPX-1 analysis task force, MSM (modified source multiplication) correction factors have been produced for a series of control rod configurations established in the first critical core C1D with minimum fissile loading and in the fully loaded core CMP. The report gives a complete description of the method used at KfK to produce these correction factors and summarises the evaluated experimental results obtained. The KfK method is characterized by a 'two-step-adjustment': A basic reactivity scale adjustment and a subsequent rod worth adjustment. The first adjustment was achieved by 'tuning' either the axialbuckling in the leakage term D B 2 or the average number of neutrons per fission in the production term so that the excess reactivity of the so-called 'Follower-core' with all control rods fully raised was properly reproduced. As this excess reactivity could not be directly determined by an experiment, it had to be assessed from the shut-down worth of the main control system in combination with measured fractions of the S-curve of this system. In the second adjustment, the absorber cross sections were tuned to reproduce experimental rod worths. While for the analysis of the C1D experiments, MSM correction factor calculations were performed in 2D centre-plane geometry only, the analysis of the CMP measurements employed both 2D and 3D calculations. (orig./HP)

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.

Sensitivity and uncertainty analysis of reactivities for UO2 and MOX fueled PWR cells

Energy Technology Data Exchange (ETDEWEB)

Foad, Basma [Research Institute of Nuclear Engineering, University of Fukui, Kanawa-cho 1-2-4, Tsuruga-shi, Fukui-ken, 914-0055 (Japan); Egypt Nuclear and Radiological Regulatory Authority, 3 Ahmad El Zomar St., Nasr City, Cairo, 11787 (Egypt); Takeda, Toshikazu [Research Institute of Nuclear Engineering, University of Fukui, Kanawa-cho 1-2-4, Tsuruga-shi, Fukui-ken, 914-0055 (Japan)

2015-12-31

The purpose of this paper is to apply our improved method for calculating sensitivities and uncertainties of reactivity responses for UO{sub 2} and MOX fueled pressurized water reactor cells. The improved method has been used to calculate sensitivity coefficients relative to infinite dilution cross-sections, where the self-shielding effect is taken into account. Two types of reactivities are considered: Doppler reactivity and coolant void reactivity, for each type of reactivity, the sensitivities are calculated for small and large perturbations. The results have demonstrated that the reactivity responses have larger relative uncertainty than eigenvalue responses. In addition, the uncertainty of coolant void reactivity is much greater than Doppler reactivity especially for large perturbations. The sensitivity coefficients and uncertainties of both reactivities were verified by comparing with SCALE code results using ENDF/B-VII library and good agreements have been found.

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…

Bubble and Dew Point Calculations in Multicomponent and Multireactive Mixtures

OpenAIRE

Bonilla-Petriciolet, A.; Acosta-Martínez, A.; Bravo-Sánchez, U. I.; Segovia-Hernández, J. G.

2006-01-01

Bubble and dew point calculations are useful in chemical engineering and play an important role in the study of separation equipments for non-reactive and reactive mixtures. To the best of the authors’s knowledge, few methods have been proposed for these calculations in systems with several chemical reactions. The objective of this paper is to introduce new conditions for performing bubble and dew point calculations in reactive mixtures. We have developed these conditions based on the a...

Analysis of MOZART critical experiment using the IRPhEP handbook data

International Nuclear Information System (INIS)

Chiba, Go

2010-12-01

Using the experimental data described in the IRPhEP handbook, an experimental analysis of the MOZART experiment is carried out with the nuclear data JENDL-4.0, and the reactor physics codes SLAROM-UF and CBG. The following results are obtained: -The C/E values for criticality are 0.9981 for the small-sized core MZA and 1.0006 for the middle-sized core MZB. Good agreement between calculation and experimental values has been observed similarly in the analyses for criticality of other MOX-fueled fast reactors. Hence, consistency between the present analysis and the others is confirmed. -In reaction rate ratios at the core center, calculation values agree with experimental values within 1.0% for F25/F49 and C28/F49, and within 4.0% for F28/F25, F40/F49 and F41/F49. -In sodium void reactivity worths, calculation values are about 10% larger than experimental values for the non-leakage-dominated data. For the data to which the leakage component largely contributes, absolute differences normalized by the leakage component are less than 10%. -In material worths, calculation values are about 5% larger than experimental values for plutonium. Calculation values agree with experimental values within 10% differences for uranium and SS. -In control rod worths, calculation values are 2% to 5% larger than experimental values. -In reaction rate distributions, calculation values agree well with experimental values in core regions. On the other hand, underestimation is observed systematically in calculation values of threshold reactions in blanket regions. For the reactivity characteristic, overestimation is systematically observed in calculations. While the reason has not yet been investigated, the result suggests underestimation of β eff . (author)

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.

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.

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.

Quantum mechanical calculations of vibrational population inversion in chemical reactions - Numerically exact L-squared-amplitude-density study of the H2Br reactive system

Science.gov (United States)

Zhang, Y. C.; Zhang, J. Z. H.; Kouri, D. J.; Haug, K.; Schwenke, D. W.

1988-01-01

Numerically exact, fully three-dimensional quantum mechanicl reactive scattering calculations are reported for the H2Br system. Both the exchange (H + H-prime Br to H-prime + HBr) and abstraction (H + HBR to H2 + Br) reaction channels are included in the calculations. The present results are the first completely converged three-dimensional quantum calculations for a system involving a highly exoergic reaction channel (the abstraction process). It is found that the production of vibrationally hot H2 in the abstraction reaction, and hence the extent of population inversion in the products, is a sensitive function of initial HBr rotational state and collision energy.

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.

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…

Point kinetics improvements to evaluate three-dimensional effects in transients calculation

International Nuclear Information System (INIS)

Castellotti, U.

1987-01-01

A calculation method, which considers the flux axial perturbations in the parameters related to the reactivity within a point kinetics model, is described. The method considered uses axial factors of consideration which act on the thermohydraulic variables included in the reactivity calculation. The PUMA three-dimensional code as reference model for the comparisons, is used. The limitations inherent to the reactivity balance of the point models used in the transients calculation, are given. (Author)

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...

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.

Critical mass, rod values and reactivity coefficients for Rapsodie; Masse critique, valeur des barres et coefficients de reactivite de rapsodie

Energy Technology Data Exchange (ETDEWEB)

Stevens, L; Gourdon, J [Commissariat a l' Energie Atomique, Cadarache (France). Centre d' Etudes Nucleaires

1967-07-01

Besides a brief general description, the report contains a description and discussion of the aims, the methods used and the results of critical mass, rod worth and static reactivity coefficient measurements on the Rapsodie reactor. (authors) [French] Apres une breve description generale, le rapport decrit et discute le but, les methodes employees et les resultats des mesures de masse critique, de reactivite des barres et des coefficients de reactivite statiques du reacteur RAPSODIE. (auteurs)

Analysis of the Techniques for Measuring the Reactivity of Far Sub-Critical Multiplying Systems

International Nuclear Information System (INIS)

Vandeplas, P.

1968-01-01

The methods of measuring control-rod worth on the basis of reactor response to a periodic excitation source are analysed and it is shown that the ratio of the integrals of the prompt and delayed neutron densities over one period is independent of the shape of the signal used. This ratio will thus be equal to the ratio of the neutron densities in the presence of a time-independent source. The pulsed source, sinusoidal source and source removed methods therefore give identical worth values. Experimentation and numerical analysis show that the ratio of prompt to delayed neutron density, which is a linear combination of successive eigenvalues of the system, is largely dependent on the position of the source. It is demonstrated analytically that with a suitably chosen volume of integration, the contribution of the space harmonics is drastically reduced in the integrals of both the prompt neutron density and the delayed neutron density. Numerical studies show that the ratio of these two integrals is practically equal to the reactivity corresponding to the fundamental static mode and is almost independent of the position of the source. Integration of the neutron densities over a volume can be reduced to an integration along a straight line by a suitable choice of source location. Such an integration can be performed by means of an integral detector, thereby only requiring one measurement. Optimization of the position of the source and of the limits of integration to give the best value for the reactivity can be achieved on the basis of purely experimental criteria. (author) [fr

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

Fast Neutron Spectrum Potassium Worth for Space Power Reactor Design Validation

Energy Technology Data Exchange (ETDEWEB)

Bess, John D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Marshall, Margaret A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Briggs, J. Blair [Idaho National Lab. (INL), Idaho Falls, ID (United States); Tsiboulia, Anatoli [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rozhikhin, Yevgeniy [Idaho National Lab. (INL), Idaho Falls, ID (United States); Mihalczo, John T. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-03-01

graphite reflected (2 inches or less) experiments also using the same set of highly enriched uranium metal parts are evaluated in HEU MET FAST 071. Polyethylene-reflected configurations are evaluated in HEU-MET-FAST-076. A stack of highly enriched metal discs with a thick beryllium top reflector is evaluated in HEU-MET-FAST-069, and two additional highly enriched uranium annuli with beryllium cores are evaluated in HEU-MET-FAST-059. Both detailed and simplified model specifications are provided in this evaluation. Both of these fast neutron spectra assemblies were determined to be acceptable benchmark experiments. The calculated eigenvalues for both the detailed and the simple benchmark models are within ~0.26 % of the benchmark values for Configuration 1 (calculations performed using MCNP6 with ENDF/B-VII.1 neutron cross section data), but under-calculate the benchmark values by ~7s because the uncertainty in the benchmark is very small: ~0.0004 (1s); for Configuration 2, the under-calculation is ~0.31 % and ~8s. Comparison of detailed and simple model calculations for the potassium worth measurement and potassium mass coefficient yield results approximately 70 – 80 % lower (~6s to 10s) than the benchmark values for the various nuclear data libraries utilized. Both the potassium worth and mass coefficient are also deemed to be acceptable benchmark experiment measurements.

Dependence of calculated void reactivity on film boiling representation in a CANDU lattice

Energy Technology Data Exchange (ETDEWEB)

Whitlock, J [McMaster Univ., Hamilton, ON (Canada). Dept. of Engineering Physics

1994-12-31

The distribution dependence of void reactivity in a CANDU (CANada Deuterium Uranium) lattice is studied, specifically in the regime of film boiling. A heterogeneous model of this phenomenon predicts a 4% increase in void reactivity over a homogeneous model for fresh fuel, and 11% at discharge. An explanation for this difference is offered, with regard to differing changes in neutron mean free path upon voiding. (author). 9 refs., 4 tabs., 6 figs.

Reactivity margins in heavy water moderated production reactors

International Nuclear Information System (INIS)

Benton, F.D.

1981-11-01

The design of the reactor core and components of the heavy water moderated reactors at the Savannah River Plant (SFP) can be varied to produce a number of isotopes. For the past decade, the predominant reactor core design has been the enriched-depleted lattice. In this lattice, fuel assemblies of highly enriched uranium and target assemblies of depleted uranium, which produce plutonium, occupy alternate lattice positions. This heterogeneous lattice arrangement and a nonuniform control rod distribution result in a reactor core that requires sophisticated calculational methods for accurate reactivity margin and power distribution predictions. For maximum accuracy, techniques must exist to provide a base of observed data for the calculations. Frequent enriched-depleted lattice design changes are required as product demands vary. These changes provided incentive for the development of techniques to combine the results of calculations and observed reactivity data to accurately and conveniently monitor reactivity margins during operation

Accuracy of WWR-M criticality calculations with code MCU-RFFI

International Nuclear Information System (INIS)

Petrov, Yu.V.; Erykalov, A.N.; Onegin, M.S.

1999-01-01

The scattering and deviation of fuel element parameters by manufacturing, approximations of the reactor structure in the computer model, the partly inadequate neutron cross sections in the computer codes etc. lead to a discrepancy between the reactivity computations and data. We have compared reactivity calculations using the MCU-RRFI Monte Carlo code of critical assemblies containing WWR-M2 (36 enriched) an WWR-M5 (90%) fuel elements with benchmark experiments. The agreement was about Δρ≅±0.3%. A strong influence of the water ratio on reactivity was shown and a significant heterogeneous effect was found. We have also investigated, by full scale reactor calculations for the RETR program, the contribution to the reactivity of the main reactor structure elements: beryllium reflector, experimental channels irradiation devices inside the core, etc. Calculations show the importance of a more thorough study of the contributions of products of the (n, α) reaction in the Be reflector to the reactivity. Ways of improving the accuracy of the calculations are discussed. (author)

Accuracy of WWR-M criticality calculations with code MCU-RFFI

Energy Technology Data Exchange (ETDEWEB)

Petrov, Yu V [Petersburg Nuclear Physics Institute RAS, 188350 Gatchina, St. Petersburg (Russian Federation); Erykalov, A N; Onegin, M S [Petersburg Nuclear Physics Institute RAS, 188350 Gatchina, St. Petersburg (Russian Federation)

1999-10-01

The scattering and deviation of fuel element parameters by manufacturing, approximations of the reactor structure in the computer model, the partly inadequate neutron cross sections in the computer codes etc. lead to a discrepancy between the reactivity computations and data. We have compared reactivity calculations using the MCU-RRFI Monte Carlo code of critical assemblies containing WWR-M2 (36 enriched) an WWR-M5 (90%) fuel elements with benchmark experiments. The agreement was about {delta}{rho}{approx_equal}{+-}0.3%. A strong influence of the water ratio on reactivity was shown and a significant heterogeneous effect was found. We have also investigated, by full scale reactor calculations for the RETR program, the contribution to the reactivity of the main reactor structure elements: beryllium reflector, experimental channels irradiation devices inside the core, etc. Calculations show the importance of a more thorough study of the contributions of products of the (n, {alpha}) reaction in the Be reflector to the reactivity. Ways of improving the accuracy of the calculations are discussed. (author)

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).

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

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.

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.

Prediction uncertainty and data worth assessment for groundwater transport times in an agricultural catchment

Science.gov (United States)

Zell, Wesley O.; Culver, Teresa B.; Sanford, Ward E.

2018-06-01

Uncertainties about the age of base-flow discharge can have serious implications for the management of degraded environmental systems where subsurface pathways, and the ongoing release of pollutants that accumulated in the subsurface during past decades, dominate the water quality signal. Numerical groundwater models may be used to estimate groundwater return times and base-flow ages and thus predict the time required for stakeholders to see the results of improved agricultural management practices. However, the uncertainty inherent in the relationship between (i) the observations of atmospherically-derived tracers that are required to calibrate such models and (ii) the predictions of system age that the observations inform have not been investigated. For example, few if any studies have assessed the uncertainty of numerically-simulated system ages or evaluated the uncertainty reductions that may result from the expense of collecting additional subsurface tracer data. In this study we combine numerical flow and transport modeling of atmospherically-derived tracers with prediction uncertainty methods to accomplish four objectives. First, we show the relative importance of head, discharge, and tracer information for characterizing response times in a uniquely data rich catchment that includes 266 age-tracer measurements (SF6, CFCs, and 3H) in addition to long term monitoring of water levels and stream discharge. Second, we calculate uncertainty intervals for model-simulated base-flow ages using both linear and non-linear methods, and find that the prediction sensitivity vector used by linear first-order second-moment methods results in much larger uncertainties than non-linear Monte Carlo methods operating on the same parameter uncertainty. Third, by combining prediction uncertainty analysis with multiple models of the system, we show that data-worth calculations and monitoring network design are sensitive to variations in the amount of water leaving the system via

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…

Reduction of fluctuations in reactivity using symmetric matrix

International Nuclear Information System (INIS)

Suescun D, D.; Segovia Ch, F. A.; Bonilla L, H. F.

2016-09-01

A new filtering method is presented in this work known as Savitzky-Golay (SG); allows reducing the fluctuations in the calculation of the reactivity. The filter softens and does not attenuate the nuclear power regardless of its shape, guaranteeing to decrease different degrees of noise with different steps of calculation time. This formulation employs a polynomial approximation of a certain degree to calculate the convolution coefficients. Its implementation is computational and avoids problems of bad conditioning, caused by the inversion of a linear system. The results show values in the maximum difference and in the averages absolute errors of the reactivity in comparison with that reported in the literature. (Author)

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, ...

Experimental Studies on Assemblies 1 and 2 of the Fast Reactor FR-0. Part 2

Energy Technology Data Exchange (ETDEWEB)

Hellstrand, E; Andersson, T L; Brunfelter, B; Kockum, J; Londen, S O; Tiren, L I

1965-12-15

In a first part of this report, published as AE-195, an account was given of critical mass determinations and measurements of flux distribution and reaction ratios in the first assemblies of the fast zero power reactor FR0. This second part of the report deals with various investigations involving the measurement of reactivity. Control rod calibrations have been made using the positive period, the inverse multiplication, the rod drop and the pulsed source techniques, and show satisfactory agreement between the various methods. The reactivity worths of samples of different materials and different sizes have been measured at the core centre. Comparisons with perturbation calculations show that the regular and adjoint fluxes are well predicted in the central region of the core. The variation in the prompt neutron life-time with reactivity has been studied by means of the pulsed source and the Rossi-{alpha} techniques. Comparison with one region calculations reveals large discrepancies, indicating that this simple model is inadequate. Some investigations of streaming effects in an empty channel in the reactor and of interaction effects between channels have been made and are compared with theoretical estimates. Measurements of the reactivity worth of an air gap between the reactor halves and of the temperature coefficient are also described in the report. The work has been performed as a joint effort by AB Atomenergi and the Research Institute of National Defence.

The impact of fuel temperature reactivity coefficient on loss of reactivity control accident

International Nuclear Information System (INIS)

Park, J. H.; Ryu, E. H.; Song, Y. M.; Jung, J. Y.

2012-01-01

Nuclear reactors experience small power fluctuations or anticipated operational transients during even normal power operation. During normal operation, the reactivity is mainly controlled by liquid zone controllers, adjuster rods, mechanical control absorbers, and moderator poison. Even when the reactor power is increased abruptly and largely from an accident and when reactor control systems cannot be actuated quickly due to a fast transient, the reactor should be controlled and stabilized by its inherent safety parameter, such as a negative PCR (Power Coefficient of Reactivity) feedback. A PWR (Pressurized Water Reactor), it is well designed for the reactor to have a negative PCR so that the reactor can be safely shut down or stabilized whenever an abrupt reactivity insertion into the reactor core occurs or the reactor power is abruptly increased. However, it is known that a CANDU reactor has a small amount of PCR, as either negative or positive, because of the different design basis and safety concepts from a PWR. CNSC's regulatory and safety regime has stated that; The PCR of CANDU reactors does not pose a significant risk. Consistent with Canadian nuclear safety requirements, nuclear power plants must have an appropriate combination of inherent and engineered safety features incorporated into the design of the reactor safety and control systems. A reactor design that has a PCR is quite acceptable provided that the reactor is stable against power fluctuations, and that the probability and consequences of any potential accidents that would be aggravated by a positive reactivity feedback are maintained within CNSCprescribed limits. Recently, it was issued licensing the refurbished Wolsong unit 1 in Korea to be operated continuously after its design lifetime in which the calculated PCR was shown to have a small positive value by applying the recent physics code systems, which are composed of WIMS IST, DRAGON IST, and RFSP IST. These code systems were transferred

CRITICALITY CALCULATION FOR THE MOST REACTIVE DEGRADED CONFIGURATIONS OF THE FFTF SNF CODISPOSAL WP CONTAINING AN INTACT IDENT-69 CONTAINER

International Nuclear Information System (INIS)

D.R. Moscalu

2002-01-01

The objective of this calculation is to perform additional degraded mode criticality evaluations of the Department of Energy's (DOE) Fast Flux Test Facility (FFTF) Spent Nuclear Fuel (SNF) codisposed in a 5-Defense High-Level Waste (5-DHLW) Waste Package (WP). The scope of this calculation is limited to the most reactive degraded configurations of the codisposal WP with an almost intact Ident-69 container (breached and flooded but otherwise non-degraded) containing intact FFTF SNF pins. The configurations have been identified in a previous analysis (CRWMS M andO 1999a) and the present evaluations include additional relevant information that was left out of the original calculations. The additional information describes the exact distribution of fissile material in each container (DOE 2002a). The effects of the changes that have been included in the baseline design of the codisposal WP (CRWMS M andO 2000) are also investigated. The calculation determines the effective neutron multiplication factor (k eff ) for selected degraded mode internal configurations of the codisposal waste package. These calculations will support the demonstration of the technical viability of the design solution adopted for disposing of MOX (FFTF) spent nuclear fuel in the potential repository. This calculation is subject to the Quality Assurance Requirements and Description (QARD) (DOE 2002b) per the activity evaluation under work package number P6212310M2 in the technical work plan TWP-MGR-MD-0000101 (BSC 2002)

Total OH Reactivity Measurements in the Boreal Forest

Science.gov (United States)

Praplan, A. P.; Hellén, H.; Hakola, H.; Hatakka, J.

2015-12-01

INTRODUCTION Atmospheric total OH reactivity (Rtotal) can be measured (Kovacs and Brune, 2001; Sinha et al., 2008) or it can be calculated according to Rtotal = ∑i kOH+X_i [Xi] where kOH+X_i corresponds to the reaction rate coefficient for the reaction of OH with a given compound Xi and [Xi] its concentration. Studies suggest that in some environments a large fraction of missing reactivity, comparing calculated Rtotal with ambient total OH reactivity measurements (Di Carlo et al., 2004; Hofzumahaus et al., 2009). In this study Rtotal has been measured using the Comparative Reactivity Method (Sinha et al., 2008). Levels of the reference compound (pyrrole, C4H5N) are monitored by gas chromatography every 2 minutes and Rtotal is derived from the difference of reactivity between zero and ambient air. RESULTS Around 36 hours of preliminary total OH reactivity data (30 May until 2 June 2015) are presented in Fig. 1. Its range matches previous studies for this site (Nölscher et al., 2012; Sinha et al., 2010) and is similar to values in another pine forest (Nakashima et al., 2014). The setup used during the period presented here has been updated and more recent data will be presented, as well as a comparison with calculated OH reactivity from measured individual species. ACKNOWLEDGEMENTS This work was supported by Academy of Finland (Academy Research Fellowship No. 275608). The authors acknowledge Juuso Raine for technical support. REFERENCES Di Carlo et al. (2004). Science 304, 722-725.Hofzumahaus et al. (2009). Science 324, 1702-1704.Kovacs and Brune (2001). J. Atmos. Chem. 39, 105-122.Nakashima et al. (2014). Atmos. Env. 85, 1-8.Nölscher et al. (2012). Atmos. Chem. Phys. 12, 8257-8270.Sinha et al. (2008). Atmos. Chem. Phys. 8, 2213-2227.Sinha et al. (2010). Environ. Sci. Technol. 44, 6614-6620.

Parallel computation of multigroup reactivity coefficient using iterative method

Science.gov (United States)

Susmikanti, Mike; Dewayatna, Winter

2013-09-01

One of the research activities to support the commercial radioisotope production program is a safety research target irradiation FPM (Fission Product Molybdenum). FPM targets form a tube made of stainless steel in which the nuclear degrees of superimposed high-enriched uranium. FPM irradiation tube is intended to obtain fission. The fission material widely used in the form of kits in the world of nuclear medicine. Irradiation FPM tube reactor core would interfere with performance. One of the disorders comes from changes in flux or reactivity. It is necessary to study a method for calculating safety terrace ongoing configuration changes during the life of the reactor, making the code faster became an absolute necessity. Neutron safety margin for the research reactor can be reused without modification to the calculation of the reactivity of the reactor, so that is an advantage of using perturbation method. The criticality and flux in multigroup diffusion model was calculate at various irradiation positions in some uranium content. This model has a complex computation. Several parallel algorithms with iterative method have been developed for the sparse and big matrix solution. The Black-Red Gauss Seidel Iteration and the power iteration parallel method can be used to solve multigroup diffusion equation system and calculated the criticality and reactivity coeficient. This research was developed code for reactivity calculation which used one of safety analysis with parallel processing. It can be done more quickly and efficiently by utilizing the parallel processing in the multicore computer. This code was applied for the safety limits calculation of irradiated targets FPM with increment Uranium.

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

Analysis Of Temperature Effects On Reactivity Of The Rsg-Gas Core Using Silicide Fuels

International Nuclear Information System (INIS)

Surbakti, Tukiran; Pinem, Surian

2001-01-01

RSG-GAS has been operating using new silicide fuels so that it is necessary to estimate and to measure the effect of temperature on reactivity of the core. The parameters to be determined due to temperature effect are reactivity coefficient of moderator temperature, temperature coefficient of fuel element and power reactivity coefficient. By doing a couple compensation method, determination of reactivity coefficient as well as the reactivity coefficient of moderator temperature can be obtained. Furthermore, coefficient of the reactivity was successfully estimated using the combination of WIMS-D4 and Batan-2DIFF. The cell calculation was done by using WIMS-D4 code to get macroscopic cross section and Batan-2DIFF code is used for core calculation. The calculation and experimental results of reactivity coefficient do not show any deviation from RSG-GAS safety margin. The results are -2,84 sen/ o C, -1,29 sen/MW and -0,64 sen/ o C for reactivity coefficients of temperature, power, fuel element and moderator temperature, respectively. All of 3 parameters are absolutely met with safety criteria

Reactivity effects due to beryllium poisoning of BR2

International Nuclear Information System (INIS)

Kalcheva, S.; Ponsard, B.; Koonen, E.

2004-01-01

This paper illustrates the impact of the poisoning of the beryllium reflector on reactivity variations of the Belgian MTR BR2 in SCK.CEN. Detailed calculations by MCNP-4C of reactivity effects caused by strong neutron absorbers 3 He and 6 Li during reactor operation history are presented. The importance of beryllium poisoning for the accuracy of reactivity predictions is discussed. (authors)

Application of the low disturbances theory in operation calculations of the BMK reactor

International Nuclear Information System (INIS)

Isaev, N.V.; Shmonin, Yu.V.; Pogosbekyan, L.R.; Druzhinin, V.E.

1985-01-01

Calculation algorithm of direct and contingent tasks in a two-group diffusion approximation for RBMK-1000 of Smolensk-1 nuclear power plant is presented. Examples of numeric calculation of the reactivity change caused by neutron field disturbance reactivity effect in case of refueling, refueling, rate and reactivity reserve on control rods are given. Calculations are made according to PEPO-program. The program is written in FORTRAN-4 for ES computer. The modificated low disturbances theory used in this program allows to reduce sufficiently the calculation error

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.

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.

Monte Carlo technique for local perturbations in multiplying systems

International Nuclear Information System (INIS)

Bernnat, W.

1974-01-01

The use of the Monte Carlo method for the calculation of reactivity perturbations in multiplying systems due to changes in geometry or composition requires a correlated sampling technique to make such calculations economical or in the case of very small perturbations even feasible. The technique discussed here is suitable for local perturbations. Very small perturbation regions will be treated by an adjoint mode. The perturbation of the source distribution due to the changed system and its reaction on the reactivity worth or other values of interest is taken into account by a fission matrix method. The formulation of the method and its application are discussed. 10 references. (U.S.)

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.

Development of an alternative reactivity meter for nuclear reactor control

International Nuclear Information System (INIS)

Ferreira, P.S.B.

1991-01-01

This work describes an alternative version of the IPEN-CNEN/SP reactivity-meter. This new version utilizes a programmable electrometer (to realize the data acquisition) and a IBM-PC microcomputer to process the reactivity calculation. The aim of development of this alternative reactivity-meter is to have available a equipment of measurements of reactivity in the case of the later version show any problem during an experiment. (author)

Reactive Hypertrophy of an Accessory Spleen Mimicking Tumour Recurrence of Metastatic Renal Cell Carcinoma

Directory of Open Access Journals (Sweden)

Christin Tjaden

2011-01-01

Full Text Available De novo occurrence of an accessory spleen after splenectomy is worth noting for two reasons. First, it is known that splenectomy can cause reactive hypertrophy of initially inactive and macroscopically invisible splenic tissue. Second, it can mimic tumour recurrence in situations in which splenectomy has been performed for oncological reasons. This might cause difficulties in differential diagnosis and the clinical decision for reoperation. We report the case of a patient with suspected recurrence of renal cell carcinoma after total pancreatectomy and splenectomy for metastatic renal cell carcinoma, which finally revealed an accessory spleen as the morphological correlate of the newly diagnosed mass in the left retroperitoneum.

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...

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.)

Assessment of reactivity devices for CANDU-6 with DUPIC fuel

International Nuclear Information System (INIS)

Jeong, Chang Joon; Choi, Hang Bok

1998-01-01

Reactivity device characteristics for a CANDU-6 reactor loaded with DUPIC fuel have been assessed. A transport code WIMS-AECL and a three-dimensional diffusion code RFSP were used for the lattice parameter generation and the core calculation, respectively. Three major reactivity devices have been assessed for their inherent functions. For the zone controller system, damping capability for spatial oscillation was investigated. The restart capability of the adjuster system was investigated. The shim operation and power stepback calculation were also performed to confirm the compatibility of the current adjuster rod system. The mechanical control absorber was assessed for the capability to compensate the temperature reactivity feedback following a power reduction. This study has shown that the current reactivity device systems retain their functions when used in a DUPIC fuel CANDU reactor

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.

Exploring the relationship between appearance-contingent self-worth and self-esteem: The roles of self-objectification and appearance anxiety.

Science.gov (United States)

Adams, Katherine E; Tyler, James M; Calogero, Rachel; Lee, Jenifer

2017-12-01

Previous work has shown that both an appearance-contingent self-worth (i.e., staking one's overall self-evaluation on one's physical appearance) and self-objectification are associated with higher appearance anxiety and lower self-esteem among women. Although prior evidence separately links both appearance-contingent self-worth and self-objectification to these negative outcomes, no work has examined the mediating processes that may underlie this relationship. With the current project, we examined the relationship between appearance-contingent self-worth and self-objectification, and the degree to which this relationship is associated with higher appearance anxiety and lower overall self-esteem. We hypothesized that appearance-contingent self-worth would be positively associated with self-objectification; in turn, we expected self-objectification to be related to higher appearance anxiety, and ultimately, lower self-esteem. Across two studies, one cross-sectional (N=208) and one short-term longitudinal (N=191), we found compelling support for this hypothesis. These findings have practical and theoretical significance for both the self-objectification and contingent self-worth literatures. Published by Elsevier Ltd.

High-resolution CT of airway reactivity

International Nuclear Information System (INIS)

Herold, C.J.; Brown, R.H.; Hirshman, C.A.; Mitzner, W.; Zerhouni, E.A.

1990-01-01

Assessment of airway reactivity has generally been limited to experimental nonimaging models. This authors of this paper used high-resolution CT (HRCT) to evaluate airway reactivity and to calculate airway resistance (Raw) compared with lung resistance (RL). Ten anesthetized and ventilated dogs were investigated with HRCT (10 contiguous 2-mm sections through the lower lung lobes) during control state, following aerosol histamine challenge, and following posthistamine hyperinflation. The HRCT scans were digitized, and areas of 10 airways per dog (diameter, 1-10 mm) were measured with a computer edging process. Changes in airway area and Raw (calculated by 1/[area] 2 ) were measured. RL was assessed separately, following the same protocol. Data were analyzed by use of a paired t-test with significance at p < .05

Characteristic test of initial HTTR core

International Nuclear Information System (INIS)

Nojiri, Naoki; Shimakawa, Satoshi; Fujimoto, Nozomu; Goto, Minoru

2004-01-01

This paper describes the results of core physics test in start-up and power-up of the HTTR. The tests were conducted in order to ensure performance and safety of the high temperature gas cooled reactor, and was carried out to measure the critical approach, the excess reactivity, the shutdown margin, the control rod worth, the reactivity coefficient, the neutron flux distribution and the power distribution. The expected core performance and the required reactor safety characteristics were verified from the results of measurements and calculations

Glaciotectonic deformation and reinterpretation of the Worth Point stratigraphic sequence: Banks Island, NT, Canada

Science.gov (United States)

Vaughan, Jessica M.; England, John H.; Evans, David J. A.

2014-05-01

Hill-hole pairs, comprising an ice-pushed hill and associated source depression, cluster in a belt along the west coast of Banks Island, NT. Ongoing coastal erosion at Worth Point, southwest Banks Island, has exposed a section (6 km long and ˜30 m high) through an ice-pushed hill that was transported ˜ 2 km from a corresponding source depression to the southeast. The exposed stratigraphic sequence is polydeformed and comprises folded and faulted rafts of Early Cretaceous and Late Tertiary bedrock, a prominent organic raft, Quaternary glacial sediments, and buried glacial ice. Three distinct structural domains can be identified within the stratigraphic sequence that represent proximal to distal deformation in an ice-marginal setting. Complex thrust sequences, interfering fold-sets, brecciated bedrock and widespread shear structures superimposed on this ice-marginally deformed sequence record subsequent deformation in a subglacial shear zone. Analysis of cross-cutting relationships within the stratigraphic sequence combined with OSL dating indicate that the Worth Point hill-hole pair was deformed during two separate glaciotectonic events. Firstly, ice sheet advance constructed the hill-hole pair and glaciotectonized the strata ice-marginally, producing a proximal to distal deformation sequence. A glacioisostatically forced marine transgression resulted in extensive reworking of the strata and the deposition of a glaciomarine diamict. A readvance during this initial stage redeformed the strata in a subglacial shear zone, overprinting complex deformation structures and depositing a glaciotectonite ˜20 m thick. Outwash channels that incise the subglacially deformed strata record a deglacial marine regression, whereas aggradation of glaciofluvial sand and gravel infilling the channels record a subsequent marine transgression. Secondly, a later, largely non-erosive ice margin overrode Worth Point, deforming only the most surficial units in the section and depositing a

Assessment of PWR safety with regard to disturbances due to reactivity changes

International Nuclear Information System (INIS)

Pernica, R.

1980-01-01

The steady state method is briefly described for reactivity disturbances assessment using steady state calculations for two sets of reactivity coefficients and four values of the thermal conductivity of the gap. The variations were processed of the limit values of reactivity being applied with the thermal conductivity of the gap between the fuel and the can. All calculations were performed for a reactor with four core zones exposed to different radial thermal stresses with different fuel element proportional stresses. The results are shown in graphs. (J.B.)

Measurements of total OH reactivity during PROPHET-AMOS 2016

Science.gov (United States)

Rickly, P.; Sakowski, J.; Bottorff, B.; Lew, M.; Stevens, P. S.; Sklaveniti, S.; Locoge, N.; Dusanter, S.

2017-12-01

As one of the main oxidant in the atmosphere, the hydroxyl radical (OH) initiates the oxidation of volatile organic compounds that can lead to the formation of ozone and secondary organic aerosols. Understanding both the sources and sinks of OH is therefore important to address issues related to air quality and climate change. Measurements of total OH reactivity can provide an important test of our understanding of the OH radical budget. Recent measurements of total reactivity in many environments have been greater than calculated based on the measured concentration of VOCs, suggesting that important OH sinks in these environments are not well characterized. Measurements of total OH reactivity were performed in a forested environment during the PROPHET - AMOS field campaign (Program for Research on Oxidants: PHotochemisty, Emissions, and Transport - Atmospheric Measurements of Oxidants in Summer) using the Comparative Reactivity Method (CRM) and the Total OH Loss Rate Method (TOHLM). The site is characterized by large emissions of isoprene and monoterpenes and low anthropogenic influence. Measurements of total OH reactivity using these two techniques agree to within their respective uncertainties, giving confidence in the measured OH reactivity. In addition, measurements of trace gases (VOCs, NOx, O3) were used to perform a comprehensive apportionment of OH sinks. These measurements are used in a chemical model using the Master Chemical Mechanism to calculate the expected OH reactivity. The results will be compared to previous measurements of total OH reactivity at this site.

Using influence diagrams for data worth analysis

International Nuclear Information System (INIS)

Sharif Heger, A.; White, Janis E.

1997-01-01

Decision-making under uncertainty describes most environmental remediation and waste management problems. Inherent limitations in knowledge concerning contaminants, environmental fate and transport, remedies, and risks force decision-makers to select a course of action based on uncertain and incomplete information. Because uncertainties can be reduced by collecting additional data., uncertainty and sensitivity analysis techniques have received considerable attention. When costs associated with reducing uncertainty are considered in a decision problem, the objective changes; rather than determine what data to collect to reduce overall uncertainty, the goal is to determine what data to collect to best differentiate between possible courses of action or decision alternatives. Environmental restoration and waste management requires cost-effective methods for characterization and monitoring, and these methods must also satisfy regulatory requirements. Characterization and monitoring activities imply that, sooner or later, a decision must be made about collecting new field data. Limited fiscal resources for data collection should be committed only to those data that have the most impact on the decision at lowest possible cost. Applying influence diagrams in combination with data worth analysis produces a method which not only satisfies these requirements but also gives rise to an intuitive representation of complex structures not possible in the more traditional decision tree representation. This paper demonstrates the use of influence diagrams in data worth analysis by applying to a monitor-and-treat problem often encountered in environmental decision problems

Vaccination against meningitis B: is it worth it?

Directory of Open Access Journals (Sweden)

Peter English

2013-01-01

Full Text Available Summary: when a new vaccine is licensed having passed the tests for efficacy and safety, governments who have to pay for it to be used, as in the UK, will carry out a careful economic appraisal before making it generally available. In this challenging article, Dr Peter English discusses whether or not a new vaccine for meningitis that has just been licensed and that could save lives is actually worth it.

Influence of the temperature distribution on the reactivity of the reactor channel; Uticaj aksijalne raspodele temperature na reaktivnost kanala

Energy Technology Data Exchange (ETDEWEB)

Stojadinovic, A; Pop-Jordanov, J; Zivkovic, Z [The Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1964-07-01

For calculating the reactivity in the reactor channel, it was estimated that there is a linear increase of the neutron temperature along the channel. The channel is divide into 5 regions. Reactivity of the channel was calculated by using the reactivity curves for each region. it has been compared to the reactivity values obtained for different mean temperature values. The calculations were done on the digital computer Zuse-Z-23.

Men seek social standing, women seek companionship: sex differences in deriving self-worth from relationships.

Science.gov (United States)

Kwang, Tracy; Crockett, Erin E; Sanchez, Diana T; Swann, William B

2013-07-01

Do men base their self-worth on relationships less than do women? In an assessment of lay beliefs, men and women alike indicated that men are less reliant on relationships as a source of self-worth than are women (Study 1). Yet relationships may make a different important contribution to the self-esteem of men. Men reported basing their self-esteem on their own relationship status (whether or not they were in a relationship) more than did women, and this link was statistically mediated by the perceived importance of relationships as a source of social standing (Studies 1 and 2). Finally, when relationship status was threatened, men displayed increased social-standing concerns, whereas women displayed increased interdependence concerns (Study 3). Together, these findings demonstrate that both men and women rely on relationships for self-worth, but that they derive self-esteem from relationships in different ways.

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.

AIREKMOD-RR, Reactivity Transients in Nuclear Research Reactors

International Nuclear Information System (INIS)

Baggoura, B.; Mazrou, H.

2001-01-01

1 - Description of program or function: AIREMOD-RR is a point kinetics code which can simulate fast transients in nuclear research reactor cores. It can also be used for theoretical reactor dynamics studies. It is used for research reactor kinetic analysis and provides a point neutron kinetic capability. The thermal hydraulic behavior is governed by a one-dimensional heat balance equation. The calculations are restricted to a single equivalent unit cell which consists of fuel, clad and coolant. 2 - Method of solution: For transient reactor kinetic calculations a modified Runge Kutta numerical method is used. The external reactivity insertion, specified as a function of time, is converted in dollar ($) unit. The neutron density, energy release and feedback variables are given at each time step. The two types of reactivity feedback considered are: Doppler effect and moderator effect. A new expression for the reactivity dependence on the feedback variables has been introduced in the present version of the code. The feedback reactivities are fitted in power series expression. 3 - Restrictions on the complexity of the problem: The number of delayed neutron groups and the total number of equations are limited only by computer storage capabilities. - Coolant is always in liquid phase. - Void reactivity feedback is not considered

Cold and ultracold dynamics of the barrierless D{sup +} + H{sub 2} reaction: Quantum reactive calculations for ∼R{sup −4} long range interaction potentials

Energy Technology Data Exchange (ETDEWEB)

Lara, Manuel, E-mail: manuel.lara@uam.es [Departamento de Química Física Aplicada, Facultad de Ciencias, Universidad Autónoma de Madrid, 28049 Madrid (Spain); Jambrina, P. G.; Aoiz, F. J. [Departamento de Química Física, Facultad de Química, Universidad Complutense, 28040 Madrid (Spain); Launay, J.-M. [Institut de Physique de Rennes, UMR CNRS 6251, Université de Rennes I, F-35042 Rennes (France)

2015-11-28

Quantum reactive and elastic cross sections and rate coefficients have been calculated for D{sup +} + H{sub 2} (v = 0, j = 0) collisions in the energy range from 10{sup −8} K (deep ultracold regime), where only one partial wave is open, to 150 K (Langevin regime) where many of them contribute. In systems involving ions, the ∼R{sup −4} behavior extends the interaction up to extremely long distances, requiring a special treatment. To this purpose, we have used a modified version of the hyperspherical quantum reactive scattering method, which allows the propagations up to distances of 10{sup 5} a{sub 0} needed to converge the elastic cross sections. Interpolation procedures are also proposed which may reduce the cost of exact dynamical calculations at such low energies. Calculations have been carried out on the PES by Velilla et al. [J. Chem. Phys. 129, 084307 (2008)] which accurately reproduces the long range interactions. Results on its prequel, the PES by Aguado et al. [J. Chem. Phys. 112, 1240 (2000)], are also shown in order to emphasize the significance of the inclusion of the long range interactions. The calculated reaction rate coefficient changes less than one order of magnitude in a collision energy range of ten orders of magnitude, and it is found in very good agreement with the available experimental data in the region where they exist (10-100 K). State-to-state reaction probabilities are also provided which show that for each partial wave, the distribution of HD final states remains essentially constant below 1 K.

Neutron flux calculations for the Rossendorf research reactor in (hex)- and (hex,z)-geometry using SNAP-3D

International Nuclear Information System (INIS)

Koch, R.; Findeisen, A.

1986-04-01

The multigroup neutron diffusion theory code SNAP-3D has been used to perform time independent neutron flux and power calculations of the 10 MW Rossendorf research reactor of the type WWR-SM. The report describes these calculations, as well as the actual reactor configuration, some details of the code SNAP-3D, and two- and three-dimensional reactor models. For evaluating the calculations some flux values and control rod worths have been compared with those of measurements. (author)

Reactivity feedback coefficients of a low enriched uranium fuelled material test research reactor at end-of-life

International Nuclear Information System (INIS)

Muhammad, Farhan

2011-01-01

Highlights: → The isotopic concentration in the fuel changes as soon as it starts its operation. → The neutronic properties of a reactor also change with fuel burnup. → The reactivity feedbacks at end-of-life of a material test reactor fuelled with low enriched uranium fuel are calculated. → Codes used include WIMS-D4 and CITATION. - Abstract: The reactivity feedback coefficients at end-of-life of a material test reactor fuelled with low enriched uranium fuel were calculated. The reactor used for the study was the IAEA's 10 MW benchmark reactor. Simulations were carried out to calculate the different reactivity feedback coefficients including Doppler feedback coefficient, reactivity coefficient for change of water temperature and reactivity coefficient for change of water density. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It was observed that the magnitude of all the reactivity feedback coefficients increased at end of life of the reactor by almost 2-5%.

Reactivity feedback coefficients Pakistan research reactor-1 using PRIDE code

Energy Technology Data Exchange (ETDEWEB)

Mansoor, Ali; Ahmed, Siraj-ul-Islam; Khan, Rustam [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Inam-ul-Haq [Comsats Institute of Information Technology, Islamabad (Pakistan). Dept. of Physics

2017-05-15

Results of the analyses performed for fuel, moderator and void's temperature feedback reactivity coefficients for the first high power core configuration of Pakistan Research Reactor - 1 (PARR-1) are summarized. For this purpose, a validated three dimensional model of PARR-1 core was developed and confirmed against the reference results for reactivity calculations. The ''Program for Reactor In-Core Analysis using Diffusion Equation'' (PRIDE) code was used for development of global (3-dimensional) model in conjunction with WIMSD4 for lattice cell modeling. Values for isothermal fuel, moderator and void's temperature feedback reactivity coefficients have been calculated. Additionally, flux profiles for the five energy groups were also generated.

For What It's Worth: Historical Financial Bubbles and the Boundaries of Economic Rationality.

Science.gov (United States)

Deringer, William

2015-09-01

This essay is a historical and epistemological exploration of a traditionally crazy economic event: the financial bubble. Venturing into two different moments in the history of economic thinking, it investigates financial bubbles as epistemic frontiers, where rationality has reached its limits. The first half forays into late twentieth-century economics. Since 1980, an interpretive battle over the ir/rationality of bubbles has made those peculiar events, long beyond the pale of the rational, contested terrain on which the limits of rationality have been fought out. The essay's second half turns to one historical crisis, the South Sea Bubble. For contemporaries in 1720, the bubble was a different kind of epistemic frontier. As they tried to reckon what South Sea Company stock was worth, investors were confronted not with clearly rational or irrational choices but with a decidedly unruly collection of similarly plausible calculations. The story of 1720 suggests that studying historical confusion might be a profitable enterprise for scholars of the economic and epistemological past.

Synthesis and DFT calculations of some 2-aminothiazoles

Science.gov (United States)

Rezania, Jafar; Behzadi, Hadi; Shockravi, Abbas; Ehsani, Morteza; Akbarzadeh, Elahe

2018-04-01

A series of 2-aminothiazole derivatives have been synthesized by the reaction of acetyl compounds with thiourea and iodine as catalyst under solvent-free condition, a green chemistry method. The quantum chemical calculations at the DFT/B3LYP level of theory in gas phase were carried out for starting acetyl derivatives. The highest occupied molecular orbital (HOMO) and lowest unoccupied molecular orbital (LUMO) and related reactivity descriptor of acetyl derivatives, as well as, enthalpy of reactions are calculated in order to investigate the reaction properties of acetyl compounds and yields of the reactions. The calculated reactivity descriptors are well correlated to activity of different acetyl derivatives.

How well can global chemistry models calculate the reactivity of short-lived greenhouse gases in the remote troposphere, knowing the chemical composition

Science.gov (United States)

Prather, Michael J.; Flynn, Clare M.; Zhu, Xin; Steenrod, Stephen D.; Strode, Sarah A.; Fiore, Arlene M.; Correa, Gustavo; Murray, Lee T.; Lamarque, Jean-Francois

2018-05-01

We develop a new protocol for merging in situ measurements with 3-D model simulations of atmospheric chemistry with the goal of integrating these data to identify the most reactive air parcels in terms of tropospheric production and loss of the greenhouse gases ozone and methane. Presupposing that we can accurately measure atmospheric composition, we examine whether models constrained by such measurements agree on the chemical budgets for ozone and methane. In applying our technique to a synthetic data stream of 14 880 parcels along 180° W, we are able to isolate the performance of the photochemical modules operating within their global chemistry-climate and chemistry-transport models, removing the effects of modules controlling tracer transport, emissions, and scavenging. Differences in reactivity across models are driven only by the chemical mechanism and the diurnal cycle of photolysis rates, which are driven in turn by temperature, water vapor, solar zenith angle, clouds, and possibly aerosols and overhead ozone, which are calculated in each model. We evaluate six global models and identify their differences and similarities in simulating the chemistry through a range of innovative diagnostics. All models agree that the more highly reactive parcels dominate the chemistry (e.g., the hottest 10 % of parcels control 25-30 % of the total reactivities), but do not fully agree on which parcels comprise the top 10 %. Distinct differences in specific features occur, including the spatial regions of maximum ozone production and methane loss, as well as in the relationship between photolysis and these reactivities. Unique, possibly aberrant, features are identified for each model, providing a benchmark for photochemical module development. Among the six models tested here, three are almost indistinguishable based on the inherent variability caused by clouds, and thus we identify four, effectively distinct, chemical models. Based on this work, we suggest that water vapor

Estimating NIRR-1 burn-up and core life time expectancy using the codes WIMS and CITATION

Science.gov (United States)

Yahaya, B.; Ahmed, Y. A.; Balogun, G. I.; Agbo, S. A.

The Nigeria Research Reactor-1 (NIRR-1) is a low power miniature neutron source reactor (MNSR) located at the Centre for Energy Research and Training, Ahmadu Bello University, Zaria Nigeria. The reactor went critical with initial core excess reactivity of 3.77 mk. The NIRR-1 cold excess reactivity measured at the time of commissioning was determined to be 4.97 mk, which is more than the licensed range of 3.5-4 mk. Hence some cadmium poison worth -1.2 mk was inserted into one of the inner irradiation sites which act as reactivity regulating device in order to reduce the core excess reactivity to 3.77 mk, which is within recommended licensed range of 3.5 mk and 4.0 mk. In this present study, the burn-up calculations of the NIRR-1 fuel and the estimation of the core life time expectancy after 10 years (the reactor core expected cycle) have been conducted using the codes WIMS and CITATION. The burn-up analyses carried out indicated that the excess reactivity of NIRR-1 follows a linear decreasing trend having 216 Effective Full Power Days (EFPD) operations. The reactivity worth of top beryllium shim data plates was calculated to be 19.072 mk. The result of depletion analysis for NIRR-1 core shows that (7.9947 ± 0.0008) g of U-235 was consumed for the period of 12 years of operating time. The production of the build-up of Pu-239 was found to be (0.0347 ± 0.0043) g. The core life time estimated in this research was found to be 30.33 years. This is in good agreement with the literature

Method for determining detailed rod worth profiles at low power in the fast test reactor

International Nuclear Information System (INIS)

Sevenich, R.A.

1975-08-01

A method for obtaining a detailed rod worth profile at low power for a slow control rod insertion is presented. The accuracy of the method depends on a preparatory experiment in which the test rod is dropped quickly to yield, upon analysis, the magnitude of the rod worth and an effective source value. These numbers are employed to initialize the inverse kinetics analysis for the slow insertion. Corrections for changes in detection efficiency are not included for the simulated experiments. (U.S.)

Actinides integral measurements on FCA assemblies

International Nuclear Information System (INIS)

Mukaiyama, Takehiko; Okajima, Shigeaki

1984-01-01

Actinide integral measurements were performed on eight assemblies of FCA where neutron energy spectra were shifted systematically from soft to hard in order to evaluate and modify the nuclear cross section data of major actinides. Experimental values on actinide fission rates and sample reactivity worths are compared with the calculated values using JENDL-2 and ENDF/B-V (or IV) data sets. (author)

Is Self-Worth Protection Best Regarded as Intentional Self-Handicapping Behaviour or an Outcome of Choking under Pressure?

Science.gov (United States)

Thompson, Ted; Dinnel, Dale L.

2007-01-01

Self-worth protective students characteristically perform poorly when they anticipate that poor performance is likely to reveal low ability, yet perform well in situations that involve little threat to self-worth. The present study sought a further understanding of this variable pattern of achievement, assessing two possibilities: (1) that the…

The performance of ENDF/B-V.2 nuclear data for fast reactor calculations

International Nuclear Information System (INIS)

Atkinson, C.A.; Collins, P.J.

1987-01-01

Calculations with ENDF/B-V.2 data have been made for twenty-five fast-spectrum integral assemblies covering a wide range of sizes and compositions. Analysis was done by transport codes with refined cross section processing methods and detailed reactor modelling. The predictions of fission rate distributions and control rod worths were emphasized for the more prototypic benchmark cores. The results show considerable improvements in agreement with experiment compared with analysis using ENDF/B-IV data, but it is apparent that significant errors remain for fast reactor design calculations

Gender Differences in Comparisons and Entitlement: Implications for Comparable Worth.

Science.gov (United States)

Major, Brenda

1989-01-01

Addresses the role of comparison processes in the persistence of the gender wage gap, its toleration by those disadvantaged by it, and resistance to comparable worth as a corrective strategy. Argues that gender segregation and undercompensation for women's jobs leads women to use different comparison standards when evaluating what they deserve.…

An analysis of reactivity prediction during the reactor start-up process

International Nuclear Information System (INIS)

Bajgl, Josef; Krysl, Vaclav; Svarny, Jiri

2015-01-01

The different VVER-440 core fuel loadings subcriticality evaluations are performed during the start-up process by boron dilution or control assembly withdrawn by macrocode MOBY-DICK calculations. The dynamic reactivity and quasicritical reactivity are compared and sensitivity of reactivity prediction at the low boundary of start-up interval (ρ = -0,01) has been provided on the basis of different modelling of ionization chamber (IC) response calculation. Special attention is paid to the impact of power distribution and spontaneous fission distribution form factor on IC response correction during control assembly movement. Precision and robustness of different corrections of IC signal processing in real core start-up processed IC signals was evaluated.

Benchmark tests of JENDL-3.2 for thermal and fast reactors

International Nuclear Information System (INIS)

Takano, Hideki

1995-01-01

Benchmark calculations for a variety of thermal and fast reactors have been performed by using the newly evaluated JENDL-3 Version-2 (JENDL-3.2) file. In the thermal reactor calculations for the uranium and plutonium fueled cores of TRX and TCA, the k eff and lattice parameters were well predicted. The fast reactor calculations for ZPPR-9 and FCA assemblies showed that the k eff , reactivity worth of Doppler, sodium void and control rod, and reaction rate distribution were in a very good agreement with the experiments. (author)

Analysis Of Control Rod Ejection Of APR1400 By RELAP5

International Nuclear Information System (INIS)

Le Thi Thu; Hoang Minh Giang; Vo Thi Huong; Le Dai Dien

2011-01-01

This paper presents the analysis of Reactivity Induced Accident caused by ejection of a Control Element Assembly (CEA) from APR 1400 reactor vessel within 0.05 second. The initial condition were assumed as following: power level at 102%, delayed neutron fraction β = 412 pcm and CEA worth = 110 pcm. The analysis was simulated by RELAP5 code through two step: calculation of steady state and calculation of transient with initial condition mentioned as above. Some output results were presented with explanation: sequence of events corresponding to the time of the accident, the system behavior as power, reactivity feedback from fuel temperature changes (Doppler) as well as temperature, pressure, DNBR within 6 second of the accident. (author)

The tension between self governance and absolute inner worth in Kant's moral philosophy.

Science.gov (United States)

Häyry, M

2005-11-01

The concepts of autonomy as the self governance of individuals and dignity as the inner worth of human beings play an important role in contemporary bioethics. Since both notions are crucial to Immanuel Kant's moral theory, it would be tempting to think that Kantian ethics could ease the friction between the two concepts. It is argued in this paper, however, that this line of thought cannot be supported by Kant's original ideas. While he did make a conscious effort to bring autonomy and dignity together, his emphasis on the absolute inner worth of our collective humanity made it impossible for him to embrace fully the personal self determination of individuals, as it is usually understood in today's liberal thinking.

The reactivation time in the treatment of AMD: a forgotten key parameter?

Science.gov (United States)

Real, J P; Luna, J D; Palma, S D

2018-06-01

Summarize and compare the available evidence on the reactivation times in patients with age-related macular degeneration treated with Ranibizumab (RNB). Systematic review of studies that reported the reactivation time of patients (direct method) or the number of injections received in a certain period of follow-up (indirect method). Only 18 of 89 selected studies reported the average reactivation time of patients in a manifest form, without the need of any calculation. The average calculated, weighted reactivation time was 101.8 days with the direct method and 99.8 days in the indirect method (84 studies included). With both methods, it was found that the average reactivation time of the RCTs was between 2 and 3 weeks less than the average time identified in the observational studies. These differences are also reflected in the clinical results, there being a correlation between the number of doses received and the change in BCVA. The analysis of 11 comparative studies showed a difference in reactivation times between patients treated with RNB or Bevacizumab (BVZ). There are few direct studies of reactivation time, but calculation from the PRN dose number turns out to be a good approximation for retrospective study of the variable. The use of the PRN, with criteria not based on optical coherence tomography scans, delays the application of doses between 2 or 3 weeks, and patients suffer loss of clinical benefits. RNB enables patients to receive less injections than BVZ throughout treatment.

38 CFR 3.274 - Relationship of net worth to pension entitlement.

Science.gov (United States)

2010-07-01

... 38 Pensions, Bonuses, and Veterans' Relief 1 2010-07-01 2010-07-01 false Relationship of net worth to pension entitlement. 3.274 Section 3.274 Pensions, Bonuses, and Veterans' Relief DEPARTMENT OF VETERANS AFFAIRS ADJUDICATION Pension, Compensation, and Dependency and Indemnity Compensation Regulations...

Introducing a new bond reactivity index: Philicities for natural bond orbitals.

Science.gov (United States)

Sánchez-Márquez, Jesús; Zorrilla, David; García, Víctor; Fernández, Manuel

2017-12-22

In the present work, a new methodology defined for obtaining reactivity indices (philicities) is proposed. This is based on reactivity functions such as the Fukui function or the dual descriptor, and makes it possible to project the information from reactivity functions onto molecular orbitals, instead of onto the atoms of the molecule (atomic reactivity indices). The methodology focuses on the molecules' natural bond orbitals (bond reactivity indices) because these orbitals have the advantage of being localized, allowing the reaction site of an electrophile or nucleophile to be determined within a very precise molecular region. This methodology provides a "philicity" index for every NBO, and a representative set of molecules has been used to test the new definition. A new methodology has also been developed to compare the "finite difference" and the "frontier molecular orbital" approximations. To facilitate their use, the proposed methodology as well as the possibility of calculating the new indices have been implemented in a new version of UCA-FUKUI software. In addition, condensation schemes based on atomic populations of the "atoms in molecules" theory, the Hirshfeld population analysis, the approximation of Mulliken (with a minimal basis set) and electrostatic potential-derived charges have also been implemented, including the calculation of "bond reactivity indices" defined in previous studies. Graphical abstract A new methodology defined for obtaining bond reactivity indices (philicities) is proposed and makes it possible to project the information from reactivity functions onto molecular orbitals. The proposed methodology as well as the possibility of calculating the new indices have been implemented in a new version of UCA-FUKUI software. In addition, this version can use new atomic condensation schemes and new "utilities" have also been included in this second version.

Measurement of reactivity coefficients for code validation

International Nuclear Information System (INIS)

Nuding, Matthias; Loetsch, Thomas

2005-01-01

In the year 2003 measurements in the cold reactor state have been performed at the NPP KKI 2 in order to validate the codes that are used for reactor core calculations and especially for the proof of the shutdown margin that is produced by calculations only. For full power states code verification is quite easy because the calculations can be compared with different measured values, e.g. with the activation values determined by the aeroball system. For cold reactor states, however the data base is smaller, especially for reactor cores that are quite 'inhomogeneous' and have rather high Pu-fiss-and 235 U-contents. At the same time the cold reactor state is important regarding the shutdown margin. For these reasons the measurements mentioned above have been performed in order to check the accuracy of the codes that are used by the operator and by our organization for many years. Basically, boron concentrations and control rod worths for different configurations have been measured. The results of the calculation show a very good agreement with the measured values. Therefore, it can be stated that the operator's as well as our code system is suitable for routine use, e.g. during licensing procedures (Authors)

Supercomputer algorithms for reactivity, dynamics and kinetics of small molecules

International Nuclear Information System (INIS)

Lagana, A.

1989-01-01

Even for small systems, the accurate characterization of reactive processes is so demanding of computer resources as to suggest the use of supercomputers having vector and parallel facilities. The full advantages of vector and parallel architectures can sometimes be obtained by simply modifying existing programs, vectorizing the manipulation of vectors and matrices, and requiring the parallel execution of independent tasks. More often, however, a significant time saving can be obtained only when the computer code undergoes a deeper restructuring, requiring a change in the computational strategy or, more radically, the adoption of a different theoretical treatment. This book discusses supercomputer strategies based upon act and approximate methods aimed at calculating the electronic structure and the reactive properties of small systems. The book shows how, in recent years, intense design activity has led to the ability to calculate accurate electronic structures for reactive systems, exact and high-level approximations to three-dimensional reactive dynamics, and to efficient directive and declaratory software for the modelling of complex systems

Monte Carlo analysis of Musashi TRIGA mark II reactor core

International Nuclear Information System (INIS)

Matsumoto, Tetsuo

1999-01-01

The analysis of the TRIGA-II core at the Musashi Institute of Technology Research Reactor (Musashi reactor, 100 kW) was performed by the three-dimensional continuous-energy Monte Carlo code (MCNP4A). Effective multiplication factors (k eff ) for the several fuel-loading patterns including the initial core criticality experiment, the fuel element and control rod reactivity worth as well as the neutron flux measurements were used in the validation process of the physical model and neutron cross section data from the ENDF/B-V evaluation. The calculated k eff overestimated the experimental data by about 1.0%Δk/k for both the initial core and the several fuel-loading arrangements. The calculated reactivity worths of control rod and fuel element agree well the measured ones within the uncertainties. The comparison of neutron flux distribution was consistent with the experimental ones which were measured by activation methods at the sample irradiation tubes. All in all, the agreement between the MCNP predictions and the experimentally determined values is good, which indicated that the Monte Carlo model is enough to simulate the Musashi TRIGA-II reactor core. (author)