Systematic of delayed neutron parameters
International Nuclear Information System (INIS)
Isaev, S.G.; Piksaikin, V.M.
2000-01-01
The experimental studies of the energy dependence of the delayed neutron (DN) parameters for various fission systems has shown that the behaviour of a some combination of delayed neutron parameters has a similar features. On the basis of this findings the systematics of delayed neutron experimental data for thorium, uranium, plutonium and americium isotopes have been investigated with the purpose to find a correlation of DN parameters with characteristics of fissioning system as well as a correlation between the delayed neutron parameters themselves. It was presented the preliminary results which were obtained during study the physics interpretation of the results [ru
Neutron delayed choice experiments
International Nuclear Information System (INIS)
Bernstein, H.J.
1986-01-01
Delayed choice experiments for neutrons can help extend the interpretation of quantum mechanical phenomena. They may also rule out alternative explanations which static interference experiments allow. A simple example of a feasible neutron test is presented and discussed. (orig.)
International Nuclear Information System (INIS)
Wall, T.
1988-01-01
Delayed neutron analysis carried out at the Australian Nuclear Scientific and Technology Organization facilities, provides a fast, high sensitivity, low cost, reliable method, particularly suitable for large batches of samples, and for non destructive analysis of a range of materials. While its main use has been in uranium exploration, other applications include archeological investigations, agriculture, oceanography and biology
Correlation properties of delayed neutrons from fast neutron induced fission
International Nuclear Information System (INIS)
Piksaikin, V.M.; Isaev, S.G.
1998-01-01
The experimental studies of the energy dependence of the delayed neutron parameters for various fissioning systems has shown that the behavior of a some combination of delayed neutron parameters (group relative abundances a i and half lives T i ) has a similar features. On the basis of this findings the systematics of delayed neutron experimental data for thorium, uranium, plutonium and americium isotopes have been investigated with the purpose to find a correlation of DN parameters with characteristics of fissioning system as well as a correlation between the delayed neutron parameters themselves. Below we will present the preliminary results which were obtained during this study omitting the physics interpretation of the results. (author)
Neutron stochastic transport theory with delayed neutrons
International Nuclear Information System (INIS)
Munoz-Cobo, J.L.; Verdu, G.
1987-01-01
From the stochastic transport theory with delayed neutrons, the Boltzmann transport equation with delayed neutrons for the average flux emerges in a natural way without recourse to any approximation. From this theory a general expression is obtained for the Feynman Y-function when delayed neutrons are included. The single mode approximation for the particular case of a subcritical assembly is developed, and it is shown that Y-function reduces to the familiar expression quoted in many books, when delayed neutrons are not considered, and spatial and source effects are not included. (author)
Analytical applications for delayed neutrons
International Nuclear Information System (INIS)
Eccleston, G.W.
1983-01-01
Analytical formulations that describe the time dependence of neutron populations in nuclear materials contain delayed-neutron dependent terms. These terms are important because the delayed neutrons, even though their yields in fission are small, permit control of the fission chain reaction process. Analytical applications that use delayed neutrons range from simple problems that can be solved with the point reactor kinetics equations to complex problems that can only be solved with large codes that couple fluid calculations with the neutron dynamics. Reactor safety codes, such as SIMMER, model transients of the entire reactor core using coupled space-time neutronics and comprehensive thermal-fluid dynamics. Nondestructive delayed-neutron assay instruments are designed and modeled using a three-dimensional continuous-energy Monte Carlo code. Calculations on high-burnup spent fuels and other materials that contain a mix of uranium and plutonium isotopes require accurate and complete information on the delayed-neutron periods, yields, and energy spectra. A continuing need exists for delayed-neutron parameters for all the fissioning isotopes
Systematics in delayed neutron yields
Energy Technology Data Exchange (ETDEWEB)
Ohsawa, Takaaki [Kinki Univ., Higashi-Osaka, Osaka (Japan). Atomic Energy Research Inst.
1998-03-01
An attempt was made to reproduce the systematic trend observed in the delayed neutron yields for actinides on the basis of the five-Gaussian representation of the fission yield together with available data sets for delayed neutron emission probability. It was found that systematic decrease in DNY for heavier actinides is mainly due to decrease of fission yields of precursors in the lighter side of the light fragment region. (author)
Delayed neutron yield from fast neutron induced fission of 238U
International Nuclear Information System (INIS)
Piksaikin, V.M.; Kazakov, L.E.; Isaev, S.G.; Roshchenko, V.A.; Goverdovski, A.A.; Tertytchnyi, R.G.
2002-01-01
The measurements of the total delayed neutron yield from fast neutron induced fission of 238 U were made. The experimental method based on the periodic irradiation of the fissionable sample by neutrons from a suitable nuclear reaction had been employed. The preliminary results on the energy dependence of the total delayed neutron yield from fission of 238 U are obtained. According to the comparison of experimental data with our prediction based on correlation properties of delayed neutron characteristics, it is concluded that the value of the total delayed neutron yield near the threshold of (n,f) reaction is not a constant. (author)
International Nuclear Information System (INIS)
Gross, K.C.
1980-07-01
The application of the computer code GIRAFFE (General Isotope Release Analysis For Failed Elements) written in FORTRAN IV is described. GIRAFFE was designed to provide parameter estimates of the nonlinear discrete-measurement models that govern the transport and decay of delayed-neutron precursors in a liquid-metal fast breeder reactor (LMFBR). The code has been organized into a set of small, relatively independent and well-defined modules to facilitate modification and maintenance. The program logic, the numerical techniques, and the methods of solution used by the code are presented, and the functions of the MAIN program and of each subroutine are discussed
Study of calculated and measured time dependent delayed neutron yields
International Nuclear Information System (INIS)
Waldo, R.W.
1980-05-01
Time-dependent delayed neutron emission is of interest in reactor design, reactor dynamics, and nuclear physics studies. The delayed neutrons from neutron-induced fission of 232 U, 237 Np, 238 Pu, 241 Am, /sup 242m/Am, 245 Cm, and 249 Cf were studied for the first time. The delayed neutron emission from 232 Th, 233 U, 235 U, 238 U, 239 Pu, 241 Pu, and 242 Pu were measured as well. The data were used to develop an empirical expression for the total delayed neutron yield. The expression gives accurate results for a large variety of nuclides from 232 Th to 252 Cf. The data measuring the decay of delayed neutrons with time were used to derive another empirical expression predicting the delayed neutron emission with time. It was found that nuclides with similar mass-to-charge ratios have similar decay patterns. Thus the relative decay pattern of one nuclide can be established by any measured nuclide with a similar mass-to-charge ratio. A simple fission product yield model was developed and applied to delayed neutron precursors. It accurately predicts observed yield and decay characteristics. In conclusion, it is possible to not only estimate the total delayed neutron yield for a given nuclide but the time-dependent nature of the delayed neutrons as well. Reactors utilizing recycled fuel or burning actinides are likely to have inventories of fissioning nuclides that have not been studied until now. The delayed neutrons from these nuclides can now be incorporated so that their influence on the stability and control of reactors can be delineated. 8 figures, 39 tables
Delayed neutrons in liquid metal spallation targets
International Nuclear Information System (INIS)
Ridikas, D.; Bokov, P.; David, J.C.; Dore, D.; Giacri, M.L.; Van Lauwe, A.; Plukiene, R.; Plukis, A.; Ignatiev, S.; Pankratov, D.
2003-01-01
The next generation spallation neutron sources, neutrino factories or RIB production facilities currently being designed and constructed around the world will increase the average proton beam power on target by a few orders of magnitude. Increased proton beam power results in target thermal hydraulic issues leading to new target designs, very often based on flowing liquid metal targets such as Hg, Pb, Pb-Bi. Radioactive nuclides produced in liquid metal targets are transported into hot cells, past electronics, into pumps with radiation sensitive components, etc. Besides the considerable amount of photon activity in the irradiated liquid metal, a significant amount of the delayed neutron precursor activity can be accumulated in the target fluid. The transit time from the front of a liquid metal target into areas, where delayed neutrons may be important, can be as short as a few seconds, well within one half-life of many delayed neutron precursors. Therefore, it is necessary to evaluate the total neutron flux (including delayed neutrons) as a function of time and determine if delayed neutrons contribute significantly to the dose rate. In this study the multi-particle transport code MCNPX combined with the material evolution program CINDER'90 will be used to evaluate the delayed neutron flux and spectra. The following scientific issues will be addressed in this paper: - Modeling of a typical geometry of the liquid metal spallation target; - Predictions of the prompt neutron fluxes, fission fragment and spallation product distributions; - Comparison of the above parameters with existing experimental data; - Time-dependent calculations of delayed neutron precursors; - Neutron flux estimates due to the prompt and delayed neutron emission; - Proposal of an experimental program to measure delayed neutron spectra from high energy spallation-fission reactions. The results of this study should be directly applicable in the design study of the European MegaPie (1 MW
Radiochemical Means of Investigating Delayed Neutron Precursors
International Nuclear Information System (INIS)
Marmol, P. del
1968-01-01
Fast radiochemical methods used now for the determination of delayed neutron precursors are classified and reviewed: precipitations, solvent extractions, range experiments, milking, gas sweeping, isotopic and ion exchange, hot atom reactions and diffusion loss. Advantages and limitations of irradiation systems with respect to fast separations are discussed: external beams which allow faster separations only have low neutron fluxes, internal beams which are mostly fit for gaseous reactions; and rabbits for solution irradiations. Future prospects of radiochemical procedures are presented; among these, studies should be mostly oriented towards gaseous reactions which offer possibilities of isolating very short-lived delayed neutron precursors. Chemical procedures for delayed neutron precursor detection are compared with mass spectrometric and isotope separator techniques; it is concluded that the methods are complementary. (author)
Radiochemical Means of Investigating Delayed Neutron Precursors
International Nuclear Information System (INIS)
Marmol, P. del
1968-01-01
Fast radiochemical methods used now for the determination of delayed neutron precursors are classified and reviewed: precipitations, solvent extractions, range experiments, milking, gas sweeping, isotopic and ion exchange, hot-atom reactions and diffusion loss. Advantages and limitations of irradiation systems with respect to fast separations are discussed: external beams which allow faster separations only have low neutron fluxes, internal beams which are mostly fit for gaseous reactions; and rabbits for solution irradiations. Future prospects of radiochemical procedures are presented; among these, studies should be mostly oriented towards gaseous reactions which offer possibilities of isolating very short-lived delayed neutron precursors. Chemical procedures for delayed neutron precursor detection are compared with mass spectrometric and isotope-separator techniques; it is concluded that the methods are complementary. (author)
The delayed neutron method of uranium analysis
International Nuclear Information System (INIS)
Wall, T.
1989-01-01
The technique of delayed neutron analysis (DNA) is discussed. The DNA rig installed on the MOATA reactor, the assay standards and the types of samples which have been assayed are described. Of the total sample throughput of about 55,000 units since the uranium analysis service began, some 78% has been concerned with analysis of uranium ore samples derived from mining and exploration. Delayed neutron analysis provides a high sensitivity, low cost uranium analysis method for both uranium exploration and other applications. It is particularly suitable for analysis of large batch samples and for non-destructive analysis over a wide range of matrices. 8 refs., 4 figs., 3 tabs
The effective delayed neutron fraction for bare-metal criticals
International Nuclear Information System (INIS)
Pearlstein, S.
1999-01-01
Given sufficient material, a large number of actinides could be used to form bare-metal criticals. The effective delayed neutron fraction for a bare critical comprised of a fissile material is comparable with the absolute delayed neutron fraction. The effective delayed neutron fraction for a bare critical composed of a fissionable material is reduced by factors of 2 to 10 when compared with the absolute delayed neutron fraction. When the effective delayed neutron fraction is small, the difference between delayed and prompt criticality is small, and extreme caution must be used in critical assemblies of these materials. This study uses an approximate but realistic model to survey the actinide region to compare effective delayed neutron fractions with absolute delayed neutron fractions
Improved Delayed-Neutron Spectroscopy Using Trapped Ions
Energy Technology Data Exchange (ETDEWEB)
Norman, Eric
2018-04-24
The neutrons emitted following the decay of fission fragments (known as delayed neutrons because they are emitted after fission on a timescale of the -decay half-lives) play a crucial role in reactor performance and control. Reviews of delayed-neutron properties highlight the need for high-quality data for a wide variety of delayed-neutron emitters to better understand the timedependence and energy spectrum of the neutrons as these properties are essential for a detailed understanding of reactor kinetics needed for reactor safety and to understand the behavior of these reactors under various accident and component-failure scenarios. For fast breeder reactors, criticality calculations require accurate delayed-neutron energy spectra and approximations that are acceptable for light-water reactors such as assuming the delayed-neutron and fission-neutron energy spectra are identical are not acceptable and improved -delayed neutron data is needed for safety and accident analyses for these reactors. With improved nuclear data, the delayedneutrons flux and energy spectrum could be calculated from the contributions from individual isotopes and therefore could be accurately modeled for any fuel-cycle concept, actinide mix, or irradiation history. High-quality -delayed neutron measurements are also critical to constrain modern nuclear-structure calculations and empirical models that predict the decay properties for nuclei for which no data exists and improve the accuracy and flexibility of the existing empirical descriptions of delayed neutrons from fission such as the six-group representation
Proceedings of the specialists' meeting on delayed neutron nuclear data
International Nuclear Information System (INIS)
Katakura, Jun-ichi
1999-07-01
This report is the Proceedings of the Specialists' Meeting on Delayed Neutron Nuclear Data. The meeting was held on January 28-29, 1999, at the Tokai Research Establishment of Japan Atomic Energy Research Institute with the participation of thirty specialists, who are evaluators, theorist, experimentalists. Although the fraction of the delayed neutron is no more than 1% in the total neutrons emitted in the fission process, it plays an important roll in the control of fission reactor. In the meeting, the following topics were reported: the present status of delayed neutron data in the major evaluated data libraries, measurements of effective delayed neutron fraction using FCA (Fast Critical Assembly) and TCA (Tank-type Critical Assembly) and their analyses, sensitivity analysis for fast reactor, measurements of delayed neutron emission from actinides and so on. As another topics, delayed neutron in transmutation system and fission yield data were also presented. Free discussion was held on the future activity of delayed neutron data evaluation. The discussion was helpful for the future activity of the delayed neutron working group of JNDC aiming to the evaluation of delayed neutron data for JENDL-3.3. The 15 of the presented papers are indexed individually. (J.P.N.)
Measuring thermal neutron characteristics
International Nuclear Information System (INIS)
Johnstone, C.W.; Jacobson, L.A.
1983-01-01
A method for providing a background-compensated measurement of the level of inducted radiation within an earth formation is claimed. The formation is irradiated with a discrete burst of neutrons and the level of radiation in the formation measured. The level of background radiation is then measured. An average level of both measurements is obtained
Leakage monitoring equipment of fuel element by delayed neutron method
International Nuclear Information System (INIS)
Ji Changsong; Zhang Shulan; Zhang Shuheng
1999-01-01
Based on monitoring results of delayed neutrons from reactor first circle water, the leakage of reactor fuel elements is monitored. A monitoring equipment consisted of an array of 3 He proportional counter tubes with 75 s delay has been developed. The neutron detection efficiency of 6.1% is obtained
An experimental facility for studying delayed neutron emission
International Nuclear Information System (INIS)
Dermendzhiev, E.; Nazarov, V.M.; Pavlov, S.S.; Ruskov, Iv.; Zamyatin, Yu.S.
1993-01-01
A new experimental facility for studying delayed neutron emission has been designed and tested. A method based on utilization of the Dubna IBR-2 pulsed reactor, has been proposed and realized for periodical irradiation of targets composed of fissionable isotopes. Such a powerful pulsed neutron source in combination with a slow neutron chopper synchronized with the reactor bursts makes possible variation of the exposure duration and effective suppression of the fast neutron background due to delay neutrons emitted from the reactor core. Detection of delayed neutrons from the target is carried out by a high-efficiency multicounter neutron detector with a near-4π geometry. Some test measurements and results are briefly described. Possible use of the facility for other tasks is also discussed. 14 refs.; 14 figs
8-group relative delayed neutron yields for monoenergetic neutron induced fission of 239Pu
International Nuclear Information System (INIS)
Piksaikin, V.M.; Kazakov, L.E.; Isaev, S.G.; Korolev, G.G.; Roshchenko, V.A.; Tertychnyj, R.G
2002-01-01
The energy dependence of the relative yield of delayed neutrons in an 8-group model representation was obtained for monoenergetic neutron induced fission of 239 Pu. A comparison of this data with the available experimental data by other authors was made in terms of the mean half-life of the delayed neutron precursors. (author)
Automated uranium analysis by delayed-neutron counting
International Nuclear Information System (INIS)
Kunzendorf, H.; Loevborg, L.; Christiansen, E.M.
1980-10-01
Automated uranium analysis by fission-induced delayed-neutron counting is described. A short description is given of the instrumentation including transfer system, process control, irradiation and counting sites, and computer operations. Characteristic parameters of the facility (sample preparations, background, and standards) are discussed. A sensitivity of 817 +- 22 counts per 10 -6 g U is found using irradiation, delay, and counting times of 20 s, 5 s, and 10 s, respectively. Presicion is generally less than 1% for normal geological samples. Critical level and detection limits for 7.5 g samples are 8 and 16 ppb, respectively. The importance of some physical and elemental interferences are outlined. Dead-time corrections of measured count rates are necessary and a polynomical expression is used for count rates up to 10 5 . The presence of rare earth elements is regarded as the most important elemental interference. A typical application is given and other areas of application are described. (auther)
Gamma/neutron competition above the neutron separation energy in delayed neutron emitters
Directory of Open Access Journals (Sweden)
Valencia E.
2014-03-01
Full Text Available To study the β-decay properties of some well known delayed neutron emitters an experiment was performed in 2009 at the IGISOL facility (University of Jyväskylä in Finland using Total Absorption γ-ray Spectroscopy (TAGS technique. The aim of these measurements is to obtain the full β-strength distribution below the neutron separation energy (Sn and the γ/neutron competition above. This information is a key parameter in nuclear technology applications as well as in nuclear astrophysics and nuclear structure. Preliminary results of the analysis show a significant γ-branching ratio above Sn.
Modeling delayed neutron monitoring systems for fast breeder reactors
International Nuclear Information System (INIS)
Bunch, W.L.; Tang, E.L.
1983-10-01
The purpose of the present work was to develop a general expression relating the count rate of a delayed neutron monitoring system to the introduction rate of fission fragments into the sodium coolant of a fast breeder reactor. Most fast breeder reactors include a system for detecting the presence of breached fuel that permits contact between the sodium coolant and the mixed oxide fuel. These systems monitor for the presence of fission fragments in the sodium that emit delayed neutrons. For operational reasons, the goal is to relate the count rate of the delayed neutron monitor to the condition of the breach in order that appropriate action might be taken
Beta-delayed neutron decay of $^{33}$Na
Radivojevic, Z; Caurier, E; Cederkäll, J; Courtin, S; Dessagne, P; Jokinen, A; Knipper, A; Le Scornet, G; Lyapin, V G; Miehé, C; Nowacki, F; Nummela, S; Oinonen, M; Poirier, E; Ramdhane, M; Trzaska, W H; Walter, G; Äystö, J
2002-01-01
Beta-delayed neutron decay of /sup 33/Na has been studied using the on-line mass separator ISOLDE. The delayed neutron spectra were measured by time-of-flight technique using fast scintillators. Two main neutron groups at 800(60) and 1020(80) keV were assigned to the /sup 33/Na decay, showing evidence for strong feeding of states at about 4 MeV in /sup 33/Mg. By simultaneous beta - gamma -n counting the delayed neutron emission probabilities P/sub 1n/ = 47(6)% and P /sub 2n/ = 13(3)% were determined. The half-life value for /sup 33 /Na, T/sub 1/2/ = 8.0(3) ms, was measured by three different techniques, one employing identifying gamma transitions and two employing beta and neutron counting. (21 refs).
Information about the new 8-group delayed neutron set preparation
International Nuclear Information System (INIS)
Svarny, J.
1998-01-01
Some comments to the present state concerning delayed neutron data preparation is given and preliminary analysis of the new 8-group delayed data (relative abundances) is presented. Comparisons of the 8-group to 6-group set is given for rod drop experiment (Unit 1, Cycle 14, NPP Dukovany).(Author)
Delayed neutron emission near the shell-closures
Directory of Open Access Journals (Sweden)
Borzov Ivan
2016-01-01
Full Text Available The self-consistent Density Functional + Continuum QRPA approach (DF+CQRPA provides a good description of the recent experimental beta-decay half-lives and delayed neutron emission branchings for the nuclei approaching to (and beyond the neutron closed shells N = 28; 50; 82. Predictions of beta-decay properties are more reliable than the ones of standard global approaches traditionally used for the r-process modelling. An impact of the quasi-particle phonon coupling on the delayed multi-neutron emission rates P2n, P3n,… near the closed shells is also discussed.
Delayed neutrons as a probe of nuclear charge distribution in fission of heavy nuclei by neutrons
Isaev, S G; Piksaikin, V M; Roshchenko, V A
2001-01-01
A method of the determination of cumulative yields of delayed neutron precursors is developed. This method is based on the iterative least-square procedure applied to delayed neutron decay curves measured after irradiation of sup 2 sup 3 sup 5 U sample by thermal neutrons. Obtained cumulative yields in turns were used for deriving the values of the most probable charge in low-energy fission of the above-mentioned nucleus.
Delayed neutrons as a probe of nuclear charge distribution in fission of heavy nuclei by neutrons
International Nuclear Information System (INIS)
Isaev, S.G.; Piksaikin, V.M.; Kazakov, L.E.; Roshchenko, V.A.
2002-01-01
A method of the determination of cumulative yields of delayed neutron precursors is developed. This method is based on the iterative least-square procedure applied to delayed neutron decay curves measured after irradiation of 235 U sample by thermal neutrons. Obtained cumulative yields in turns were used for deriving the values of the most probable charge in low-energy fission of the above-mentioned nucleus. (author)
A possible island of beta-delayed neutron precursors in heavy nucleus region
International Nuclear Information System (INIS)
Zhang Li
1991-01-01
The possible Beta-Delayed neutron precursors in the elements Tl, Hg, and Au were predicted following a systematic research on the known Beta-Delayed neutron precursors. The masses of the unknown nuclei and neutron emission probabilities were calculated
Importance of delayed neutron data in transmutation system
International Nuclear Information System (INIS)
Tsujimoto, Kazufumi
1999-01-01
The accelerator-driven transmutation system has been studied at the Japan Atomic Energy Research Institute. This system is a hybrid system which consists of a high intensity accelerator, a spallation target and a subcritical core region. The subcritical core is driven by neutrons generated by spallation reaction in the target region. There is no control rod in this system, so the power is controlled only by proton beam current. The beam current to keep constant power change with effective multiplication factor of subcritical core. So, the evaluation of delayed neutron fraction which is strongly connected to the measurement of subcritical level is important factor in operation of accelerator-driven system. In this paper, important nuclides for the delayed neutron fraction of ADS will be discussed, moreover, present state of delayed neutron data in evaluated nuclear data library is presented. (author)
International Nuclear Information System (INIS)
Doroshenko, A.Yu.; Tarasko, M.Z.; Piksaikin, V.M.
2002-01-01
The energy spectrum of the delayed neutrons is the poorest known of all input data required in the calculation of the effective delayed neutron fractions. In addition to delayed neutron spectra based on the aggregate spectrum measurements there are two different approaches for deriving the delayed neutron energy spectra. Both of them are based on the data related to the delayed neutron spectra from individual precursors of delayed neutrons. In present work these two different data sets were compared with the help of an approximation by gamma-function. The choice of this approximation function instead of the Maxwellian or evaporation type of distribution is substantiated. (author)
MONSTER: a TOF Spectrometer for beta-delayed Neutron Spetroscopy
Martinez, T; Castilla, J; Garcia, A R; Marin, J; Martinez, G; Mendoza, E; Santos, C; Tera, F; Jordan, M D; Rubio, B; Tain, J L; Bhattacharya, C; Banerjee, K; Bhattacharya, S; Roy, P; Meena, J K; Kundu, S; Mukherjee, G; Ghosh, T K; Rana, T K; Pandey, R; Saxena, A; Behera, B; Penttila, H; Jokinen, A; Rinta-Antila, S; Guerrero, C; Ovejero, M C; Villamarin, D; Agramunt, J; Algora, A
2014-01-01
Beta-delayed neutron (DN) data, including emission probabilities, P-n, and energy spectrum, play an important role in our understanding of nuclear structure, nuclear astrophysics and nuclear technologies. A MOdular Neutron time-of-flight SpectromeTER (MONSTER) is being built for the measurement of the neutron energy spectra and branching ratios. The TOF spectrometer will consist of one hundred liquid scintillator cells covering a significant solid angle. The MONSTER design has been optimized by using Monte Carlo (MC) techniques. The response function of the MONSTER cell has been characterized with mono-energetic neutron beams and compared to dedicated MC simulations.
International Nuclear Information System (INIS)
Talamo, A.; Gohar, Y.; Aliberti, G.; Zhong, Z.; Bournos, V.; Fokov, Y.; Kiyavitskaya, H.; Routkovskaya, C.; Serafimovich, I.
2010-01-01
In 1997, Bretscher calculated the effective delayed neutron fraction by the k-ratio method. The Bretscher's approach is based on calculating the multiplication factor of a nuclear reactor core with and without the contribution of delayed neutrons. The multiplication factor set by the delayed neutrons (the delayed multiplication factor) is obtained as the difference between the total and the prompt multiplication factors. Bretscher evaluated the effective delayed neutron fraction as the ratio between the delayed and total multiplication factors (therefore the method is often referred to as k-ratio method). In the present work, the k-ratio method is applied by deterministic nuclear codes. The ENDF/B nuclear data library of the fuel isotopes ( 238 U and 238 U) have been processed by the NJOY code with and without the delayed neutron data to prepare multigroup WIMSD nuclear data libraries for the DRAGON code. The DRAGON code has been used for preparing the PARTISN macroscopic cross sections. This calculation methodology has been applied to the YALINA-Thermal assembly of Belarus. The assembly has been modeled and analyzed using PARTISN code with 69 energy groups and 60 different material zones. The deterministic and Monte Carlo results for the effective delayed neutron fraction obtained by the k-ratio method agree very well. The results also agree with the values obtained by using the adjoint flux. (authors)
Some properties of zero power neutron noise in a time-varying medium with delayed neutrons
International Nuclear Information System (INIS)
Kitamura, Y.; Pal, L.; Pazsit, I.; Yamamoto, A.; Yamane, Y.
2008-01-01
The temporal evolution of the distribution of the number of neutrons in a time-varying multiplying system, producing only prompt neutrons, was treated recently with the master equation technique by some of the present authors. Such a treatment gives account of both the so-called zero power reactor noise and the power reactor noise simultaneously. In particular, the first two moments of the neutron number, as well as the concept of criticality for time-varying systems, were investigated and discussed. The present paper extends these investigations to the case when delayed neutrons are also taken into account. Due to the complexity of the description, only the expectation of the neutron number is calculated. The concept of criticality of a time-varying system is also generalized to systems with delayed neutrons. The temporal behaviour of the expectation of the number of neutrons and its asymptotic properties are displayed and discussed
Method and apparatus for measuring thermal neutron characteristics
International Nuclear Information System (INIS)
Johnstone, C.W.
1983-01-01
The thermal neutron decay characteristics of an earth formation are measured by detecting indications of the thermal neutron concentration in the formation during a selected set of two measurement intervals following irradiation of the formation with a burst of fast neutrons. These measurement intervals may comprise a sequence of time gates following a delay after the neutron burst. The duration of the neutron bursts, of the delay between the burst and the start of the sequence, and of the individual time gates, may all be adjusted by a common, selected one of a finite number of scale factor values. The set of two measurement intervals is selected from among a number of possible sets as a function of a previously measured value of the decay characteristic. Each measurement interval set is used over only a specific range of decay characteristic values for which it has been determined, in accordance with a previously established relationship between the decay characteristic value and a function of the thermal neutron concentration measurements for the set, to afford enhanced statistical accuracy in the measured value of the decay characteristic. (author)
Kalman filter analysis of delayed neutron nondestructive assay measurements
International Nuclear Information System (INIS)
Aumeier, S. E.
1998-01-01
The ability to nondestructively determine the presence and quantity of fissile and fertile nuclei in various matrices is important in several nuclear applications including international and domestics safeguards, radioactive waste characterization and nuclear facility operations. Material irradiation followed by delayed neutron counting is a well known and useful nondestructive assay technique used to determine the fissile-effective content of assay samples. Previous studies have demonstrated the feasibility of using Kalman filters to unfold individual isotopic contributions to delayed neutron measurements resulting from the assay of mixes of uranium and plutonium isotopes. However, the studies in question used simulated measurement data and idealized parameters. We present the results of the Kalman filter analysis of several measurements of U/Pu mixes taken using Argonne National Laboratory's delayed neutron nondestructive assay device. The results demonstrate the use of Kalman filters as a signal processing tool to determine the fissile and fertile isotopic content of an assay sample from the aggregate delayed neutron response following neutron irradiation
A general formula considering one group delayed neutron under nonequilibrium condition
International Nuclear Information System (INIS)
Li Haofeng; Chen Wenzhen; Zhu Qian; Luo Lei
2008-01-01
A general neutron breeder formula is developed when the reactor does not reach the steady state and the reactivity changes in phase. This formula can be used to calculate the results of six groups delayed neutron model through a way of amending λ in one group delayed neutron model. The analysis shows that the solution of amended single group delayed neutron model is approximately equal to that of six-group delayed neutron model, and the amended model meets the engineering accuracy. (authors)
Calibration of the JET neutron yield monitors using the delayed neutron counting technique
International Nuclear Information System (INIS)
van Belle, P.; Jarvis, O.N.; Sadler, G.; de Leeuw, S.; D'Hondt, P.; Pillon, M.
1990-01-01
The time-resolved neutron yield is routinely measured on the JET tokamak using a set of fission chambers. At present, the preferred technique is to employ activation reactions to determine the neutron fluence at a well-chosen position and to relate the measured fluence to the total neutron emission by means of neutron transport calculations. The delayed neutron counting method is a particularly convenient method of performing the activation measurement and the fission cross sections are accurately known. This paper outlines the measurement technique as used on JET
Bioassay method for Uranium in urine by Delay Neutron counting
International Nuclear Information System (INIS)
Suratman; Purwanto; Sukarman-Aminjoyo
1996-01-01
A bioassay method for uranium in urine by neutron counting has been studied. The aim of this research is to obtain a bioassay method for uranium in urine which is used for the determination of internal dose of radiation workers. The bioassay was applied to the artificially uranium contaminated urine. The weight of the contaminant was varied. The uranium in the urine was irradiated in the Kartini reactor core, through pneumatic system. The delayed neutron was counted by BF3 neutron counter. Recovery of the bioassay was between 69.8-88.8 %, standard deviation was less than 10 % and the minimum detection was 0.387 μg
Statistical precision of delayed-neutron nondestructive assay techniques
International Nuclear Information System (INIS)
Bayne, C.K.; McNeany, S.R.
1979-02-01
A theoretical analysis of the statistical precision of delayed-neutron nondestructive assay instruments is presented. Such instruments measure the fissile content of nuclear fuel samples by neutron irradiation and delayed-neutron detection. The precision of these techniques is limited by the statistical nature of the nuclear decay process, but the precision can be optimized by proper selection of system operating parameters. Our method is a three-part analysis. We first present differential--difference equations describing the fundamental physics of the measurements. We then derive and present complete analytical solutions to these equations. Final equations governing the expected number and variance of delayed-neutron counts were computer programmed to calculate the relative statistical precision of specific system operating parameters. Our results show that Poisson statistics do not govern the number of counts accumulated in multiple irradiation-count cycles and that, in general, maximum count precision does not correspond with maximum count as first expected. Covariance between the counts of individual cycles must be considered in determining the optimum number of irradiation-count cycles and the optimum irradiation-to-count time ratio. For the assay system in use at ORNL, covariance effects are small, but for systems with short irradiation-to-count transition times, covariance effects force the optimum number of irradiation-count cycles to be half those giving maximum count. We conclude that the equations governing the expected value and variance of delayed-neutron counts have been derived in closed form. These have been computerized and can be used to select optimum operating parameters for delayed-neutron assay devices
Neutron irradiation facility and its characteristics
International Nuclear Information System (INIS)
Oyama, Yukio; Noda, Kenji
1995-01-01
A neutron irradiation facility utilizing spallation reactions with high energy protons is conceived as one of the facilities in 'Proton Engineering center (PEC)' proposed at JAERI. Characteristics of neutron irradiation field of the facility for material irradiation studies are described in terms of material damage parameters, influence of the pulse irradiation, irradiation environments other than neutronics features, etc., comparing with the other sorts of neutron irradiation facilities. Some perspectives for materials irradiation studies using PEC are presented. (author)
Study of beta-delayed neutron with proton-neutron QRPA plus statistical model
International Nuclear Information System (INIS)
Minato, Futoshi; Iwamoto, Osamu
2015-01-01
β-delayed neutron is known to be important for safety operation of nuclear reactor and prediction of elemental abundance after freeze-out of r-process. A lot of researches on it have been performed. However, the experimental data are far from complete since the lifetime of most of the relevant nuclei is so short that one cannot measure in a high efficiency. In order to estimate half-lives and delayed neutron emission probabilities of unexplored nuclei, we developed a new theoretical method which combines a proton-neutron quasi-particle random-phase-approximation and the Hauser-Feshbach statistical model. The present method reproduces experimentally known β-decay half-lives within a factor of 10 and about 40% of within a factor of 2. However it fails to reproduce delayed neutron emission probabilities. We discuss the problems and remedy for them to be made in future. (author)
Experimental methods of effective delayed neutron fraction
International Nuclear Information System (INIS)
Yamaye, Yoshihiro
1995-01-01
The defining principle and examples of β eff measurement method: the substitutional method, Cf neutron source method, Bennett method, the coupling coefficient method and Nelson method were introduced and surveyed. Measurement errors and C/E value of the substitutional, Cf ray source and Bennett method were of the order of 3%, 5% and 3 - 6% and 0.903 - 0.965, 1.85 and 1.019 - 1.165, respectably. Evaluation of the absolute value is so hard that β eff measurement belongs to the difficult experiment. The dependence on nuclear calculation in decreasing order is the substitutional, Cf ray source, Bennett, the coupling coefficient and Nelson number method. If good substitute materials were selected, the substitutional method has possibility to determine β eff by small correction value or independent on calculation. (S.Y.)
Delayed neutron spectra from short pulse fission of uranium-235
International Nuclear Information System (INIS)
Atwater, H.F.; Goulding, C.A.; Moss, C.E.; Pederson, R.A.; Robba, A.A.; Wimett, T.F.; Reeder, P.; Warner, R.
1986-01-01
Delayed neutron spectra from individual short pulse (∼50 μs) fission of small 235 U samples (50 mg) were measured using a small (5 cm OD x 5 cm length) NE 213 neutron spectrometer. The irradiating fast neutron flux (∼10 13 neutrons/cm 2 ) for these measurements was provided by the Godiva fast burst reactor at the Los Alamos Critical Experiment Facility (LACEF). A high speed pneumatic transfer system was used to transfer the 50 mg 235 U samples from the irradiation position near the Godiva assembly to a remote shielded counting room containing the NE 213 spectrometer and associated electronics. Data were acquired in sixty-four 0.5 s time bins and over an energy range 1 to 7 MeV. Comparisons between these measurements and a detailed model calculation performed at Los Alamos is presented
Subcritical Neutron Multiplication Measurements of HEU Using Delayed Neutrons as the Driving Source
International Nuclear Information System (INIS)
Hollas, C.L.; Goulding, C.A.; Myers, W.L.
1999-01-01
A new method for the determination of the multiplication of highly enriched uranium systems is presented. The method uses delayed neutrons to drive the HEU system. These delayed neutrons are from fission events induced by a pulsed 14-MeV neutron source. Between pulses, neutrons are detected within a medium efficiency neutron detector using 3 He ionization tubes within polyethylene enclosures. The neutron detection times are recorded relative to the initiation of the 14-MeV neutron pulse, and subsequently analyzed with the Feynman reduced variance method to extract singles, doubles and triples neutron counting rates. Measurements have been made on a set of nested hollow spheres of 93% enriched uranium, with mass values from 3.86 kg to 21.48 kg. The singles, doubles and triples counting rates for each uranium system are compared to calculations from point kinetics models of neutron multiplicity to assign multiplication values. These multiplication values are compared to those from MC NP K-Code calculations
The neutron long counter NERO for studies of β-delayed neutron emission in the r-process
International Nuclear Information System (INIS)
Pereira, J.; Hosmer, P.; Lorusso, G.; Santi, P.; Couture, A.; Daly, J.; Del Santo, M.; Elliot, T.
2010-01-01
The neutron long counter NERO was built at the National Superconducting Cyclotron Laboratory (NSCL), Michigan State University, for measuring β-delayed neutron-emission probabilities. The detector was designed to work in conjunction with a β-delay implantation station, so that β decays and β-delayed neutrons emitted from implanted nuclei can be measured simultaneously. The high efficiency of about 40%, for the range of energies of interest, along with the small background, are crucial for measuring β-delayed neutron emission branchings for neutron-rich r-process nuclei produced as low intensity fragmentation beams in in-flight separator facilities.
Review of experimental methods for evaluating effective delayed neutron fraction
Energy Technology Data Exchange (ETDEWEB)
Yamane, Yoshihiro [Nagoya Univ. (Japan). School of Engineering
1997-03-01
The International Effective Delayed Neutron Fraction ({beta}{sub eff}) Benchmark Experiments have been carried out at the Fast Critical Assembly of Japan Atomic Energy Research Institute since 1995. Researchers from six countries, namely France, Italy, Russia, U.S.A., Korea, and Japan, participate in this FCA project. Each team makes use of each experimental method, such as Frequency Method, Rossi-{alpha} Method, Nelson Number Method, Cf Neutron Source Method, and Covariance Method. In this report these experimental methods are reviewed. (author)
Neutron sources and their characteristics
International Nuclear Information System (INIS)
McCall, R.C.; Swanson, W.P.
1979-03-01
The significant sources of photoneutrons within a linear-accelerator treatment head are identified and absolute estimates of neutron production per treatment dose are given for typical components. It is found that the high-Z materials within the treatment head do not significantly alter the neutron fluence but do substantially reduce the average energy of the transmitted spectrum. Reflection of neutrons from the concrete treatment room contribute to the neutron fluence, but not substantially to the patient integral dose, because of a further reduction in average energy. The ratio of maximum fluence to the treatment dose at the same distance is given as a function of electron energy. This ratio rises with energy to an almost constant value of 2.1 x 10 5 neutrons cm -2 rad -1 at electron energies above about 25 MeV. Measured data obtained at a variety of accelerator installations are presented and compared with these calculations. Reasons for apparent deviations are suggested. Absolute depth-dose and depth-dose-equivalent distributions for realistic neutron spectra that occur at therapy installations are calculated, and a rapid falloff with depth is found. The ratio of neutron integral absorbed dose to leakage photon absorbed dose is estimated to be 0.04 and 0.2 for 14 to 25 MeV incident electron energy, respectively. Possible reasons are given for lesser neutron production from betatrons than from linear accelerators. Possible ways in which neutron production can be reduced are discussed
8-group relative delayed neutron yields for epithermal neutron induced fission of 235U and 239Pu
International Nuclear Information System (INIS)
Piksaikin, V.M.; Kazakov, L.E.; Isaev, S.G.; Korolev, G.G.; Roshchenko, V.A.; Tertychnyj, R.G
2002-01-01
An 8-group representation of relative delayed neutron yields was obtained for epithermal neutron induced fission of 235 U and 239 Pu. These data were compared with ENDF/B-VI data in terms of the average half- life of the delayed neutron precursors and on the basis of the dependence of reactivity on the asymptotic period. (author)
Some characteristics of a miniature neutron spectrometer
International Nuclear Information System (INIS)
Sekimoto, H.; Oishi, K.; Hojo, K.; Hojo, T.
1984-01-01
Some characteristics of an NE213 miniature spherical spectrometer for in-assembly fast-neutron spectrometry were measured. As the bubbling time changed, the pulse-height did not change appreciably, but the n-γ discrimination characteristics changed considerably. As the count rate changed, the pulse-height did not change appreciably, and the change of the n-γ discrimination characteristics was acceptable. The neutron response function was measured to be almost isotropic except for the backward direction. (orig.)
Calibration of the delayed-gamma neutron activation facility
International Nuclear Information System (INIS)
Ma, R.; Zhao, X.; Rarback, H.M.; Yasumura, S.; Dilmanian, F.A.; Moore, R.I.; Lo Monte, A.F.; Vodopia, K.A.; Liu, H.B.; Economos, C.D.; Nelson, M.E.; Aloia, J.F.; Vaswani, A.N.; Weber, D.A.; Pierson, R.N. Jr.; Joel, D.D.
1996-01-01
The delayed-gamma neutron activation facility at Brookhaven National Laboratory was originally calibrated using an anthropomorphic hollow phantom filled with solutions containing predetermined amounts of Ca. However, 99% of the total Ca in the human body is not homogeneously distributed but contained within the skeleton. Recently, an artificial skeleton was designed, constructed, and placed in a bottle phantom to better represent the Ca distribution in the human body. Neutron activation measurements of an anthropomorphic and a bottle (with no skeleton) phantom demonstrate that the difference in size and shape between the two phantoms changes the total body calcium results by less than 1%. To test the artificial skeleton, two small polyethylene jerry-can phantoms were made, one with a femur from a cadaver and one with an artificial bone in exactly the same geometry. The femur was ashed following the neutron activation measurements for chemical analysis of Ca. Results indicate that the artificial bone closely simulates the real bone in neutron activation analysis and provides accurate calibration for Ca measurements. Therefore, the calibration of the delayed-gamma neutron activation system is now based on the new bottle phantom containing an artificial skeleton. This change has improved the accuracy of measurement for total body calcium. Also, the simple geometry of this phantom and the artificial skeleton allows us to simulate the neutron activation process using a Monte Carlo code, which enables us to calibrate the system for human subjects larger and smaller than the phantoms used as standards. copyright 1996 American Association of Physicists in Medicine
Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core
International Nuclear Information System (INIS)
Lashkari, A.; Khalafi, H.; Kazeminejad, H.
2013-01-01
Highlights: ► Kinetic parameters of Tehran research reactor mixed-core have been calculated. ► Burn-up effect on TRR kinetics parameters has been studied. ► Replacement of LEU-CFE with HEU-CFE in the TRR core has been investigated. ► Results of each mixed core were compared to the reference core. ► Calculation of kinetic parameters are necessary for reactivity and power excursion transient analysis. - Abstract: In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR P C package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change
MODICO, 1-D Time-Dependent 1 Group, 2 Group Neutron Diffusion with Delayed Neutron Precursors
International Nuclear Information System (INIS)
Camiciola, P.; Cundari, D.; Montagnini, B.
1992-01-01
1 - Description of program or function: The program solves the 1-D time-dependent one and two group coarse-mesh neutron diffusion equations, coupled with the equations for the delayed-neutron precursor, in plane geometry. 2 - Method of solution: The program is based on a simple coarse-mesh cubic approximation formula for the spatial behaviour of the flux inside each interval. An implicit scheme (the time-integrated method) is used for the advancement of the solution. The resulting (block three-diagonal) matrix is inverted at each time step by Thomas' method. 3 - Restrictions on the complexity of the problem: Number of coarse- mesh intervals LE 80; number of material regions LE 10; number of delayed-neutron precursor groups LE 10. Typical mesh sizes range from 5 cm to 20 cm; typical step length (non-prompt critical transients) ranges from 0.005 to 0.1 seconds
International Nuclear Information System (INIS)
Anon.
1981-01-01
The method covers the detection and measurement of delayed neutron-emitting fission products contained in nuclear reactor coolant water while the reactor is operating. The method is limited to the measurement of the delayed neutron-emitting bromine isotope of mass 87 and the delayed neutron-emitting iodine isotope of mass 137. The other delayed neutron-emitting fission products cannot be accurately distinguished from nitrogen 17, which is formed under some reactor conditions by neutron irradiation of the coolant water molecules. The method includes a description of significance, measurement variables, interferences, apparatus, sampling, calibration, standardization, sample measurement procedures, system efficiency determination, calculations, and precision
An accurate solution of point reactor neutron kinetics equations of multi-group of delayed neutrons
International Nuclear Information System (INIS)
Yamoah, S.; Akaho, E.H.K.; Nyarko, B.J.B.
2013-01-01
Highlights: ► Analytical solution is proposed to solve the point reactor kinetics equations (PRKE). ► The method is based on formulating a coefficient matrix of the PRKE. ► The method was applied to solve the PRKE for six groups of delayed neutrons. ► Results shows good agreement with other traditional methods in literature. ► The method is accurate and efficient for solving the point reactor kinetics equations. - Abstract: The understanding of the time-dependent behaviour of the neutron population in a nuclear reactor in response to either a planned or unplanned change in the reactor conditions is of great importance to the safe and reliable operation of the reactor. In this study, an accurate analytical solution of point reactor kinetics equations with multi-group of delayed neutrons for specified reactivity changes is proposed to calculate the change in neutron density. The method is based on formulating a coefficient matrix of the homogenous differential equations of the point reactor kinetics equations and calculating the eigenvalues and the corresponding eigenvectors of the coefficient matrix. A small time interval is chosen within which reactivity relatively stays constant. The analytical method was applied to solve the point reactor kinetics equations with six-groups delayed neutrons for a representative thermal reactor. The problems of step, ramp and temperature feedback reactivities are computed and the results compared with other traditional methods. The comparison shows that the method presented in this study is accurate and efficient for solving the point reactor kinetics equations of multi-group of delayed neutrons
Study of $\\beta$-delayed neutron decay of $^{8}$He
The goal of the present proposal is to study $\\beta$-delayed neutron decay branch of $^{8}$He. The energy spectra of the emitted neutrons will be measured in the energy range of 0.1 – 6 MeV using the VANDLE spectrometer. Using coincident $\\gamma$-ray measurement, components of the spectrum corresponding to transitions to the ground- and first- excited states of $^{7}$Li will be disentangled. The new data will allow us to get a more complete picture of the $\\beta$-decay of $^{8}$He and to clarify the discrepancy between the B(GT) distributions derived from the $\\beta$-decay and $^{8}$He(p, n)$^{8}$Li reaction studies.
Proceedings of the specialists' meeting on delayed neutron nuclear data
Energy Technology Data Exchange (ETDEWEB)
Katakura, Jun-ichi [ed.] [Japanese Nuclear Data Committee, Tokai, Ibaraki (Japan)
1999-07-01
This report is the Proceedings of the Specialists' Meeting on Delayed Neutron Nuclear Data. The meeting was held on January 28-29, 1999, at the Tokai Research Establishment of Japan Atomic Energy Research Institute with the participation of thirty specialists, who are evaluators, theorist, experimentalists. Although the fraction of the delayed neutron is no more than 1% in the total neutrons emitted in the fission process, it plays an important roll in the control of fission reactor. In the meeting, the following topics were reported: the present status of delayed neutron data in the major evaluated data libraries, measurements of effective delayed neutron fraction using FCA (Fast Critical Assembly) and TCA (Tank-type Critical Assembly) and their analyses, sensitivity analysis for fast reactor, measurements of delayed neutron emission from actinides and so on. As another topics, delayed neutron in transmutation system and fission yield data were also presented. Free discussion was held on the future activity of delayed neutron data evaluation. The discussion was helpful for the future activity of the delayed neutron working group of JNDC aiming to the evaluation of delayed neutron data for JENDL-3.3. The 15 of the presented papers are indexed individually. (J.P.N.)
Rapid uranium analysis by delayed neutron counting of neutron activated samples
International Nuclear Information System (INIS)
Papadopoulos, N.N.
1985-01-01
The uranium analyzer at the Nuclear Research Center ''Demokritos'' and the delayed neutron method have been used to determine the uranium content in lignite, in chemically enriched samples and in solutions of extractable uranium. The results are compared with data obtained by other methods. In the case of dissolved extractable uranium. The results are in good agreement with X-ray fluorescence data in the range 100 ppm to 2000 ppm while beyond these limits the discrepancies between neutron and spectrophotometric data are observed. The results for lignite samples are in good agreement with gamma spectrometric data. Discrepancies indicate that more extensive intercomparisons are needed to check the reliability of various methods
International Nuclear Information System (INIS)
Piksaikin, V.M.; Isaev, S.G.; Goverdovski, A.A.; Pshakin, G.M.
1998-10-01
The document includes the following two reports: 'Correlation properties of delayed neutrons from fast neutron induced fission' and 'Method and set-up for measurements of trace level content of heavy fissionable elements based on delayed neutron counting. A separate abstract was prepared for each report
An In-Pile Kinetic Method for Determining the Delayed Neutron Fraction βeff
International Nuclear Information System (INIS)
Gilad, E.; Rivin, O.; Ettedgui, H.; Yaar, I.; Geslot, B.; Pepino, A.; Di Salvo, J.; Gruel, A.; Blaise, P.
2014-01-01
Delayed neutrons are of fundamental importance in the field of nuclear reactor dynamics and control. Although only a small fraction of the neutrons emitted by fission are not prompt, the knowledge of the delayed neutrons parameters is essential for transient analysis, such as startup or shutdown of the reactor, as well as for accidents analysis and control system design [1]. One of the main delayed neutron parameters used in the point reactor model equations is the effective delayed neutron fraction, which incorporates both delayed neutron spectral properties and core geometrical configuration [1,2]. Additional delayed neutron parameters include the fraction of fission neutrons emitted in each delayed group, and the delayed neutron precursors decay constants . Experimental efforts aimed at determining the value ofβ, which provide experimental support for the evaluation of delayed neutron parameters, are extremely valuable. This is due to the fact that unlike other fields in reactor physics, e.g. criticality safety or shielding, the availability of experimental data and benchmark problems for validating delayed neutron parameters and its implementation in different models is highly limited. Furthermore, the existing experimental data exhibit significant discrepancies between the different sets of parameter, which lead to substantial disparity in the analysis of kinetic experiments and reactor dynamic behavior]. In this work, a method for determining the effective delayed neutron fraction using in-pile reactivity oscillation and Fourier analysis is presented. The method is based on measurements of the reactor's power response to small periodic in-pile reactivity perturbations and utilizes Fourier analysis for reconstruction of the reactor zero power transfer function. Knowledge of the reactor transfer function enables the estimation of theβ value using multi-parameter nonlinear fit. The method accounts for higher harmonics, which are excited by the trapezoidal
Study on calculation methods for the effective delayed neutron fraction
International Nuclear Information System (INIS)
Irwanto, Dwi; Obara, Toru; Chiba, Go; Nagaya, Yasunobu
2011-03-01
The effective delayed neutron fraction β eff is one of the important neutronic parameters from a view point of a reactor kinetics. Several Monte-Carlo-based methods to estimate β eff have been proposed to date. In order to quantify the accuracy of these methods, we study calculation methods for β eff by analyzing various fast neutron systems including the bare spherical systems (Godiva, Jezebel, Skidoo, Jezebel-240), the reflective spherical systems (Popsy, Topsy, Flattop-23), MASURCA-R2 and MASURCA-ZONA2, and FCA XIX-1, XIX-2 and XIX-3. These analyses are performed by using SLAROM-UF and CBG for the deterministic method and MVP-II for the Monte Carlo method. We calculate β eff with various definitions such as the fundamental value β 0 , the standard definition, Nauchi's definition and Meulekamp's definition, and compare these results with each other. Through the present study, we find the following: The largest difference among the standard definition of β eff , Nauchi's β eff and Meulekamp's β eff is approximately 10%. The fundamental value β 0 is quite larger than the others in several cases. For all the cases, Meulekamp's β eff is always higher than Nauchi's β eff . This is because Nauchi's β eff considers the average neutron multiplicity value per fission which is large in the high energy range (1MeV-10MeV), while the definition of Meulekamp's β eff does not include this parameter. Furthermore, we evaluate the multi-generation effect on β eff values and demonstrate that this effect should be considered to obtain the standard definition values of β eff . (author)
Response characteristics of selected personnel neutron dosimeters
International Nuclear Information System (INIS)
McDonald, J.C.; Fix, J.J.; Hadley, R.T.; Holbrook, K.L.; Yoder, R.C.; Roberson, P.L.; Endres, G.W.R.; Nichols, L.L.; Schwartz, R.B.
1983-09-01
Performance characteristics of selected personnel neutron dosimeters in current use at Department of Energy (DOE) facilities were determined from their evaluation of neutron dose equivalent received after irradiations with specific neutron sources at either the National Bureau of Standards (NBS) or the Pacific Northwest Laboratory (PNL). The characteristics assessed included: lower detection level, energy response, precision and accuracy. It was found that when all of the laboratories employed a common set of calibrations, the overall accuracy was approximately +-20%, which is within uncertainty expected for these dosimeters. For doses above 80 mrem, the accuracy improved to better than 10% when a common calibration was used. Individual differences found in this study may reflect differences in calibration technique rather than differences in the dose rates of actual calibration standards. Second, at dose rates above 100 mrem, the precision for the best participants was generally below +-10% which is also within expected limits for these types of dosimeters. The poorest results had a standard deviation of about +-25%. At the lowest doses, which were sometimes below the lower detection limit, the precision often approached or exceeded +-100%. Third, the lower level of detection for free field 252 Cf neutrons generally ranged between 20 and 50 mrem. Fourth, the energy dependence study provided a characterization of the response of the dosimeters to neutron energies far from the calibration energy. 11 references, 22 figures, 26 tables
Delaying Developmental Mathematics: The Characteristics and Costs
Johnson, Marianne; Kuennen, Eric
2004-01-01
This paper investigates which students delay taking a required developmental mathematics course and the impact of delay on student performance in introductory microeconomics. Analysis of a sample of 1462 students at a large Midwestern university revealed that, although developmental-level mathematics students did not reach the same level of…
Statistical theory for calculating energy spectra of β-delayed neutrons
International Nuclear Information System (INIS)
Kawano, Toshihiko; Moeller, Peter; Wilson, William B.
2008-01-01
Theoretical β-delayed neutron spectra are calculated based on the Quasi-particle Random Phase Approximation (QRPA) and the Hauser-Feshbach statistical model. Neutron emissions from an excited daughter nucleus after β-decay to the granddaughter residual are more accurately calculated than previous evaluations, including all the microscopic nuclear structure information, such as a Gamow-Teller strength distribution and discrete states in the granddaughter. The calculated delayed-neutron spectra reasonably agree with those evaluations in the ENDF decay library, which are based on experimental data. The model was adopted to generate the delayed-neutron spectra for all 271 precursors. (authors)
Comparison of dynamic compensation methods for delayed self-powered neutron detector
International Nuclear Information System (INIS)
In, Wang Kee; Kim, Joon Sung; Auh, Geun Sun; Yoon, Tae Young
1993-01-01
Dynamic compensation methods for rhodium self-powered neutron detector have been developed by Banda and Hoppe to compensate for the time delay associated with detector signals. The time delay is due to the decay of the neutron-activated rhodium and results in delayed detector response. Two digital dynamic compensation methods, were compared for step change of neutron flux in this paper. The inverse kinetics method gave slightly better response time and noise gain. However, the inverse kinetics method also showed overshooting of neutron flux for the step change. (Author)
Investigation of capture reactions far off stability by β-delayed neutron emission
International Nuclear Information System (INIS)
Wiescher, M.; Leist, B.; Ziegert, W.; Gabelmann, H.; Steinmueller, B.; Ohm, H.; Kratz, K.h.; Thielemann, F.h.; Hillebrandt, W.
1985-01-01
Beta-delayed neutron spectroscopy is applied to determine reaction rates of neutron capture on several neutron rich nuclei. The results of these experiments are presented and discussed in the light of their astrophysical implications. Furthermore, the experimental possibilities and limits of planned measurements are advertised
Energy Technology Data Exchange (ETDEWEB)
Piksaikine, V. [Institute of Physics and Power Engineering, Obninsk (Russian Federation)
1997-03-01
The experimental method for measurements of the delayed neutron yields and period is presented. The preliminary results of the total yield, relative abundances and periods are shown comparing with the previously reported values. (J.P.N.)
Energy Technology Data Exchange (ETDEWEB)
Mayer, M. [Department of Mechanical and Nuclear Engineering, The Pennsylvania State University, University Park, Pennsylvania 16802 (United States); Nattress, J.; Jovanovic, I., E-mail: ijov@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, Michigan 48109 (United States)
2016-06-27
Detection of unique signatures of special nuclear materials is critical for their interdiction in a variety of nuclear security and nonproliferation scenarios. We report on the observation of delayed neutrons from fission of uranium induced in dual-particle active interrogation based on the {sup 11}B(d,n γ){sup 12}C nuclear reaction. Majority of the fissions are attributed to fast fission induced by the incident quasi-monoenergetic neutrons. A Li-doped glass–polymer composite scintillation neutron detector, which displays excellent neutron/γ discrimination at low energies, was used in the measurements, along with a recoil-based liquid scintillation detector. Time-dependent buildup and decay of delayed neutron emission from {sup 238}U were measured between the interrogating beam pulses and after the interrogating beam was turned off, respectively. Characteristic buildup and decay time profiles were compared to the common parametrization into six delayed neutron groups, finding a good agreement between the measurement and nuclear data. This method is promising for detecting fissile and fissionable materials in cargo scanning applications and can be readily integrated with transmission radiography using low-energy nuclear reaction sources.
Analysis of incident-energy dependence of delayed neutron yields in actinides
Energy Technology Data Exchange (ETDEWEB)
Nasir, Mohamad Nasrun bin Mohd, E-mail: monasr211@gmail.com; Metorima, Kouhei, E-mail: kohei.m2420@hotmail.co.jp; Ohsawa, Takaaki, E-mail: ohsawa@mvg.biglobe.ne.jp; Hashimoto, Kengo, E-mail: kengoh@pp.iij4u.or.jp [Graduate School of Science and Engineering, Kindai University, Kowakae, Higashi-Osaka, 577-8502 (Japan)
2015-04-29
The changes of delayed neutron yields (ν{sub d}) of Actinides have been analyzed for incident energy up to 20MeV using realized data of precursor after prompt neutron emission, from semi-empirical model, and delayed neutron emission probability data (P{sub n}) to carry out a summation method. The evaluated nuclear data of the delayed neutron yields of actinide nuclides are still uncertain at the present and the cause of the energy dependence has not been fully understood. In this study, the fission yields of precursor were calculated considering the change of the fission fragment mass yield based on the superposition of fives Gaussian distribution; and the change of the prompt neutrons number associated with the incident energy dependence. Thus, the incident energy dependent behavior of delayed neutron was analyzed.The total number of delayed neutron is expressed as ν{sub d}=∑Y{sub i} • P{sub ni} in the summation method, where Y{sub i} is the mass yields of precursor i and P{sub ni} is the delayed neutron emission probability of precursor i. The value of Y{sub i} is derived from calculation of post neutron emission mass distribution using 5 Gaussian equations with the consideration of large distribution of the fission fragments. The prompt neutron emission ν{sub p} increases at higher incident-energy but there are two different models; one model says that the fission fragment mass dependence that prompt neutron emission increases uniformly regardless of the fission fragments mass; and the other says that the major increases occur at heavy fission fragments area. In this study, the changes of delayed neutron yields by the two models have been investigated.
International Nuclear Information System (INIS)
Artemov, V.G.; Gusev, V.I.; Zinatullin, R.E.; Karpov, A.S.
2007-01-01
Using modeled WWER cram rod drop experiments, performed at the Rostov NPP, as an example, the influence of delayed neutron parameters on the modeling results was investigated. The delayed neutron parameter values were taken from both domestic and foreign nuclear databases. Numerical modeling was carried out on the basis of SAPFIR 9 5andWWERrogram package. Parameters of delayed neutrons were acquired from ENDF/B-VI and BNAB-78 validated data files. It was demonstrated that using delay fraction data from different databases in reactivity meters led to significantly different reactivity results. Based on the results of numerically modeled experiments, delayed neutron parameters providing the best agreement between calculated and measured data were selected and recommended for use in reactor calculations (Authors)
Breached fuel location in FFTF by delayed neutron monitor triangulation
International Nuclear Information System (INIS)
Bunch, W.L.; Tang, E.L.
1985-10-01
The Fast Flux Test Facility (FFTF) features a three-loop, sodium-cooled 400 MWt mixed oxide fueled reactor designed for the irradiation testing of fuels and materials for use in liquid metal cooled fast reactors. To establish the ultimate capability of a particular fuel design and thereby generate information that will lead to improvements, many of the fuel irradiations are continued until a loss of cladding integrity (failure) occurs. When the cladding fails, fission gas escapes from the fuel pin and enters the reactor cover gas system. If the cladding failure permits the primary sodium to come in contact with the fuel, recoil fission products can enter the sodium. The presence of recoil fission products in the sodium can be detected by monitoring for the presence of delayed neutrons in the coolant. It is the present philosophy to not operate FFTF when a failure has occurred that permits fission fragments to enter the sodium. Thus, it is important that the identity and location of the fuel assembly that contains the failed cladding be established in order that it might be removed from the core. This report discusses method of location of fuel element when cladding is breached
Delayed Particle Study of Neutron Rich Lithium Isotopes
Marechal, F; Perrot, F
2002-01-01
We propose to make a systematic complete coincidence study of $\\beta$-delayed particles from the decay of neutron-rich lithium isotopes. The lithium isotopes with A=9,10,11 have proven to contain a vast information on nuclear structure and especially on the formation of halo nuclei. A mapping of the $\\beta$-strength at high energies in the daughter nucleus will make possible a detailed test of our understanding of their structure. An essential step is the comparison of $\\beta$-strength patterns in $^{11}$Li and the core nucleus $^{9}$Li, another is the full characterization of the break-up processes following the $\\beta$-decay. To enable such a measurement of the full decay process we will use a highly segmented detection system where energy and emission angles of both charged and neutral particles are detected in coincidence and with high efficiency and accuracy. We ask for a total of 30 shifts (21 shifts for $^{11}$Li, 9 shifts $^{9}$Li adding 5 shifts for setting up with stable beam) using a Ta-foil target...
Maternal mortality and delay: Socio-demographic characteristics of ...
African Journals Online (AJOL)
This study assessed the contribution of delay to maternal deaths and also determined the socio¬demographic characteristics of patients with maternal deaths with associated delay. Methods: This is a cross-sectional descriptive study of all maternal deaths in Irrua specialist Teaching Hospital, Nigeria between January 1999 ...
Possibilities of delayed neutron fraction (βeff) calculation and measurement
International Nuclear Information System (INIS)
Michalek, S.; Hascik, J.; Farkas, G.
2008-01-01
The influence of the delayed neutrons on the reactor dynamics can be understood through their impact on the reactor power change rate. In spite of the fact that delayed neutrons constitute only a very small fraction of the total number of neutrons generated from fission, they play a dominant role in the fission chain reaction control. If only the prompt neutrons existed, the reactor operation would become impossible due to the fast reactor power changes. The exact determination of delayed neutrons main parameter, the delayed neutron fraction (β eff ), is very important in the field of reactor physics. The interest in the delayed neutron data accuracy improvement started to increase at the end of 80-ties and the beginning of 90-ties, after discrepancies among the results of calculations and experiments. In consequence of difficulties in β eff experimental measurement, this value in exact state use to be determined by calculations. Subsequently, its reliability depends on the calculation method and the delayed neutron data used. Determination of β eff requires criticality calculations. In the past, k eff used to be traditionally calculated by taking the ratio of the adjoint- and spectrum-weighted delayed neutron production rate to the adjoint- and spectrum- weighted total neutron production rate. An alternative method has also been used in which β eff is calculated from simple k-eigenvalue solutions. In this work, a summary of possible β eff calculation methods can be found and a calculation of β eff for VR-1 training reactor in one operation state is made using the prompt method, by MCNP5 code. Also a method of β eff kinetic measurement on VR-1 training reactor at Czech Technical University in Prague using in-pile kinetic technique is outlined (authors)
A study on the linearity characteristics of neutron power measurement system for Hanaro
International Nuclear Information System (INIS)
Kang, Tai Ki; Kim, Young Ki; Lee, Byung Chul; Park, Sang Jun
1999-06-01
It is briefly described the general principles of neutron detection and the method of neutron measurement in the nuclear reactor which neutron flux varies widely and gamma radiation also exists. Wide-range Fission Chamber System which is excellent in electrical and mechanical performances has been selected for neutron power measurement system for Hanaro. The linearity characteristics of neutron power signals is a critical factor of the reliability in reactor power control. In particular , the linearity of the log power signal, which covers 10 decade form 10 -8 %FP to 200 %FP was a matter of primary concern during commissioning. In case of the linear power signal for reactor control at high power condition, the output signals were additionally analyzed in connection with the reactor thermal power and the delayed neutron signal from the primary pipe as well as the output signal from the compensated ion chamber as a reference signal. (author). 13 refs., 7 tabs., 33 figs
A study on the linearity characteristics of neutron power measurement system for Hanaro
Energy Technology Data Exchange (ETDEWEB)
Kang, Tai Ki; Kim, Young Ki; Lee, Byung Chul; Park, Sang Jun
1999-06-01
It is briefly described the general principles of neutron detection and the method of neutron measurement in the nuclear reactor which neutron flux varies widely and gamma radiation also exists. Wide-range Fission Chamber System which is excellent in electrical and mechanical performances has been selected for neutron power measurement system for Hanaro. The linearity characteristics of neutron power signals is a critical factor of the reliability in reactor power control. In particular , the linearity of the log power signal, which covers 10 decade form 10 {sup -8} %FP to 200 %FP was a matter of primary concern during commissioning. In case of the linear power signal for reactor control at high power condition, the output signals were additionally analyzed in connection with the reactor thermal power and the delayed neutron signal from the primary pipe as well asthe output signal from the compensated ion chamber as a reference signal. (author). 13 refs., 7 tabs., 33 figs.
Delayed neutron spectra and their uncertainties in fission product summation calculations
Energy Technology Data Exchange (ETDEWEB)
Miyazono, T.; Sagisaka, M.; Ohta, H.; Oyamatsu, K.; Tamaki, M. [Nagoya Univ. (Japan)
1997-03-01
Uncertainties in delayed neutron summation calculations are evaluated with ENDF/B-VI for 50 fissioning systems. As the first step, uncertainty calculations are performed for the aggregate delayed neutron activity with the same approximate method as proposed previously for the decay heat uncertainty analyses. Typical uncertainty values are about 6-14% for {sup 238}U(F) and about 13-23% for {sup 243}Am(F) at cooling times 0.1-100 (s). These values are typically 2-3 times larger than those in decay heat at the same cooling times. For aggregate delayed neutron spectra, the uncertainties would be larger than those for the delayed neutron activity because much more information about the nuclear structure is still necessary. (author)
Pulse-shape discrimination in radioanalytical methods. Part I. Delayed fission neutron counting
International Nuclear Information System (INIS)
Posta, S.; Vacik, J.; Hnatowicz, V.; Cervena, J.
1999-01-01
In this study the principle of pulse shape discrimination (PSD) has been employed in delayed fission neutron counting (DNC) method. Effective elimination of unwanted gamma background signals in measured radiation spectra has been proved. (author)
International Nuclear Information System (INIS)
Amin, Y.M.; Kamaluddin, B.; Mahat, R.H.
1990-01-01
Limestone stratigraphy in Malaysia has been and is dependent almost entirely in palaeontology. However fossil localities are sporadic and as such a new fossil discovery mean the necessity for a complete re-appraisal of the stratigraphy. The almost complete dependence upon palaeontology results from the difficulties of stratigraphy correlation of isolated outcrops, from the cover of tropical vegetation and from the often complex folding and faulting which has been imposed on the geosyn-clinical rocks by the Indonesian-Thai-Malayan orogeny. So by studying the elemental composition of limestones accurately, we would be able to correlate outcrops and other stratigraphic samples independent of fossil finds. The use of delayed neutron analysis would also determine the concentration of uranium and thorium accurately. This study, in conjunction with thermoluminescence and fission track studies, would able us to date the age of the limestones
Evaluation method for uncertainty of effective delayed neutron fraction βeff
International Nuclear Information System (INIS)
Zukeran, Atsushi
1999-01-01
Uncertainty of effective delayed neutron fraction β eff is evaluated in terms of three quantities; uncertainties of the basic delayed neutron constants, energy dependence of delayed neutron yield ν d m , and the uncertainties of the fission cross sections of fuel elements. The uncertainty of β eff due to the delayed neutron yield is expressed by a linearized formula assuming that the delayed neutron yield does not depend on the incident energy, and the energy dependence is supplemented by using the detailed energy dependence proposed by D'Angelo and Filip. The third quantity, uncertainties of fission cross section, is evaluated on the basis of the generalized perturbation theory in relation to reaction rate rations such as central spectral indexes or average reaction rate ratios. Resultant uncertainty of β eff is about 4 to 5%s, in which primary factor is the delayed neutron yield, and the secondary one is the fission cross section uncertainty, especially for 238 U. The energy dependence of ν d m systematically reduces the magnitude of β eff about 1.4% to 1.7%, depending on the model of the energy vs. ν d m correlation curve. (author)
Beta-delayed gamma and neutron emission near the double shell closure at 78Ni
International Nuclear Information System (INIS)
Rykaczewski, Krzysztof Piotr; Mazzocchi, C.; Grzywacz, R.; Batchelder, J. C.; Bingham, C.R.; Fong, D.; Hamilton, J.H.; Hwang, J.K.; Karny, M.; Krolas, W.; Liddick, S. N.; Morton, A. C.; Mantica, P. F.; Mueller, W. F.; Steiner, M.; Stolz, A.; Winger, J.A.
2005-01-01
An experiment was performed at the National Superconducting Cyclotron Laboratory at Michigan State University to investigate β decay of very neutron-rich cobalt isotopes. Beta-delayed neutron emission from 71-74 Co has been observed for the first time. Preliminary results are reported
New Beta-delayed Neutron Measurements in the Light-mass Fission Group
Energy Technology Data Exchange (ETDEWEB)
Agramunt, J. [Instituto de Física Corpuscular, CSIC-Univ. Valencia, Apdo. Correos 22085, E-46071 Valencia (Spain); García, A.R. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas, E-28040 Madrid (Spain); Algora, A. [Instituto de Física Corpuscular, CSIC-Univ. Valencia, Apdo. Correos 22085, E-46071 Valencia (Spain); Äystö, J. [University of Jyväskylä, FI-40014 Jyväskyä (Finland); Caballero-Folch, R.; Calviño, F. [Secció d' Enginyeria Nuclear, Universitat Politécnica de Catalunya, E-08028 Barcelona (Spain); Cano-Ott, D. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas, E-28040 Madrid (Spain); Cortés, G. [Secció d' Enginyeria Nuclear, Universitat Politécnica de Catalunya, E-08028 Barcelona (Spain); Domingo-Pardo, C. [Instituto de Física Corpuscular, CSIC-Univ. Valencia, Apdo. Correos 22085, E-46071 Valencia (Spain); Eronen, T. [University of Jyväskylä, FI-40014 Jyväskyä (Finland); Gelletly, W. [Department of Physics, University of Surrey, Guildford GU2 7XH (United Kingdom); Gómez-Hornillos, M.B. [Secció d' Enginyeria Nuclear, Universitat Politécnica de Catalunya, E-08028 Barcelona (Spain); and others
2014-06-15
A new accurate determination of beta-delayed neutron emission probabilities from nuclei in the low mass region of the light fission group has been performed. The measurements were carried out using the BELEN 4π neutron counter at the IGISOL-JYFL mass separator in combination with a Penning trap. The new results significantly improve the uncertainties of neutron emission probabilities for {sup 91}Br, {sup 86}As, {sup 85}As, and {sup 85}Ge nuclei.
TDTORT: Time-Dependent, 3-D, Discrete Ordinates, Neutron Transport Code System with Delayed Neutrons
International Nuclear Information System (INIS)
2002-01-01
1 - Description of program or function: TDTORT solves the time-dependent, three-dimensional neutron transport equation with explicit representation of delayed neutrons to estimate the fission yield from fissionable material transients. This release includes a modified version of TORT from the C00650MFMWS01 DOORS3.1 code package plus the time-dependent TDTORT code. GIP is also included for cross-section preparation. TORT calculates the flux or fluence of particles due to particles incident upon the system from extraneous sources or generated internally as a result of interaction with the system in two- or three-dimensional geometric systems. The principle application is to the deep-penetration transport of neutrons and photons. Reactor eigenvalue problems can also be solved. Numerous printed edits of the results are available, and results can be transferred to output files for subsequent analysis. TDTORT reads ANISN-format cross-section libraries, which are not included in the package. Users may choose from several available in RSICC's data library collection which can be identified by the keyword 'ANISN FORMAT'. 2 - Methods:The time-dependent spatial flux is expressed as a product of a space-, energy-, and angle-dependent shape function, which is usually slowly varying in time and a purely time-dependent amplitude function. The shape equation is solved for the shape using TORT; and the result is used to calculate the point kinetics parameters (e.g., reactivity) by using their inner product definitions, which are then used to solve the time-dependent amplitude and precursor equations. The amplitude function is calculated by solving the kinetics equations using the LSODE solver. When a new shape calculation is needed, the flux is calculated using the newly computed amplitude function. The Boltzmann transport equation is solved using the method of discrete ordinates to treat the directional variable and weighted finite-difference methods, in addition to Linear Nodal
JENDL-4.0 benchmarking for effective delayed neutron fraction of fast neutron systems
International Nuclear Information System (INIS)
Chiba, Go; Tsuji, Masashi; Sugiyama, Ken-ichiro; Narabayashi, Tadashi
2011-01-01
The performance of the latest Japanese evaluated nuclear data library JENDL-4.0 for the prediction of effective delayed neutron fraction β eff is assessed using experimental data of a wide range of fast neutron systems. Covariance data of JENDL-4.0 are used to quantify nuclear-data-induced uncertainties. Calculations with other libraries. JENDL-3.3, ENDF/B-VII.0, and JEFF-3.1, are also carried out for a quantitative comparison. JENDL-4.0 results in good agreement between calculation and experimental values within total uncertainties, and consistency between the differential nuclear data and integral experimental data is confirmed. While the other libraries also show good performance for β eff prediction, there are small differences in the predicted values of β eff among different libraries and ENDF/B-VII.0 gives the most stable results. Furthermore, a simple and convenient procedure to calculate sensitivity profiles of β eff to nuclear data is proposed. (author)
Neutron sources and its dosimetric characteristics
International Nuclear Information System (INIS)
Vega C, H.R.; Manzanares A, E.; Hernandez D, V.M.; Mercado S, G.A.; Gallego D, E.; Lorente F, A.
2005-01-01
By means of Monte Carlo methods the spectra of the produced neutrons 252 Cf, 252 Cf/D 2 O, 241 Am Be, 239 Pu Be, 140 La Be, 239 Pu 18 O 2 and 226 Ra Be have been calculated. With the information of the spectrum it was calculated the average energy of the neutrons of each source. By means of the fluence coefficients to dose it was determined, for each one of the studied sources, the fluence factors to dose. The calculated doses were H, H * (10), H p,sIab (10, 0 0 ), E AP and E ISO . During the phase of the calculations the sources were modeled as punctual and their characteristics were determined to 100 cm in the hole. Also, for the case of the sources of 239 Pu Be and 241 Am Be, were carried out calculations modeling the sources with their respective characteristics and the dosimetric properties were determined in a space full with air. The results of this last phase of the calculations were compared with the experimental results obtained for both sources. (Author)
Method and device for optimizing the measurements of the damping characteristics of therman neutrons
International Nuclear Information System (INIS)
Jacobson, L.A.; Johnstone, C.W.
1978-01-01
The borehole probe consists of a pulsed neutron generator and two detectors installed at different distances from the generator. The decay or damping characteristics of the thermal neutrons in a ground formation are measured by picking up indications of the concentration of thermal neutrons in the formation during a set of two measuring intervals offer irradiation. These measuring intervals consist of a sequence of discrete time gates. The time gates are subdivided into groups of progressive periods of time. The time delay between the pulses and the beginning of the sequence is adjusted by means of a selected scale factor value. (DG) [de
Characteristics of the JRR-3M neutron guide tubes
International Nuclear Information System (INIS)
Suzuki, Masatoshi; Ichikawa, Hiroki; Kawabata, Yuji.
1993-01-01
Large scale neutron guide tubes have been installed in the upgraded JRR-3 (Japan Research Reactor No.3, JRR-3M). The total length of the guide tubes is 232m. The neutron fluxes and spectra were measured at the end of the neutron guide tubes. The neutron fluxes of thermal neutron guide tubes with characteristic wavelength of 2A are 1.2 x 10 8 n/cm 2 · s. The neutron fluxes of cold guide tubes are 1.4 x 10 8 n/cm 2 · s with characteristic wavelength of 4A and 2.0 x 10 8 n/cm 2 · s with 6A when the cold neutron source is operated. The neutron spectra measured by time-of-flight method agree well with their designed ones. (author)
International Nuclear Information System (INIS)
Aumeier, S.E.; Forsmann, J.H.
1998-01-01
The ability to nondestructively determine the presence and quantity of fissile/fertile nuclei in various matrices is important in several areas of nuclear applications, including international and domestic safeguards, radioactive waste characterization, and nuclear facility operations. An analysis was performed to determine the feasibility of identifying the masses of individual fissionable isotopes from a cumulative delayed-neutron signal resulting form the neutron irradiation of several uranium and plutonium isotopes. The feasibility of two separate data-processing techniques was studied: Kalman filtering and genetic algorithms. The basis of each technique is reviewed, and the structure of the algorithms as applied to the delayed-neutron analysis problem is presented. The results of parametric studies performed using several variants of the algorithms are presented. The effect of including additional constraining information such as additional measurements and known relative isotopic concentration is discussed. The parametric studies were conducted using simulated delayed-neutron data representative of the cumulative delayed-neutron response following irradiation of a sample containing 238 U, 235 U, 239 Pu, and 240 Pu. The results show that by processing delayed-neutron data representative of two significantly different fissile/fertile fission ratios, both Kalman filters and genetic algorithms are capable of yielding reasonably accurate estimates of the mass of individual isotopes contained in a given assay sample
Development of a methodology for analysis of delayed-neutron signals
International Nuclear Information System (INIS)
Gross, K.C.; Strain, R.V.; Fryer, R.M.
1980-02-01
Experimental and analytical techniques have been developed for analysis and characterization of delayed-neutron (DN) signals that can provide diagnostic information to augment data from cover-gas analyses in the detection and identification of breached elements in an LMFBR. Eleven flow-reduction tests have been run in EBR-II to provide base data support for predicting DN signal characteristics during exposed-fuel operation. Results from the tests demonstrate the feasibility and practicability of response-analysis techniques for determining (a) the transit time, T/sub tr/, for DN emitters traveling from the core to the detector and (b) the isotropic holdup time, T/sub h/, of DN precursors in the fuel element
International Nuclear Information System (INIS)
Isaev, S.G.; Piksaikin, L.E.; Kazakov, L.E.; Tarasko, M.Z.
2000-01-01
The measurements of relative abundances and periods of delayed neutrons from fast neutron induced fission of 235 U and 236 U have been made at the electrostatic accelerator CG-2.5 at IPPE. The preliminary results were obtained and discussed in the frame of the systematics of the average half-life of delayed neutron precursors. It was shown that the average half-life value in both reactions depends on the energy of primary neutrons [ru
Characteristics of hypertrophic cardiomyopathy on delayed contrast-enhanced MRI
International Nuclear Information System (INIS)
Yan Chaowu; Zhao Shihua; Li Hua; Jiang Shiliang; Lu Minjie; Zhang Yan; Wei Yunqing; Ling Jian; Fang Wei
2010-01-01
Objective: To analyze the characteristics of hypertrophic cardiomyopathy (HCM) on delayed contrast-enhanced cardiac magnetic resonance imaging (CMRI). Methods: All patients underwent delayed contrast-enhanced CMRI. The left ventricle was divided into 9 segments to assess the location, extent and function of the hypertrophic segments. The t test was applied for the statistics. Results: Of 154 patients, delayed enhancement of' hypertrophic segment was found in 95 cases and non-delayed enhancement in 59 cases. The thickness and number of hypertrophic segment in patients with delayed enhancement were larger than those with non-delayed enhancement [(24.8±5.5) mm vs (20.4± 3.8) mm, t=3.82, P<0.05; (3.3±1.9) vs (2.4±1.7), t=2.26, P<0.05], and the age was younger [(46.0±15.2) years vs (55.0±11.9) years, t=-3.67, P<0.05]. The diffuse enhancement was found in 62 patients, and confluent enhancement in 33 patients. Confluent enhancement was found in all 14 patients after the alcohol ablation procedure. Conclusion: The age, thickness and number of hypertrophic segments in patients with delayed enhancement are different from those with non-delayed enhancement. (authors)
Neutronics characteristics of space power reactors
International Nuclear Information System (INIS)
Little, W.; Barner, J.
1986-01-01
The objective of the paper is to describe the neutronic characteristics of a range of possible space reactor designs, and indicate the relative advantages and disadvantages of the various designs. Fuel designs to be considered are cermets (i.e., ceramic particles embedded in a metal matrix) consisting of UO 2 or Nn ceramic particles in matrices of Nb, Mo, Ta, or W. These cermet fuels are compared to a UN pin-type design. UN was selected for the reference fuel material since it has a somewhat higher density than UO 2 (i.e., 14.32 versus 10.96 gm/cc), which allows a lower minimum critical mass for both ceramic and cermet designs
Directory of Open Access Journals (Sweden)
Tolosa-Delgado A.
2017-01-01
Full Text Available The commissioning of a new setup for β-delayed neutron measurements was carried out successfully in November-2016, at the RIKEN Nishina Center in Japan. The β-decay half-lives and Pn branching ratios of several isotopes in the 78Ni region were measured. Details of the experimental setup and the first results are given.
Delayed neutron kinetic functions for /sup 232/Th and /sup 238/U mixtures
Energy Technology Data Exchange (ETDEWEB)
Ganich, P P; Goshovskij, M V; Lendel, A I; Lomonosov, V I; Sikora, D I; Sychev, S I
1984-11-01
In order to investigate the applicability of the method based on using kinetic functions, describing the emission of delayed neutrons by samples for determination of the content of fissionable nuclides in binary mixtures, the /sup 232/Th+/sup 238/U mixtures have been analyzed with the M-30 microtron. Fresh samples containing ThO/sub 2/, U/sub 3/O/sub 8/ and their mixtures are irradiated by bremstrahlung at the 15.5 MeV energy of accelerated electrons and 9 ..mu..A average current. The mass of samples is about 6 g. To determine the kinetic functions, temporal distributions of delayed neutron pulses are used, their maximum number for different samples being (1.7-3.0) x 10/sup 4/. In processing the data obtained two methods of normalization of the delayed neutron number in the kinetic functions are used: to the total yield of delayed neutrons and to the yield of /sup 133/I ..gamma..-quanta. The conclusion is drawn that the method investigated permits to determine relative /sup 238/U concentrations in the mixtures considered with 0.06-0.2 errors. Error reduction is achieved during the normalization of the number of delayed neutrons to the yield of /sup 130/I ..gamma..-quanta.
A delayed neutron technique for measuring induced fission rates in fresh and burnt LWR fuel
Energy Technology Data Exchange (ETDEWEB)
Jordan, K.A., E-mail: kajordan@gmail.co [Paul Scherrer Institut, Laboratory for Reactor Physics and System Behaviour, 5232 Villigen (Switzerland); Perret, G. [Paul Scherrer Institut, Laboratory for Reactor Physics and System Behaviour, 5232 Villigen (Switzerland)
2011-04-01
The LIFE-PROTEUS program at the Paul Scherrer Institut is being undertaken to characterize the interfaces between burnt and fresh fuel assemblies in modern LWRs. Techniques are being developed to measure fission rates in burnt fuel following re-irradiation in the zero-power PROTEUS research reactor. One such technique utilizes the measurement of delayed neutrons. To demonstrate the feasibility of the delayed neutron technique, fresh and burnt UO{sub 2} fuel samples were irradiated in different positions in the PROTEUS reactor, and their neutron outputs were recorded shortly after irradiation. Fission rate ratios of the same sample irradiated in two different positions (inter-positional) and of two different samples irradiated in the same position (inter-sample) were derived from the measurements and compared with Monte Carlo predictions. Derivation of fission rate ratios from the delayed neutron measured signal requires correcting the signal for the delayed neutron source properties, the efficiency of the measurement setup, and the time dependency of the signal. In particular, delayed neutron source properties strongly depend on the fissile and fertile isotopes present in the irradiated sample and must be accounted for when deriving inter-sample fission rate ratios. Measured inter-positional fission rate ratios generally agree within 1{sigma} uncertainty (on the order of 1.0%) with the calculation predictions. For a particular irradiation position, however, a bias of about 2% is observed and is currently under investigation. Calculated and measured inter-sample fission rate ratios have C/E values deviating from unity by less than 1% and within 2{sigma} of the statistical uncertainties. Uncertainty arising from delayed neutron data is also assessed, and is found to give an additional 3% uncertainty factor. The measurement data indicate that uncertainty is overestimated.
Beta-Delayed Neutron Spectroscopy of 72Co with VANDLE
Keeler, Andrew; Grzywacz, Robert; King, Thomas; Taylor, Steven; Paulauskas, Stanley; Zachary, Christopher; Vandle Collaboration
2017-09-01
Measurements of simple, closed-shell isotopes far from stability provide important benchmarks for nuclear models and are a key constraint in r-process calculations. In particular, r-process models are sensitive to beta decay lifetimes and branching ratios of these neutron-rich isotopes. In this experiment, the Versatile Array of Neutron Detectors at Low Energy (VANDLE) was used to observe decays of nuclei produced by the fragmentation of 82Se at the National Superconducting Cyclotron Laboratory (NSCL). The neutron and gamma emissions of 72Co were measured to map the beta strength distribution (S_beta) above the neutron separation energy and infer the size of the Z = 28 shell gap in the 78Ni region. An implantation detector made of a radiation-hardened, inorganic scintillator was used to correlate implanted ions with beta decays as well as provide a start signal for the neutron Time of Flight measurement. Funded by the National Nuclear Security Administration under the Stewardship Science Academic Alliances program through DOE Award No. DE-NA0002132 and by the Office of Nuclear Physics, U.S. Department of Energy under Awards No. DE-FG02-96ER40983 (UTK).
We propose to use the new ISOLDE decay station and the neutron detector VANDLE to measure the $\\beta$-delayed neutron emission of N=82-84 $^{130-132}$Cd isotopes. The large delayed neutron emission probability observed in a previous ISOLDE measurement is indicative of the Gamow-Teller transitions due to the decay of deep core neutrons. Core Gamow-Teller decay has been experimentally proven in the $^{78}$Ni region for the N>50 nuclei using the VANDLE array. The spectroscopic measurement of delayed neutron emission along the cadmium isotopic chain will allow us to track the evolution of the single particle states and the shell gap.
Energy Technology Data Exchange (ETDEWEB)
Perret, G.; Jordan, K. A. [Paul Scherrer Institut, Villigen, 5232 (Switzerland)
2011-07-01
Novel techniques to measure newly induced fissions in spent fuel after re-irradiation at low power have been developed and tested at the Proteus zero-power research reactor. The two techniques are based on the detection of high energy gamma-rays emitted by short-lived fission products and delayed neutrons. The two techniques relate the measured signals to the total fission rate, the isotopic composition of the fuel, and nuclear data. They can be combined to derive better estimates on each of these parameters. This has potential for improvement in many areas. Spent fuel characterisation and safeguard applications can benefit from these techniques for non-destructive assay of plutonium content. Another application of choice is the reduction of uncertainties on nuclear data. As a first application of the combination of the delayed neutron and gamma measurement techniques, this paper shows how to reduce the uncertainties on the relative abundances of the longest delayed neutron group for thermal fissions in {sup 235}U, {sup 239}Pu and fast fissions in {sup 238}U. The proposed experiments are easily achievable in zero-power research reactors using fresh UO{sub 2} and MOX fuel and do not require fast extraction systems. The relative uncertainties (1{sigma}) on the relative abundances are expected to be reduced from 13% to 4%, 16% to 5%, and 38% to 12% for {sup 235}U, {sup 238}U and {sup 239}Pu, respectively. (authors)
Uranium borehole logging using delayed or prompt fission neutrons
International Nuclear Information System (INIS)
Schulze, G.; Wuerz, H.
1977-04-01
The measurement of induced fission neutrons using Cf 252 and 14 MeV neutrons is a sensitive method for an in situ determination of Uranium. Applying this methods requires a unique relation between concentration of Uranium and intensity of induced fission neutrons. A discussion of parameters influencing the determination of concentration is given. A simple method is developed allowing an elemination of the geochemistry of the deposit and of the borehole configuration. Borehole probes using the methods described are of considerable help during the phase of detailed exploration of uranium ore deposits. These on-line tools allow an immediate determination of concentration. Thus avoiding the expensive and time consuming step of core drilling and subsequent chemical analysis. (orig./HP) [de
Neutron spectra characteristics for the intense neutron source, INS
International Nuclear Information System (INIS)
Battat, M.; Dierckx, R.; Emigh, C.R.
1977-01-01
The Intense Neutron Source, INS, facility is presently under construction at the Los Alamos Scientific Laboratory. Its purpose is to provide a broad base for research work related to the radiation effects produced by 14-MeV neutrons from a D-T burn of a fusion reactor. The INS facility produces a D-T burn-like reaction from the collision of an intense tritium-ion beam with a supersonic jet target of deuterium gas. The reaction produces a typical D-T 14-MeV neutron spectrum. By adding a fission blanket surrounding the D-T ''burn,'' the neutron spectral shape may be tailored to match almost perfectly the anticipated first-wall spectra from presently proposed fusion reactors. With a blanket in place, the total production of neutrons can be as large as 3 x 10 16 n/s and experimental volumes of the order of 1000 cm 3 can be available at flux levels greater than 0.6 x 10 14 n/cm 2 s
Total body-calcium measurements: comparison of two delayed-gamma neutron activation facilities
International Nuclear Information System (INIS)
Ma, R.; Ellis, K.J.; Shypailo, R.J.; Pierson, R.N. Jr.
1999-01-01
This study compares two independently calibrated delayed-gamma neutron activation (DGNA) facilities, one at the Brookhaven National Laboratory (BNL), Upton, New York, and the other at the Children's Nutrition Research Center (CNRC), Houston, Texas that measure total body calcium (TBCa). A set of BNL phantoms was sent to CNRC for neutron activation analysis, and a set of CNRC phantoms was measured at BNL. Both facilities showed high precision (<2%), and the results were in good agreement, within 5%. (author)
Characteristics of a Portable Neutron Generator
Energy Technology Data Exchange (ETDEWEB)
Jin, Jeong-Tae; Oh, Byung-Hoon; Chang, Dae-Sik; In, Sang-Yeol; Huh, Sung-Ryul; Hong, Kwang-Pyo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2015-10-15
Neutron generators can be excellent tools for materials analysis, explosive material detection, nuclear weapon detection, and high quality radiography. D + D : 3He + n (2.5 MeV) D + T : 4He + n (14 MeV) Recent commercial neutron generators, fast neutron yield from 10{sup 7} to 10{sup 11} n/s, are produced by several companies and research groups around the world. But limited life time, high price, and frequent troubles make it difficult to develop related application systems by domestic companies or research groups. To remove such problems, it is necessary to develop our own domestic neutron generators. In this presentation, the design and experimental results on the developed small neutron generator are summarized. Experiments on deuterium beam extraction and fast neutron measurement by injecting deuterium beams on a drive-in target are executed. The stable deuterium beam of the energy higher than 100 keV was achieved by introducing metal cover which reduces the effect of metal-vacuum-insulator triple junction. The neutron flux of 5 n/s is measured by RadEye GN gamma Neutron (Thermo scientific) detector with about 200 mm distance and insertion of 40 mm PE plate between neutron source and the detector. The precise detector calibration is not carried out yet, so more detailed experimental results will be summarized at the presentation.
Characteristics of a Portable Neutron Generator
International Nuclear Information System (INIS)
Jin, Jeong-Tae; Oh, Byung-Hoon; Chang, Dae-Sik; In, Sang-Yeol; Huh, Sung-Ryul; Hong, Kwang-Pyo
2015-01-01
Neutron generators can be excellent tools for materials analysis, explosive material detection, nuclear weapon detection, and high quality radiography. D + D : 3He + n (2.5 MeV) D + T : 4He + n (14 MeV) Recent commercial neutron generators, fast neutron yield from 10 7 to 10 11 n/s, are produced by several companies and research groups around the world. But limited life time, high price, and frequent troubles make it difficult to develop related application systems by domestic companies or research groups. To remove such problems, it is necessary to develop our own domestic neutron generators. In this presentation, the design and experimental results on the developed small neutron generator are summarized. Experiments on deuterium beam extraction and fast neutron measurement by injecting deuterium beams on a drive-in target are executed. The stable deuterium beam of the energy higher than 100 keV was achieved by introducing metal cover which reduces the effect of metal-vacuum-insulator triple junction. The neutron flux of 5 n/s is measured by RadEye GN gamma Neutron (Thermo scientific) detector with about 200 mm distance and insertion of 40 mm PE plate between neutron source and the detector. The precise detector calibration is not carried out yet, so more detailed experimental results will be summarized at the presentation
Use of one delayed-neutron precursor group in transient analysis
International Nuclear Information System (INIS)
Diamond, D.J.
1983-01-01
In most reactor dynamics calculations six groups of delayed-neutron precursors are usually accounted for. However, under certain circumstances it may be advantageous to simplify the calculation and utilize a single delayed-neutron group. The motivation for going to one precursor group is economy. For LWR transient codes that use point kinetics the equations are solved very rapidly and six precursor groups should always be used. However, codes with spatially dependent neutron kinetics are very long running and the use of one precursor group may save computer costs and not impair the accuracy of the results significantly. Furthermore, in some codes, the elimation of five presursor groups makes additional memory available which may be used to give a net increase in the accuracy of the calculations, e.g., by allowing for an increase in mesh density. In order to use one delayed neutron precursor group it is necessary to derive a single decay constant, 6 lambda-, which, along with the total (or one group) delayed neutron fraction β = Σ/sub i = 1/β/sub i/, will adequately describe the transeint precursor behavior. The present summary explains how a recommendation for lambda- was derived
Estimation of delayed neutron emission probability by using the gross theory of nuclear β-decay
International Nuclear Information System (INIS)
Tachibana, Takahiro
1999-01-01
The delayed neutron emission probabilities (P n -values) of fission products are necessary in the study of reactor physics; e.g. in the calculation of total delayed neutron yields and in the summation calculation of decay heat. In this report, the P n -values estimated by the gross theory for some fission products are compared with experiment, and it is found that, on the average, the semi-gross theory somewhat underestimates the experimental P n -values. A modification of the β-decay strength function is briefly discussed to get more reasonable P n -values. (author)
Power measurement in the boiling capsules in R2 using delayed neutron detector
International Nuclear Information System (INIS)
Roennberg, G.
1979-03-01
LWR fuel testing is performed in the R2 reactor by irradiation in both loops and so-called boiling capsules. The loops have forced cooling, and the power can be measured calorimetrically by conventional instrumentation. The boiling capsules have convection cooling, and it has therefore been necessary to develop a special technique for power measurement, the delayed neutron detector (DND). The DND is a pneumatic rabbit system, which activates small uranium samples in the boiling capsules and counts the delayed neutrons for determination of the fission rate. This report describes the equipment used, the procedure of measurement, and the method of evaluation. (atuhor)
International Nuclear Information System (INIS)
Coelho, Paulo Rogerio Pinto
1979-01-01
This work presents results of non destructive mass analysis of natural uranium by the pulsed source technique. Fissioning is produced by irradiating the test sample with pulses of 14 MeV neutrons and the uranium mass is calculated on a relative scale from the measured emission of delayed neutrons. Individual measurements were normalised against the integral counts of a scintillation detector measuring the 14 MeV neutron intensity. Delayed neutrons were measured using a specially constructed slab detector operated in anti synchronism with the fast pulsed source. The 14 MeV neutrons were produced via the T(d,n) 4 He reaction using a 400 kV Van de Graaff accelerated operated at 200 kV in the pulsed source mode. Three types of sample were analysed, namely: discs of metallic uranium, pellets of sintered uranium oxide and plates of uranium aluminium alloy sandwiched between aluminium. These plates simulated those of Material Testing Reactor fuel elements. Results of measurements were reproducible to within an overall error in the range 1.6 to 3.9%; the specific error depending on the shape, size and mass of the sample. (author)
A Neutron Sensitive Microchannel Plate Detector with Cross Delay Line Readout
International Nuclear Information System (INIS)
Berry, Kevin D.; Bilheux, Hassina Z.; Crow, Lowell; Diawara, Yacouba; Feller, W. Bruce; Iverson, Erik B.; Martin, Adrian; Robertson, J. Lee
2012-01-01
Microchannel plates containing neutron absorbing elements such as boron and gadolinium in the bulk glass are used as the sensing element in high spatial resolution, high rate neutron imaging systems. In this paper we describe one such device, using both 10 B and natural Gd, which employs cross delay line signal readout, with time-of-flight capability. This detector has a measured spatial resolution under 40 m FWHM, thermal neutron efficiency of 19%, and has recorded rates in excess of 500 kHz. A physical and functional description is presented, followed by a discussion of measurements of detector performance and a brief survey of some practical applications.
Non-destructive isotopic uranium assay by multiple delayed neutron measurements
International Nuclear Information System (INIS)
Papadopoulos, N.N.; Tsagas, N.F.
1991-01-01
The high accuracy and precision required in nuclear safeguards measurements can be achieved by an improved neutron activation technique based on multiple delayed fission neutron counting under various experimental conditions. For the necessary ultrahigh counting statistics required, cyclic activation of multiple subsamples has been applied. The home-made automated flexible analytical system with neutron flux and spectrum differentiation by irradiation position adjustment and cadmium screening, permits the non-destructive determination of the U235 abundance and the total U element concentration needed in nuclear safeguards sample analysis, with a high throughout and a low operational cost. Careful experimental optimization led to considerable improvement of the results
First measurement of several $\\beta$-delayed neutron emitting isotopes beyond N=126
Caballero-Folch, R.; Agramunt, J.; Algora, A.; Ameil, F.; Arcones, A.; Ayyad, Y.; Benlliure, J.; Borzov, I.N.; Bowry, M.; Calvino, F.; Cano-Ott, D.; Cortés, G.; Davinson, T.; Dillmann, I.; Estrade, A.; Evdokimov, A.; Faestermann, T.; Farinon, F.; Galaviz, D.; García, A.R.; Geissel, H.; Gelletly, W.; Gernhäuser, R.; Gómez-Hornillos, M.B.; Guerrero, C.; Heil, M.; Hinke, C.; Knöbel, R.; Kojouharov, I.; Kurcewicz, J.; Kurz, N.; Litvinov, Y.; Maier, L.; Marganiec, J.; Marketin, T.; Marta, M.; Martínez, T.; Martínez-Pinedo, G.; Montes, F.; Mukha, I.; Napoli, D.R.; Nociforo, C.; Paradela, C.; Pietri, S.; Podolyák, Zs.; Prochazka, A.; Rice, S.; Riego, A.; Rubio, B.; Schaffner, H.; Scheidenberger, Ch.; Smith, K.; Sokol, E.; Steiger, K.; Sun, B.; Taín, J.L.; Takechi, M.; Testov, D.; Weick, H.; Wilson, E.; Winfield, J.S.; Wood, R.; Woods, P.; Yeremin, A.
2016-01-01
The $\\beta$-delayed neutron emission probabilities of neutron rich Hg and Tl nuclei have been measured together with $\\beta$-decay half-lives for 20 isotopes of Au, Hg, Tl, Pb and Bi in the mass region N$\\gtrsim$126. These are the heaviest species where neutron emission has been observed so far. These measurements provide key information to evaluate the performance of nuclear microscopic and phenomenological models in reproducing the high-energy part of the $\\beta$-decay strength distribution. In doing so, it provides important constraints to global theoretical models currently used in $r$-process nucleosynthesis.
First Measurement of Several β-Delayed Neutron Emitting Isotopes Beyond N=126.
Caballero-Folch, R; Domingo-Pardo, C; Agramunt, J; Algora, A; Ameil, F; Arcones, A; Ayyad, Y; Benlliure, J; Borzov, I N; Bowry, M; Calviño, F; Cano-Ott, D; Cortés, G; Davinson, T; Dillmann, I; Estrade, A; Evdokimov, A; Faestermann, T; Farinon, F; Galaviz, D; García, A R; Geissel, H; Gelletly, W; Gernhäuser, R; Gómez-Hornillos, M B; Guerrero, C; Heil, M; Hinke, C; Knöbel, R; Kojouharov, I; Kurcewicz, J; Kurz, N; Litvinov, Yu A; Maier, L; Marganiec, J; Marketin, T; Marta, M; Martínez, T; Martínez-Pinedo, G; Montes, F; Mukha, I; Napoli, D R; Nociforo, C; Paradela, C; Pietri, S; Podolyák, Zs; Prochazka, A; Rice, S; Riego, A; Rubio, B; Schaffner, H; Scheidenberger, Ch; Smith, K; Sokol, E; Steiger, K; Sun, B; Taín, J L; Takechi, M; Testov, D; Weick, H; Wilson, E; Winfield, J S; Wood, R; Woods, P; Yeremin, A
2016-07-01
The β-delayed neutron emission probabilities of neutron rich Hg and Tl nuclei have been measured together with β-decay half-lives for 20 isotopes of Au, Hg, Tl, Pb, and Bi in the mass region N≳126. These are the heaviest species where neutron emission has been observed so far. These measurements provide key information to evaluate the performance of nuclear microscopic and phenomenological models in reproducing the high-energy part of the β-decay strength distribution. This provides important constraints on global theoretical models currently used in r-process nucleosynthesis.
Energy Technology Data Exchange (ETDEWEB)
Huy, Ngo Quang; Thong, Ha Van [Nuclear Research Inst., Da Lat (Viet Nam)
1994-10-01
The investigation of the neutron physics characteristics of the Dalat Reactor has obtained the results as follows: 1/ The effective fraction of delayed photoneutrons and the extraneous neutron source left after reactor shut down are measured. 2/ The lowest power levels of critical states of the reactor are determined. 3/The perturbation effect is investigated when a water column or a plexiglass rod is substituted for a fuel element. 4/ The relative axial and radial distributions of the thermal neutrons measured and the geometrical parameters of the core such as the inhomogeneous coefficients, the buckling, the effective height and radius, the extrapolated distances are obtained. 4/ The thermal neutron distributions are measured around the old graphite reflector. (author). 10 refs., 10 figs., 2 tabs.
Delayed Neutron Fraction (beta-effective) Calculation for VVER 440 Reactor
International Nuclear Information System (INIS)
Hascik, J.; Michalek, S.; Farkas, G.; Slugen, V.
2008-01-01
Effective delayed neutron fraction (β eff ) is the main parameter in reactor dynamics. In the paper, its possible determination methods are summarized and a β eff calculation for a VVER 440 power reactor as well as for training reactor VR1 using stochastic transport Monte Carlo method based code MCNP5 is made. The uncertainties in determination of basic delayed neutron parameters lead to the unwished conservatism in the reactor control system design and operation. Therefore, the exact determination of the β eff value is the main requirement in the field of reactor dynamics. The interest in the delayed neutron data accuracy improvement started to increase at the end of 80-ties and the beginning of 90-ties, after discrepancies among the results of experiments and measurements what do you mean differences between different calculation approaches and experimental results. In consequence of difficulties in β eff experimental measurement, this value in exact state is determined by calculations. Subsequently, its reliability depends on the calculation method and the delayed neutron data used. An accurate estimate of β eff is essential for converting reactivity, as measured in dollars, to an absolute reactivity and/or to an absolute k eff . In the past, k eff has been traditionally calculated by taking the ratio of the adjoint- and spectrum-weighted delayed neutron production rate to the adjoint- and spectrum-weighted total neutron production rate. An alternative method has also been used in which β eff is calculated from simple k-eigenvalue solutions. The summary of the possible β eff determination methods can be found in this work and also a calculation of β eff first for the training reactor VR1 in one operation state and then for VVER 440 power reactor in two different operation states are made using the prompt method, by MCNP5 code.(author)
International Nuclear Information System (INIS)
Kiyanagi, Yoshiaki
2012-01-01
Neutron beam is one of important tools to obtain the transmission image of an object. Until now, steady state neutron sources such as reactors are mainly used for this imaging purpose. Recently, it has been demonstrated that pulsed neutron imaging based on accelerator neutron sources can provide a real-space distribution of physical information of materials such as crystallographic structure, element, temperature, hydrogen bound state, magnetic field and so on, by analyzing wavelength dependent transmission spectrum, which information cannot be observed or difficult to obtain with a traditional imaging method using steady state neutrons. Here, characteristics of the pulsed neutron source and principle of the pulsed neutron imaging are explained as a basic concept of the new method. (author)
International Nuclear Information System (INIS)
Nagaya, Yasunobu
2013-01-01
Benchmark calculations with a continuous-energy Monte Carlo code have been performed for delayed neutron data of JENDL-4.0. JENDL-4.0 gives good prediction for the effective delayed neutron fraction in the present benchmarks but further detailed analysis is required for some cores. (author)
Reference neutron radiations. Part 1: Characteristics and methods of production
International Nuclear Information System (INIS)
2001-01-01
ISO 8529 consists of the following parts, under the general title Reference neutron radiations: Part 1: Characteristics and methods of production; Part 2: Calibration fundamentals of radiation protection devices related to the basic quantities characterizing the radiation field; Part 3: Calibration of area and personal dosimeters and determination of response as a function of energy and angle of incidence. This Part 1. of ISO 8529 specifies the reference neutron radiations, in the energy range from thermal up to 20 MeV, for calibrating neutron-measuring devices used for radiation protection purposes and for determining their response as a function of neutron energy. Reference radiations are given for neutron fluence rates of up to 1x10 9 m 2 s-1 , corresponding, at a neutron energy of 1 MeV, to dose-equivalent rates of up to 100 mSv h -1 . This part of ISO 8529 is concerned only with the methods of producing and characterizing the neutron reference radiations. The procedures for applying these radiations for calibrations are described in ISO 8529-2 and ISO 8529-3. The reference radiations specified are the following: neutrons from radionuclide sources, including neutrons from sources in a moderator; neutrons produced by nuclear reactions with charged particles from accelerators; neutrons from reactors. In view of the methods of production and use of them, these reference radiations are divided, for the purposes of this part of ISO 8529, into the following two separate sections. In clause 4, radionuclide neutron sources with wide spectra are specified for the calibration of neutron measuring devices. These sources should be used by laboratories engaged in the routine calibration of neutron-measuring devices, the particular design of which has already been type tested. In clause 5, accelerator-produced monoenergetic neutrons and reactor-produced neutrons with wide or quasi monoenergetic spectra are specified for determining the response of neutron-measuring devices
First delayed neutron emission measurements at ALTO with the neutron detector TETRA
International Nuclear Information System (INIS)
Testov, D.; Ancelin, S.; Bettane, J.; Ibrahim, F.; Kolos, K.; Mavilla, G.; Niikura, M.; Verney, D.; Wilson, J.; Kuznetsova, E.; Penionzhkevich, Yu.; Smirnov, V.; Sokol, E.
2013-01-01
Beta-decay properties are among the easiest and, therefore, the first ones to be measured to study new neutron-rich isotopes. Eventually, a very small number of nuclei could be sufficient to estimate their lifetime and neutron emission probability. With the new radioactive beam facilities which have been commissioned recently (or will be constructed shortly) new areas of neutron-rich isotopes will become reachable. To study beta-decay properties of such nuclei at IPN (Orsay) in the framework of collaboration with JINR (Dubna), a new experimental setup including the neutron detector of high efficiency TETRA was developed and commissioned
Calculation of the pulsed Feynman- and Rossi-alpha formulae with delayed neutrons
International Nuclear Information System (INIS)
Kitamura, Y.; Pazsit, I.; Wright, J.; Yamamoto, A.; Yamane, Y.
2005-01-01
In previous works, the authors have developed an effective solution technique for calculating the pulsed Feynman- and Rossi-alpha formulae. Through derivation of these formulae, it was shown that the technique can easily handle various pulse shapes of the pulsed neutron source. Furthermore, it was also shown that both the deterministic (i.e., synchronizing with the pulsing of neutron source) and stochastic (non-synchronizing) Feynman-alpha formulae can be obtained with this solution technique. However, for mathematical simplicity and the sake of insight, the formal derivation was performed in a model without delayed neutrons. In this paper, to demonstrate the robustness of the technique, the pulsed Feynman- and Rossi-alpha formulae were re-derived by taking one group of delayed neutrons into account. The results show that the advantages of this technique are retained even by inclusion of the delayed neutrons. Compact explicit formulae are derived for the Feynman- and Rossi-alpha methods for various pulse shapes and pulsing methods
Energy Technology Data Exchange (ETDEWEB)
Oyamatsu, Kazuhiro [Nagoya Univ. (Japan)
1998-03-01
Application programs for personal computers are developed to calculate the decay heat power and delayed neutron activity from fission products. The main programs can be used in any computers from personal computers to main frames because their sources are written in Fortran. These programs have user friendly interfaces to be used easily not only for research activities but also for educational purposes. (author)
The universal library of fission products and delayed neutron group yields
International Nuclear Information System (INIS)
Koldobskiy, A.B.; Zhivun, V.M.
1997-01-01
A new fission product yield library based on the Semiempirical method for the estimation of their mass and charge distribution is described. Contrary to other compilations, this library can be used with all possible excitation energies of fissionable actinides. The library of delayed neutron group yields, based on the fission product yield compilation, is described as well. (author). 15 refs, 4 tabs
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
Measurement of the most exotic beta-delayed neutron emitters at N=50 and N=126
Dillmann, Iris
2017-09-01
Beta-delayed neutron (βn)-emission will be the dominant decay mechanism of neutron-rich nuclei and plays an important role in the stellar nucleosynthesis of heavy elements in the ``r process''. It leads to a detour of the material β-decaying back to stability and the released neutrons increase the neutron-to-seed ratio, and are re-captured during the freeze-out phase and thus influence the final solar r-abundance curve. Thus the neutron branching ratio of very neutron-rich isotopes is a crucial parameter in astrophysical simulations. In addition, β-decay half-lives can be deduced from the time-dependent detection of βn's. I will talk about two recent experimental campaigns. The neutron detector BELEN was used at GSI Darmstadt to measure half-lives and neutron-branching ratios of the heaviest presently accessible βn-emitters at N=126. For isotopes between 204Au and 220Bi nine half-lives and eight neutron-branching ratios were measured for the first time and provide an important input for benchmarking theoretical models in this mass region. Its successor is the BRIKEN detector (``Beta-delayed neutron measurements at RIKEN for nuclear structure, astrophysics, and applications''), the most efficient neutron detector used so far for nuclear structure studies. In conjunction with two clover detectors and the ``Advanced Implantation Detector Array'' (AIDA) the setup has been used a few months ago to measure the most neutron-rich isotopes around 78Ni, 132Sn, and the Rare Earth Region. Some preliminary results are shown from the campaign covering the 78Ni region where the neutron-branching ratio of 78Ni and 28 more isotopes were measured for the first time, as well as the half-lives of 20 isotopes. The BRIKEN campaign aims to (re-)measure almost all βn-emitters between 76Co and 167Eu, many of them for the first time. An extension of the campaign to lighter masses is planned. This work has been supported by the NSERC and NRC in Canada, the US DOE, the Spanish
Measurement of the Effective Delayed Neutron Fraction in Three Different FR0-cores
Energy Technology Data Exchange (ETDEWEB)
Moberg, L; Kockum, J
1972-06-15
The effective delayed neutron fraction, beta{sub eff}, has been measured in the three cores 3, 5 and 8 of the fast zero-power reactor FR0. The variance-to-mean method, in which the statistical fluctuations of the neutron density in the reactor is studied, was used. A 3He-gas scintillator was placed in the reflector and used as a neutron detector. It was made more sensitive to fast neutrons by surrounding it with polythene. Its efficiency, expressed as the number of counts per fission in the reactor, was determined using fission chambers with known efficiency placed in the core. The space distribution of the fission rate in the core was determined by foil activation technique. The experimental results were compared with theoretical beta{sub eff}-values calculated with perturbation theory. The difference was about 3 % which is of the same order as the accuracy in the experimental values
Statistical effects in beta-delayed neutron emission from fission product nuclides
International Nuclear Information System (INIS)
McElroy, R.D. Jr.
1986-01-01
The delayed neutron spectra for the precursors Rb-93, 94, 95, 96, 97 and Cs-145 were measured by use of the on-line isotope separator facility TRISTAN and a time-of-flight (TOF) spectrometer. Flight paths were used that provided, for energies below 70 keV, a FWHM energy resolution between 2 and 4 percent. Each spectrum showed discrete neutron peaks below 156 keV, with as many as 26 in the Rb-95 spectra. Level densities near the neutron binding energy in the neutron-emitting nuclide were deduced using a missing-level indicator based on a Porter-Thomas distribution of neutron peak intensities. The resulting level density data were compared to the predictions of the Gilbert and Cameron formulism and to those of Dilg, Schantl, Vonach and Uhl. Comparisons were made between the empirically-based level parameter a and the values predicted by each model for Sr-93, 94, 95, 97 and Ba-145. The two models appear, within the uncertainties, to be equally capable of describing these neutron-rich nuclides and equally as capable for them as they are for nuclides in the valley of beta stability. Measurements of the neutron strength function are sometimes possible with the present TOF system for neutron decays with competing neutron branches to levels in the grandchild nucleus. A value for the d-wave strength function of Sr-96 is found to be (4.2 +- 1.1)/10 4 . Improvements in the TOF system, allowing the measurement of the neutron strength function for the more general case, are discussed. 72 refs., 56 figs., 16 tabs
International Nuclear Information System (INIS)
De Oliveira, Z.M.
1980-01-01
A detailed analysis of the simple statistical model description for delayed neutron emission of 87 Br, 137 I, 85 As and 135 Sb has been performed. In agreement with experimental findings, structure in the #betta#-strength function is required to reproduce the envelope of the neutron spectrum from 87 Br. For 85 As and 135 Sb the model is found incapable of simultaneously reproducing envelopes of delayed neutron spectra and neutron branching ratios to excited states in the final nuclei for any choice of #betta#-strength function. The results indicate that partial widths for neutron emission are not compatible with optical-model transmission coefficients. The simple shell model with pairing is shown to qualitatively describe the main features of the #betta#-strength functions for decay of 87 Br and 91 93 95 97 Rb. It is found that the location of apparent resonances in the experimental data are in rough agreement with the location of centroids of strength calculated with this model. An extension of the shell model picture which includes the Gamow-Teller residual interaction is used to investigate decay properties of 84 86 As, 86 92 Br and 88 102 Rb. For a realistic choice of interaction strength, the half lives of these isotopes are fairly well reproduced and semiquantitative agreement with experimental #betta#-strength functions is found. Delayed neutron emission probabilities are reproduced for precursors nearer stability with systematic deviations being observed for the heavier nuclei. Contrary to the assumption of a structureless Gamow-Teller giant resonance as embodied gross theory of #betta#-decay, we find that structures in the tail of the Gamow-Teller giant resonances are expected which strongly influence the decay properties of nuclides in this region
Savannah River Site delayed neutron instruments for safeguards measurements
International Nuclear Information System (INIS)
Studley, R.V.
1992-01-01
The Savannah River Site (SRS) includes a variety of nuclear production facilities that, since 1953, have processed special nuclear materials (SNM) including highly-enriched uranium (>90% 235 U), recycled enriched uranium (∼50% 235 U + 40% 236 U), low burnup plutonium (> 90% 239 Pu + 240 Pu ) and several other nuclear materials such as heat source plutonium ( 238 Pu). DOE Orders, primarily 5633.3, require all nuclear materials to be safeguarded through accountability and material control. Accountability measurements determine the total amount of material in a facility, balancing inventory changes against receipts and shipments, to provide assurance (delayed) that all material was present. Material control immediately detects or deters theft or diversion by assuring materials remain in assigned locations or by impeding unplanned movement of materials within or from a material access area. Goals for accountability or material control, and, therefore, the design of measurement systems, are distinctly different. Accountability measurements are optimized for maximum precision and accuracy, usually for large amounts of special nuclear material. Material control measurements are oriented more toward security features and often must be optimized for sensitivity, to detect small amounts of materials where none should be
Population of delayed-neutron granddaughter states and the optical potential
International Nuclear Information System (INIS)
Schenter, R.E.; Mann, F.M.; Warner, R.A.; Reeder, P.L.
1982-08-01
Using a statistical treatment of beta decay and the Hauser-Feshbach model of nuclear reactions, calculations were made and compared to recent experimental measurements of the population of granddaughter states of several delayed neutron precursors ( 144 145 147 Cs and 96 Rb). Emphasis of this paper is on the sensitivity and interpretation of experimental results to various standard low energy neutron optical model potentials and variations in their forms and parameters. Results for these precursors show qualitative agreement with experiment for all the optical potential models used and good quantitative agreement for two (Moldauer and Becchetti-Greenlees). Questions such as (N-Z) terms, deformation and nonlocality dependence are presented
Determination of the effective delayed neutron fraction in the Coral-I Reactor
International Nuclear Information System (INIS)
Francisco, J. L. de; Perez-Navarro, A.; Rodriguez-Mayquez, E.
1973-01-01
The effective delayed neutron fraction, β eff, has been determined from the measurement of E / β 2 , by means of reactor noise analysis in the time domain, and the neutron detector efficiency, ε. For the ε measurement it is necessary to determine the fission rate in the reactor. This value can be obtained from the absolute measurement of the fission rate per cm 3 , at a certain point of the reactor, and the determination of these two values ratio, which has been calculated by the Monte Cario method and also measured with results in good agreement. (Author)
International Nuclear Information System (INIS)
Talamo, Alberto
2013-01-01
This study presents three numerical algorithms to solve the time dependent neutron transport equation by the method of the characteristics. The algorithms have been developed taking into account delayed neutrons and they have been implemented into the novel MCART code, which solves the neutron transport equation for two-dimensional geometry and an arbitrary number of energy groups. The MCART code uses regular mesh for the representation of the spatial domain, it models up-scattering, and takes advantage of OPENMP and OPENGL algorithms for parallel computing and plotting, respectively. The code has been benchmarked with the multiplication factor results of a Boiling Water Reactor, with the analytical results for a prompt jump transient in an infinite medium, and with PARTISN and TDTORT results for cross section and source transients. The numerical simulations have shown that only two numerical algorithms are stable for small time steps
Energy Technology Data Exchange (ETDEWEB)
Talamo, Alberto, E-mail: alby@anl.gov [Nuclear Engineering Division, Argonne National Laboratory, 9700 South Cass Avenue, Lemont, IL 60439 (United States)
2013-05-01
This study presents three numerical algorithms to solve the time dependent neutron transport equation by the method of the characteristics. The algorithms have been developed taking into account delayed neutrons and they have been implemented into the novel MCART code, which solves the neutron transport equation for two-dimensional geometry and an arbitrary number of energy groups. The MCART code uses regular mesh for the representation of the spatial domain, it models up-scattering, and takes advantage of OPENMP and OPENGL algorithms for parallel computing and plotting, respectively. The code has been benchmarked with the multiplication factor results of a Boiling Water Reactor, with the analytical results for a prompt jump transient in an infinite medium, and with PARTISN and TDTORT results for cross section and source transients. The numerical simulations have shown that only two numerical algorithms are stable for small time steps.
International Nuclear Information System (INIS)
Ohm, H.
1982-01-01
Using the example of the delayed neutron spectrum of 24 s- 137 I the statistical model is tested in view of its applicability. A computer code was developed which simulates delayed neutron spectra by the Monte Carlo method under the assumption that the transition probabilities of the ν and the neutron decays obey the Porter-Thomas distribution while the distances of the neutron emitting levels are Wigner distribution. Gramow-Teller ν-transitions and simply forbidden ν-transitions from the preceding nucleus to the emitting nucleus were regarded. (orig./HSI) [de
International Nuclear Information System (INIS)
Kophazi, J.; Czifrus, Sz.; Feher, S.; Por, G.
2001-01-01
The paper describes the measurement of the delayed signal of a Rh emitter Self Powered Neutron Detector (SPND) separately from other signal components originating from (n-gamma-e), (background gamma-e) and other effects. In order to separate the delayed signal, the detector was removed from the reactor core and placed to an adequately distant location during the measurement, where the radiation from the core was negligible. The experiment was carried out on the 100kW light water tank-type reactor of Technical University of Budapest and the results of the measurement were compared with the results of Monte Carlo calculations.(author)
Study and building of a detection array for delayed neutrons: TONNERRE
International Nuclear Information System (INIS)
Martin, Thierry
1998-01-01
This work has been undertaken within a French-Romanian collaboration in order to build a high efficiency detector array for delayed neutrons: barrel-shaped TONNERRE. Some neutron-rich nuclei decay through 1, 2 or 3 neutron emission after β - decay. More exotic nuclei will be produced by SPIRAL at GANIL. An array with high efficiency and good resolution is then required. Thirty two BC400 plastic scintillators (160 x 20 x 4 cm 3 ) allow us to get the time of flight neutron spectra. They are bent for uniform flight path and viewed by a photomultiplier tube at both ends. Simulations have allowed to establish scintillator size and to minimize light attenuation. Intrinsic efficiency and crosstalk have been measured with 252 Cf and compared to GEANT. 1 to 5 MeV neutrons are detected with good timing and position properties. Other counters will be built for neutrons from 300 keV to 1 MeV. Planned to run at several particle accelerators (GANIL, CERN, and others), TONNERRE is modular and many geometries are possible. (author)
International Nuclear Information System (INIS)
Yedvab, Y.; Reiss, I.; Bettan, M.; Harari, R.; Grober, A.; Ettedgui, H.; Caspi, E. N.
2006-01-01
A method for determining delayed neutrons source in the frequency domain based on measuring power oscillations in a non-critical reactor is presented. This method is unique in the sense that the delayed neutrons source is derived from the dynamic behavior of the reactor, which serves as the measurement system. An algorithm for analyzing power oscillation measurements was formulated, which avoids the need for a multi-parameter non-linear fit process used by other methods. Using this algorithm results of two sets of measurements performed in IRR-I and IRR-II (Israeli Research Reactors I and II) are presented. The agreement between measured values from both reactors and calculated values based on Keepin (and JENDL-3.3) group parameters is very good. (authors)
^{235}U Determination using In-Beam Delayed Neutron Counting Technique at the NRU Reactor
Energy Technology Data Exchange (ETDEWEB)
Andrews, M. T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Bentoumi, G. [Canadian Nuclear Labs., Chalk River, ON (Canada); Corcoran, E. C. [Royal Military College of Canada, Kingston, ON (United States); Dimayuga, I. [Canadian Nuclear Labs., Chalk River, ON (Canada); Kelly, D. G. [Royal Military College of Canada, Kingston, ON (United States); Li, L. [Canadian Nuclear Labs., Chalk River, ON (Canada); Sur, B. [Canadian Nuclear Labs., Chalk River, ON (Canada); Rogge, R. B. [Canadian Nuclear Labs., Chalk River, ON (Canada)
2015-11-17
This paper describes a collaborative effort that saw the Royal Military College of Canada (RMC)’s delayed neutron and gamma counting apparatus transported to Canadian Nuclear Laboratories (CNL) for use in the neutron beamline at the National Research Universal (NRU) reactor. Samples containing mg quantities of fissile material were re-interrogated, and their delayed neutron emissions measured. This collaboration offers significant advantages to previous delayed neutron research at both CNL and RMC. This paper details the determination of ^{235}U content in enriched uranium via the assay of in-beam delayed neutron magnitudes and temporal behavior. ^{235}U mass was determined with an average absolute error of ± 2.7 %. This error is lower than that obtained at RMCC for the assay of ^{235}U content in aqueous solutions (3.6 %) using delayed neutron counting. Delayed neutron counting has been demonstrated to be a rapid, accurate, and precise method for special nuclear material detection and identification.
A two-dimensional detector with delay line readout for slow neutron fields measurements
International Nuclear Information System (INIS)
Cheremukhina, G.A.; Chernenko, S.P.; Ivanov, A.B.
1992-01-01
This article presents the description of a two-dimensional detector of slow neutrons together with its readout and data acquisition electronics based on a PC/AT> The detector with a sensitive area of 260x140 mm 2 is based on a high pressure multiwire proportional chamber with delay line readout and gas filling of 3.0 atm. 3 He + propane. 25 refs.; 10 figs.; 2 tabs
International Nuclear Information System (INIS)
Bécares, V.; Pérez-Martín, S.; Vázquez-Antolín, M.; Villamarín, D.; Martín-Fuertes, F.; González-Romero, E.M.; Merino, I.
2014-01-01
Highlights: • Review of several Monte Carlo effective delayed neutron fraction calculation methods. • These methods have been implemented with the Monte Carlo code MCNPX. • They have been benchmarked against against some critical and subcritical systems. • Several nuclear data libraries have been used. - Abstract: The calculation of the effective delayed neutron fraction, β eff , with Monte Carlo codes is a complex task due to the requirement of properly considering the adjoint weighting of delayed neutrons. Nevertheless, several techniques have been proposed to circumvent this difficulty and obtain accurate Monte Carlo results for β eff without the need of explicitly determining the adjoint flux. In this paper, we make a review of some of these techniques; namely we have analyzed two variants of what we call the k-eigenvalue technique and other techniques based on different interpretations of the physical meaning of the adjoint weighting. To test the validity of all these techniques we have implemented them with the MCNPX code and we have benchmarked them against a range of critical and subcritical systems for which either experimental or deterministic values of β eff are available. Furthermore, several nuclear data libraries have been used in order to assess the impact of the uncertainty in nuclear data in the calculated value of β eff
International Nuclear Information System (INIS)
Yoshida, Tadashi
1999-01-01
The Delayed Neutron Working Group was established in April 1997 within the Nuclear Data Subcommittee of JNDC. It has two principal missions. One is to coordinate the Japanese activities toward the WPEC/Subgroup-6 efforts, and the other is to recommend the delayed neutron data for JENDL-3.3. The final report of Subgroup-6, which in one of the subgroups of the NEA International Evaluation Cooperation (WPEC) and is in charge of the delayed neutron data, is to be completed in 1999. Here in Japan, JENDL-3.3 is planned to be released in early 2000. Delayed Neutron Working Group is, then, going to finalize its activity by the end of the fiscal year 1999 after recommending appropriate sets of data as coherently as possible with the of Subgroup-6 efforts. (author)
Nuclear and activation characteristics of materials in 14.1-MeV and 2.5-MeV neutron field
International Nuclear Information System (INIS)
Seki, Yasushi; Takeyasu, Yuuichi.
1988-11-01
The nuclear and activation characteristics of various materials and elements of interest in terms of fusion reactor design are calculated and the results are graphically shown. The elements and materials are placed in a simple geometry modelling a blanket and shield of a fusion reactor. The neutrons with 14.1-MeV and 2.5-MeV energy are generated from the region represented as D-T and D-D plasma, respectively. The following activation characteristics after neutron irradiation are shown for each material and element; 1. Time evolution of induced activity, 2. Time evolution of decay heat, 3. Delayed gamma-ray dose distribution, 4. Decay heat distribution. In addition to the above activation characteristics, nuclear characteristics during the neutron irradiation, e.g. neutron energy spectra, neutron and gamma-ray flux distribution, nuclear heating distributions, and neutron and gamma-ray dose rate are also shown. (author)
Energy Technology Data Exchange (ETDEWEB)
Dillmann, Iris [TRIUMF, Vancouver BC, V6T 2A3, Canada and GSI Helmholtzzentrum für Schwerionenforschung GmbH, D-64291 Darmstadt (Germany); Abriola, Daniel [Laboratorio Tandar, Comisión Nacional de Energía Atómica, B1650KINA, San Martín, Buenos Aires (Argentina); Singh, Balraj [Department of Physics and Astronomy, McMaster University, Hamilton ON, L8S 4M1 (Canada)
2014-05-02
Beta-delayed neutron (βn) emitters play an important, two-fold role in the stellar nucleosynthesis of heavy elements in the 'rapid neutron-capture process' (r process). On one hand they lead to a detour of the material β-decaying back to stability. On the other hand, the released neutrons increase the neutron-to-seed ratio, and are re-captured during the freeze-out phase and thus influence the final solar r-abundance curve. A large fraction of the isotopes inside the r-process reaction path are not yet experimentally accessible and are located in the (experimental) 'Terra Incognita'. With the next generation of fragmentation and ISOL facilities presently being built or already in operation, one of the main motivation of all projects is the investigation of these very neutron-rich isotopes. A short overview of one of the planned programs to measure βn-emitters at the limits of the presently know isotopes, the BRIKEN campaign (Beta delayed neutron emission measurements at RIKEN) will be given. Presently, about 600 β-delayed one-neutron emitters are accessible, but only for a third of them experimental data are available. Reaching more neutron-rich isotopes means also that multiple neutron-emission becomes the dominant decay mechanism. About 460 β-delayed two-, three-or four-neutron emitters are identified up to now but for only 30 of them experimental data about the neutron branching ratios are available, most of them in the light mass region below A=30. The International Atomic and Energy Agency (IAEA) has identified the urgency and picked up this topic recently in a 'Coordinated Research Project' on a 'Reference Database for Beta-Delayed Neutron Emission Data'. This project will review, compile, and evaluate the existing data for neutron-branching ratios and half-lives of β-delayed neutron emitters and help to ensure a reliable database for the future discoveries of new isotopes and help to constrain astrophysical and
Online failed fuel identification using delayed neutron detector signals in pool type reactors
International Nuclear Information System (INIS)
Upadhyay, Chandra Kant; Sivaramakrishna, M.; Nagaraj, C.P.; Madhusoodanan, K.
2011-01-01
In todays world, nuclear reactors are at the forefront of modern day innovation and reactor designs are increasingly incorporating cutting edge technology. It is of utmost importance to detect failure or defects in any part of a nuclear reactor for healthy operation of reactor as well as the safety aspects of the environment. Despite careful fabrication and manufacturing of fuel pins, there is a chance of clad failure. After fuel pin clad rupture takes place, it allows fission products to enter in to sodium pool. There are some potential consequences due to this such as Total Instantaneous Blockage (TIB) of coolant and primary component contamination. At present, the failed fuel detection techniques such as cover gas monitoring (alarming the operator), delayed neutron detection (DND-automatic trip) and standalone failed fuel localization module (FFLM) are exercised in various reactors. The first technique is a quantitative measurement of increase in the cover gas activity background whereas DND system causes automatic trip on detecting certain level of activity during clad wet rupture. FFLM is subsequently used to identify the failed fuel subassembly. The later although accurate, but mainly suffers from downtime and reduction in power during identification process. The proposed scheme, reported in this paper, reduces the operation of FFLM by predicting the faulty sector and therefore reducing reactor down time and thermal shocks. The neutron evolution pattern gets modulated because fission products are the delay neutron precursors. When they travel along with coolant to Intermediate heat Exchangers, experienced three effects i.e. delay; decay and dilution which make the neutron pulse frequency vary depending on the location of failed fuel sub assembly. This paper discusses the method that is followed to study the frequency domain properties, so that it is possible to detect exact fuel subassembly failure online, before the reactor automatically trips. (author)
$\\beta$-delayed neutrons from oriented $^{137,139}$I and $^{87,89}$Br nuclei
We propose a world-first measurement of the angular distribution of $\\beta$‐delayed n and $\\gamma$-radiation from oriented $^{137, 139}$I and $^{87,89}$Br nuclei, polarised at low temperature at the NICOLE facility. $\\beta$-delayed neutron emission is an increasingly important decay mechanism as the drip line is approached and its detailed understanding is essential to phenomena as fundamental as the r‐process and practical as the safe operation of nuclear power reactors. The experiments offer sensitive tests of theoretical input concerning the allowed and first‐forbidden $\\beta$‐decay strength, the spin-density of neutron emitting states and the partial wave barrier penetration as a function of nuclear deformation. In $^{137}$I and $^{87}$Br the decay feeds predominantly the ground state of the daughters $^{136}$Xe and $^{86}$Kr whereas in $^{139}$I and $^{89}$Br we will explore the use of n-$\\gamma$- coincidence to study neutron transitions to the first and second excited states in the daughters...
$\\beta$-delayed neutrons from oriented $^{137,139}$I and $^{87,89}$Br nuclei
Grzywacz, Robert; Stone, Nicholas; Köster, Ulli; Singh, Barlaj; Bingham, Carrol; Gaulard, S; Kolos, Karolina; Madurga, Miguel; Nikolov, J; Otsubo, T; Roccia, S; Veskovic, Miroslav; Walker, Phil; Walters, William
2013-01-01
We propose a world-‐first measurement of the angular distribution of $\\beta$-‐delayed n and $\\gamma$- radiation from oriented $^{137, 139}$I and $^{87,89}$Br nuclei, polarised at low temperature at the NICOLE facility. $\\beta$-‐delayed neutron emission is an increasingly important decay mechanism as the drip line is approached and its detailed understanding is essential to phenomena as fundamental as the r‐process and practical as the safe operation of nuclear power reactors. The experiments offer sensitive tests of theoretical input concerning the allowed and first-‐forbidden $\\beta$‐decay strength, the spin-‐density of neutron emitting states and the partial wave barrier penetration as a function of nuclear deformation. In $^{137}$I and $^{87}$Br the decay feeds predominantly the ground state of the daughters $^{136}$Xe and $^{86}$Kr whereas in $^{139}$I and $^{89}$Br we will explore the use of n-$\\gamma$- coincidence to study neutron transitions to the first and second excited state...
Beta-delayed fission and neutron emission calculations for the actinide cosmochronometers
International Nuclear Information System (INIS)
Meyer, B.S.; Howard, W.M.; Mathews, G.J.; Takahashi, K.; Moeller, P.; Leander, G.A.
1989-01-01
The Gamow-Teller beta-strength distributions for 19 neutron-rich nuclei, including ten of interest for the production of the actinide cosmochronometers, are computed microscopically with a code that treats nuclear deformation explicitly. The strength distributions are then used to calculate the beta-delayed fission, neutron emission, and gamma deexcitation probabilities for these nuclei. Fission is treated both in the complete damping and WKB approximations for penetrabilities through the nuclear potential-energy surface. The resulting fission probabilities differ by factors of 2 to 3 or more from the results of previous calculations using microscopically computed beta-strength distributions around the region of greatest interest for production of the cosmochronometers. The indications are that a consistent treatment of nuclear deformation, fission barriers, and beta-strength functions is important in the calculation of delayed fission probabilities and the production of the actinide cosmochronometers. Since we show that the results are very sensitive to relatively small changes in model assumptions, large chronometric ages for the Galaxy based upon high beta-delayed fission probabilities derived from an inconsistent set of nuclear data calculations must be considered quite uncertain
International Nuclear Information System (INIS)
Matthews, I.P.
1979-09-01
Early attempts at determining the elemental composition of the body by radioactive isotope dilution techniques are reviewed. The development and current status of in-vivo neutron activation analysis and the ways in which it supersedes or supplements certain of the former techniques are outlined. An irradiation facility is described which employs a 5 Ci neutron source and is capable of performing prompt and delay γ-ray measurements as well as cyclic activation. The uniformity of thermal neutron flux in a phantom is demonstrated and the neutron spectrum at a depth in the phantom has been obtained by means of threshold detectors. An examination is made of the possible applications of the Monte Carlo method to the design of irradiation and detection facilities and in yielding information about inaccessible areas. Detection limits for the bulk body elements and trace elements are presented. It is shown that the depth of a region of the body can be determined from a prompt gamma ray spectrum. This technique can be used to correct measurements when it is known that activation and detection is non-uniform. The feasibility of using a C.T. whole body scanner to measure bone demineralisation is explored. (author)
On solution to the problem of reactor kinetics with delayed neutrons by Monte Carlo method
International Nuclear Information System (INIS)
Kyncl, Jan
2013-07-01
The initial value problem is addressed for the neutron transport equation and for the system of equations that describe the behaviour of emitters of delayed neutrons. Examination of the solution to this problem is based on several main assumptions concerning the behaviour of macroscopic effective cross-sections describing the reaction of the neutron with the medium, the temperature of medium and the remaining parameters of the equations. Formulation of these assumptions is adequately general and is in agreement with the properties of all known models of the physical quantities involved. Among others, the assumptions admit dependence of the macroscopic effective cross-sections and temperature on spatial coordinates and time that can be arbitrary to a great extent. The problem starts from a set of integro-differential equations. This problem is first transposed into the equivalent problem of solving a linear integral equation for neutron flux. This integral equation is solved by the method of successive iterations and its uniqueness is demonstrated. Numeric solution to the integral equation by Monte Carlo method consists in finding a functional of the exact solution. For this, a random process is set up and some random variables are proposed. Then it is demonstrated that each of these variables is an unbiased estimator of that functional. (author)
International Nuclear Information System (INIS)
Hashimoto, Kengo; Mouri, Tomoaki; Ohtani, Nobuo
1999-01-01
The difference-filtering correlation analysis was applied to time-sequence neutron count data measured in a slightly subcritical assembly, where the Feynman-α analysis suffered from large contribution of delayed neutron to the variance-to-mean ratio of counts. The prompt-neutron decay constant inferred from the present filtering analysis agreed very closely with that by pulsed neutron experiment, and no dependence on the gate-time range specified could be observed. The 1st-order filtering was sufficient for the reduction of the delayed-neutron contribution. While the conventional method requires a choice of analysis formula appropriate to a gate-time range, the present method is applicable to a wide variety of gate-time ranges. (author)
International Nuclear Information System (INIS)
Balestrini, S.J.; Balagna, J.P.; Menlove, H.O.
1976-01-01
Two specialized neutron-sensitive detectors are described which are employed for rapid assays of fissionable elements by sensing for delayed neutrons emitted by samples after they have been irradiated in a nuclear reactor. The more sensitive of the two detectors, designed to assay for uranium in water samples, is 40% efficient; the other, designed for sediment sample assays, is 27% efficient. These detectors are also designed to operate under water as an inexpensive shielding against neutron leakage from the reactor and neutrons from cosmic rays. (Auth.)
International Nuclear Information System (INIS)
Carl, U.M.
1987-01-01
The authors have compared the effects of mixed fractionation schedules with X rays and neutrons on growth delay of a murine tumour and skin reactions in mice. The schedules were five daily fractions of X rays, neutrons or mixtures (NNXXX, XXXNN or NXXXN). For clamped tumours or skin all three mixed schedules had the same effect. In contrast, for unclamped tumours giving the neutrons first (NNXXX) was more effective than the other two mixed schedules. This represented a true therapeutic gain and implies that if neutrons are used clinically as only part of a course of fractionated radiotherapy, they should be given at the beginning rather than at the end of treatment. (author)
Characteristics of uranium oxide cathode for neutron streak camera
International Nuclear Information System (INIS)
Niki, H.; Itoga, K.; Yamanaka, M.; Yamanaka, T.; Yamanaka, C.
1986-01-01
In laser fusion research, time-resolved neutron measurements require 20ps resolution in order to obtain the time history of the D-T burn. Uranium oxide was expected to be a sensitive material as a cathode of a neutron streak camera because of its large fission cross section. The authors report their measurements of some characteristics of the uranium oxide cathode connected to a conventional streak tube. 14 MeV neutron signal were observed as the bright spots on a TV monitor using a focus mode opration. Detection efficiency was ∼ 1 x 10 -6 for 1 μm thick cathode. Each signal consisted of more than several tens of components, which were corresponding to the secondary electrons dragged out from the cathode by a fission fragment. Time resolution is thought to be limited mainly by the transit time spread of the secondary electrons. 14ps resolution was obtained by a streak mode operation for a single fission event
Summary Report of Consultants' Meeting on Beta-Delayed Neutron Emission Evaluation
International Nuclear Information System (INIS)
Abriola, Daniel; Singh, Balraj; Dillmann, Iris
2011-12-01
A summary is given of a Consultants' Meeting assembled to assess the viability of a new IAEA Co-ordinated Research Project (CRP) on Beta-delayed neutron emission evaluation. The current status of the field was reviewed, cases in which new measurements are needed were identified and the current theoretical models were examined. The best known cases were selected as standards and were assessed and preliminary best values of the emission probabilities were obtained. The need of such a CRP was strongly agreed. Both the technical discussions and the expected outcome of such a project are described, along with detailed recommendations for its implementation. (author)
International Nuclear Information System (INIS)
Lee, Y.K.; Hugot, F.X.
2011-01-01
The effective delayed neutron fraction βeff is an important reactor physics parameter. Its calculation within the multi-group deterministic transport code can be performed with the aid of adjoint flux weighted integrations. However, in continuous energy Monte Carlo transport code, the adjoint weighted βeff calculation becomes complicated due to the backward treatment of the anisotropy scattering. In TRIPOLI-4 continuous energy Monte Carlo code, the βeff calculation was performed by a two-run method, one run with delayed neutrons and second with only the contribution from prompt fission neutrons. To improve the uncertainty of the βeff two-run calculation for the experimental reactors, two simple and fast one-run methods to estimate the βeff in the continuous energy simulation have been implemented into the TRIPOLI-4 code. First approach is an improved one of the Bretscher's prompt method and second one based on the proposal of Nauchi and Kameyama. In these one-run methods, the prompt and the delayed neutrons are first tagged. Their tracking and statistics are separated performed. The new βeff calculations have been optimized in the power iteration cycles so as to estimate the production of prompt and delayed neutrons from the prompt and delayed neutrons of previous generation. To validate the new βeff calculation by TRIPOLI-4, several benchmarks including fast and thermal systems have been considered. In this paper the recent measurements of βeff in the research reactor IPEN/MB-01 have been benchmarked. The basic components of the βeff and the Keff have been also calculated so as to understand the influences of the cross sections and the delayed neutron yields on the reactor reactivity calculations. Three nuclear data libraries, ENDF/BVI.r4, ENDF/B-VII.0, and JEFF-3.1 were taken into account in this study. (author)
International Nuclear Information System (INIS)
Piksajkin, V.M.; Kazakov, L.E.; Isaev, S.T.; Korolev, G.G.; Roshchenko, V.A.; Tertychnyj, R.G.
2002-01-01
Relative yield and group period of delayed neutrons induced by the 239 Pu fission in the 0.37-4.97 MeV range were measured. Comparative analysis of experimental data was conducted in terms of middle period of half-life of delayed neutron nuclei-precursors. Character and scale of changing values of delayed neutron group parameters as changing excitation energy of fission compound-nucleus have been demonstrated for the first time. Considerable energy dependence of group parameters under the neutron induced 239 Pu fission that was expressed by the decreasing middle period of half-life of nuclei-precursors by 10 % in the 2.85 eV - 5 MeV range of virgin neutrons was detected [ru
Electron-capture delayed fission properties of neutron-deficient einsteinium nuclei
Energy Technology Data Exchange (ETDEWEB)
Shaughnessy, Dawn A. [Univ. of California, Berkeley, CA (United States)
2000-01-01
Electron-capture delayed fission (ECDF) properties of neutron-deficient einsteinium isotopes were investigated using a combination of chemical separations and on-line radiation detection methods. ^{242}Es was produced via the ^{233}U(^{14}N,5n)^{242}Es reaction at a beam energy of 87 MeV (on target) in the lab system, and was found to decay with a half-life of 11 ± 3 seconds. The ECDF of ^{242}Es showed a highly asymmetric mass distribution with an average pre-neutron emission total kinetic energy (TKE) of 183 ± 18 MeV. The probability of delayed fission (P_{DF}) was measured to be 0.006 ± 0.002. In conjunction with this experiment, the excitation functions of the ^{233}U(^{14}N,xn)^{247-x}Es and ^{233}U^{(15N,xn)}^{248-x}Es reactions were measured for ^{243}Es, ^{244}Es and ^{245}Es at projectile energies between 80 MeV and 100 MeV.
Electron-capture delayed fission properties of neutron-deficient einsteinium nuclei
International Nuclear Information System (INIS)
Shaughnessy, Dawn A.
2000-01-01
Electron-capture delayed fission (ECDF) properties of neutron-deficient einsteinium isotopes were investigated using a combination of chemical separations and on-line radiation detection methods. 242 Es was produced via the 233 U( 14 N,5n) 242 Es reaction at a beam energy of 87 MeV (on target) in the lab system, and was found to decay with a half-life of 11 ± 3 seconds. The ECDF of 242 Es showed a highly asymmetric mass distribution with an average pre-neutron emission total kinetic energy (TKE) of 183 ± 18 MeV. The probability of delayed fission (P DF ) was measured to be 0.006 ± 0.002. In conjunction with this experiment, the excitation functions of the 233 U( 14 N,xn) 247-x Es and 233 U( 15 N,xn) 248-x Es reactions were measured for 243 Es, 244 Es and 245 Es at projectile energies between 80 MeV and 100 MeV
Characteristics of GaAs MESFET inverters exposed to high energy neutrons
International Nuclear Information System (INIS)
Bloss, W.L.; Yamada, W.E.; Young, A.M.; Janousek, B.K.
1988-01-01
GaAs MESFET circuits have been exposed to high energy neutrons with fluences ranging from 1x10/sup 14/ n/cm/sup 2/ to 2x10/sup 15/ m/cm/sup 2/. Discrete transistors, inverters, and ring oscillators were characterized at each fluence. While the MESFETs exhibit significant threshold voltage shifts and transconductance and saturation current degradation over this range of neutron fluences, the authors have observed improvement in the DC characteristics of Schottky Diode FET Logic (SDFL) inverters. This unusual result has been successfully simulated using device parameters extracted from FETs damaged by exposure to high energy neutrons. Although the decrease in device transconductance results in an increase in inverter gate delay, as reflected in ring oscillator frequency measurements, the authors conclude that GaAs ICs fabricated from this logic family will remain functional after exposure to extreme neutron fluences. This is a consequence of the observed improvement in inverter noise margin evident in both measured and simulated circuit performance
Beta-decay rate and beta-delayed neutron emission probability of improved gross theory
Koura, Hiroyuki
2014-09-01
A theoretical study has been carried out on beta-decay rate and beta-delayed neutron emission probability. The gross theory of the beta decay is based on an idea of the sum rule of the beta-decay strength function, and has succeeded in describing beta-decay half-lives of nuclei overall nuclear mass region. The gross theory includes not only the allowed transition as the Fermi and the Gamow-Teller, but also the first-forbidden transition. In this work, some improvements are introduced as the nuclear shell correction on nuclear level densities and the nuclear deformation for nuclear strength functions, those effects were not included in the original gross theory. The shell energy and the nuclear deformation for unmeasured nuclei are adopted from the KTUY nuclear mass formula, which is based on the spherical-basis method. Considering the properties of the integrated Fermi function, we can roughly categorized energy region of excited-state of a daughter nucleus into three regions: a highly-excited energy region, which fully affect a delayed neutron probability, a middle energy region, which is estimated to contribute the decay heat, and a region neighboring the ground-state, which determines the beta-decay rate. Some results will be given in the presentation. A theoretical study has been carried out on beta-decay rate and beta-delayed neutron emission probability. The gross theory of the beta decay is based on an idea of the sum rule of the beta-decay strength function, and has succeeded in describing beta-decay half-lives of nuclei overall nuclear mass region. The gross theory includes not only the allowed transition as the Fermi and the Gamow-Teller, but also the first-forbidden transition. In this work, some improvements are introduced as the nuclear shell correction on nuclear level densities and the nuclear deformation for nuclear strength functions, those effects were not included in the original gross theory. The shell energy and the nuclear deformation for
International Nuclear Information System (INIS)
Apperson, C.E. Jr.
1981-01-01
A method is presented for studying the influence of fission product transpot on delayed neutron precursors and decay heat sources during Liquid Metal Fast Breeder Reactor (LMFBR) unprotected accidents. The model represents the LMFBR core as a closed homogeneous cell. Thermodynamic phase equilibrium theory is used to predict fission product mobility. Reactor kinetics behavior is analyzed by an extension of point kinetics theory. Group dependent delayed neutron precursor and decay heat source retention factors, which represent the fraction of each group retained in the fuel, are developed to link the kinetics and thermodynamics analysis. Application of the method to a highly simplified model of an unprotected loss-of-flow accident shows a time delay on the order of 10 ms is introduced in the predisassembly power history if fission product motion is considered when compared to the traditional transient solution. The post-transient influence of fission product transport calculated by the present model is a 24 percent reduction in the decay heat level in the fuel material which is similar to traditional approximations. Isotopes of the noble gases, Kr and Xe, and the elements I and Br are shown to be very mobile and are responsible for a major part of the observed effects. Isotopes of the elements Cs, Se, Rb, and Te were found to be moderately mobile and contribute to a lesser extent to the observed phenomena. These results obtained from the application of the described model confirm the initial hypothesis that sufficient fission product transport can occur to influence a transient. For these reasons, it is concluded that extension of this model into a multi-cell transient analysis code is warranted
Energy Technology Data Exchange (ETDEWEB)
Timis, C.N
2001-07-01
A new detection array for beta delayed neutrons was built. It includes up to 32 plastic scintillation counters 180 cm long located at 120 cm from the target. Neutron energy spectra are measured by time-of-flight in the 300 keV-15 MeV range with good energy resolution. The device was tested with several known nuclei. Its performances are discussed in comparison with Monte Carlo simulations. They very high overall detection efficiency on the TONNERRE array made it possible to study one and two neutron emission of {sup 11}Li. A complete decay scheme was obtained. The {sup 33}Mg and {sup 35}Al beta decays were investigated for the first time by neutron and gamma spectroscopy. Complete decay schemes were established and compared to large scale shell-model calculations. (authors)
International Nuclear Information System (INIS)
Aufiero, Manuele; Brovchenko, Mariya; Cammi, Antonio; Clifford, Ivor; Geoffroy, Olivier; Heuer, Daniel; Laureau, Axel; Losa, Mario; Luzzi, Lelio; Merle-Lucotte, Elsa; Ricotti, Marco E.; Rouch, Hervé
2014-01-01
Highlights: • Calculation of effective delayed neutron fraction in circulating-fuel reactors. • Extension of the Monte Carlo SERPENT-2 code for delayed neutron precursor tracking. • Forward and adjoint multi-group diffusion eigenvalue problems in OpenFOAM. • Analytical approach for β eff calculation in simple geometries and flow conditions. • Good agreement among the three proposed approaches in the MSFR test-case. - Abstract: This paper deals with the calculation of the effective delayed neutron fraction (β eff ) in circulating-fuel nuclear reactors. The Molten Salt Fast Reactor is adopted as test case for the comparison of the analytical, deterministic and Monte Carlo methods presented. The Monte Carlo code SERPENT-2 has been extended to allow for delayed neutron precursors drift, according to the fuel velocity field. The forward and adjoint eigenvalue multi-group diffusion problems are implemented and solved adopting the multi-physics tool-kit OpenFOAM, by taking into account the convective and turbulent diffusive terms in the precursors balance. These two approaches show good agreement in the whole range of the MSFR operating conditions. An analytical formula for the circulating-to-static conditions β eff correction factor is also derived under simple hypotheses, which explicitly takes into account the spatial dependence of the neutron importance. Its accuracy is assessed against Monte Carlo and deterministic results. The effects of in-core recirculation vortex and turbulent diffusion are finally analysed and discussed
Neutronic characteristics of coupled moderator proposed in integrated model
International Nuclear Information System (INIS)
Teshigawara, Makoto; Meigo, Shin-ichiro; Sakata, Hideaki; Kai, Tetsuya; Harada, Masahide; Ikeda, Yujiro; Watanabe, Noboru
2001-05-01
A pulsed spallation source for the materials science and the life science is currently developing for its construction in the High Intensity Proton Accelerator Project proposed jointly by the Japan Atomic Energy Research Institute (JAERI) and the High Energy Accelerator Research Organization (KEK). This report presents the analytical results of the neutronic characteristics of the coupled moderator based on the analytical results obtained by using an integrated model which has established on the extensive neutronic and technical study. Total heat deposition in a hydrogen (H 2 ) moderator working as the main moderator was about 420 W/MW. Maximum nuclear heat density in the H 2 moderator was about 1 W/cm 3 /MW. Also total heat deposition in a premoderator was about 9.2 kW/MW. The heat density of the premoderator was comparable to that of the moderator vessel made of aluminum alloy. The heat density of the premoderator and the moderator vessel is about 1.2-2 times higher than that of the hydrogen moderator. The temperature from 300 K to 400 K of the premoderator did not affect on neutron intensity of the H 2 moderator. This suggested an engineering advantage on the thermal and hydraulic design. 6000 or 7000 type of a aluminum alloy was considered from the viewpoint of the neutron beam transmission. The proton beams scattered by the proton beam window did not affect on the nuclear heating in the H 2 moderator. The heat deposition in the H 2 moderator and the neutron intensity of the H 2 moderator did not depend on the proton beam profile but it did on the distance between the proton beam and the moderator. (author)
Energy Technology Data Exchange (ETDEWEB)
Suratman,; Purwanto,; Sukarman-Aminjoyo, [Yogyakarta Nuclear Research Centre, National Atomic Energy Agency, Yogyakarta (Indonesia)
1996-04-15
A bioassay method for uranium in urine by neutron counting has been studied. The aim of this research is to obtain a bioassay method for uranium in urine which is used for the determination of internal dose of radiation workers. The bioassay was applied to the artificially uranium contaminated urine. The weight of the contaminant was varied. The uranium in the urine was irradiated in the Kartini reactor core, through pneumatic system. The delayed neutron was counted by BF3 neutron counter. Recovery of the bioassay was between 69.8-88.8 %, standard deviation was less than 10 % and the minimum detection was 0.387 {mu}g.
International Nuclear Information System (INIS)
Bazin, D.
1987-07-01
Among the nuclear mechanisms used for the production of nuclei far from stability, the projectile fragmentation process has recently proved its efficiency. However, at Fermi energies, one has to take into account some collective and relaxation effects which drastically modify the production cross-sections. The spectroscopic study of very neutron-rich nuclei is very dependent of these production rates. A study of beta-delayed neutron emission which leads to new measurements of half-lives and neutron delayed emission probabilities is achieved with a liquid scintillator detector. The results which are then compared to different theories are of interest for the understanding of natural production of heavy elements (r processus) [fr
International Nuclear Information System (INIS)
Stromswold, D.C.; Peurrung, A.J.; Reeder, P.L.; Perkins, R.W.
1996-06-01
Feasibility experiments conducted at Pacific Northwest National Laboratory demonstrate that either delayed neutrons or energetic gamma rays from short-lived fission products can be used to monitor the blending of UF 6 gas streams. A 252 Cf neutron source was used to induce 235 U fission in a sample, and delayed neutrons and gamma rays were measured after the sample moved open-quotes down-stream.close quotes The experiments used a UO 2 powder that was transported down the pipe to simulate the flowing UF 6 gas. Computer modeling and analytic calculation extended the test results to a flowing UF 6 gas system. Neutron or gamma-ray measurements made at two downstream positions can be used to indicate both the 235 U content and UF 6 flow rate. Both the neutron and gamma-ray techniques have the benefits of simplicity and long-term reliability, combined with adequate sensitivity for low-intrusion monitoring of the blending process. Alternatively, measuring the neutron emission rate from (a, n) reactions in the UF 6 provides an approximate measure of the 235 U content without using a neutron source to induce fission
International Nuclear Information System (INIS)
Menlove, H.O.
1981-08-01
An active neutron interrogation method has been developed for the measurement of 235 U content in fresh fuel assemblies. The neutron Coincidence Collar uses neutron interrogation with an AmLi neutron source and coincidence counting the induced fission reaction neutrons from the 235 U. This manual describes the system components, operation, and performance characteristics. Applications of the Coincidence Collar to PWR and BWR types of reactor fuel assemblies are described
Benchmark experiments of effective delayed neutron fraction βeff at FCA
International Nuclear Information System (INIS)
Sakurai, Takeshi; Okajima, Shigeaki
1999-01-01
Benchmark experiments of effective delayed neutron fraction β eff were performed at Fast Critical Assembly (FCA) in the Japan Atomic Energy Research Institute. The experiments were made in three cores providing systematic change of nuclide contribution to the β eff : XIX-1 core fueled with 93% enriched uranium, XIX-2 core fueled with plutonium and uranium (23% enrichment) and XIX-3 core fueled with plutonium (92% fissile Pu). Six organizations from five countries participated in these experiments and measured the β eff by using their own methods and instruments. Target accuracy in the β eff was achieved to be better than ±3% by averaging the β eff values measured using a wide variety of experimental methods. (author)
Delayed Fission Gamma-ray Characteristics of Th-232 U-233 U-235 U-238 and Pu-239
Energy Technology Data Exchange (ETDEWEB)
Lane, Taylor [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Parma, Edward J. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)
2015-08-01
Delayed fission gamma-rays play an important role in determining the time dependent ioniz- ing dose for experiments in the central irradiation cavity of the Annular Core Research Reactor (ACRR). Delayed gamma-rays are produced from both fission product decay and from acti- vation of materials in the core, such as cladding and support structures. Knowing both the delayed gamma-ray emission rate and the time-dependent gamma-ray energy spectrum is nec- essary in order to properly determine the dose contributions from delayed fission gamma-rays. This information is especially important when attempting to deconvolute the time-dependent neutron, prompt gamma-ray, and delayed gamma-ray contribution to the response of a diamond photo-conducting diode (PCD) or fission chamber in time frames of milliseconds to seconds following a reactor pulse. This work focused on investigating delayed gamma-ray character- istics produced from fission products from thermal, fast, and high energy fission of Th-232, U-233, U-235, U-238, and Pu-239. This work uses a modified version of CINDER2008, a transmutation code developed at Los Alamos National Laboratory, to model time and energy dependent photon characteristics due to fission. This modified code adds the capability to track photon-induced transmutations, photo-fission, and the subsequent radiation caused by fission products due to photo-fission. The data is compared against previous work done with SNL- modified CINDER2008 [ 1 ] and experimental data [ 2 , 3 ] and other published literature, includ- ing ENDF/B-VII.1 [ 4 ]. The ability to produce a high-fidelity (7,428 group) energy-dependent photon fluence at various times post-fission can improve the delayed photon characterization for radiation effects tests at research reactors, as well as other applications.
Energy Technology Data Exchange (ETDEWEB)
Perret, Gregory [Paul Scherrer Institute, Villigen, 5232, (Switzerland)
2015-07-01
The critical decay constant (B/A), delayed neutron fraction (B) and generation time (A) of the Minerve reactor were measured by the Paul Scherrer Institut (PSI) and the Commissariat a l'Energie Atomique (CEA) in September 2014 using the Feynman-alpha and Power Spectral Density neutron noise measurement techniques. Three slightly subcritical configuration were measured using two 1-g {sup 235}U fission chambers. This paper reports on the results obtained by PSI in the near critical configuration (-2g). The most reliable and precise results were obtained with the Cross-Power Spectral Density technique: B = 708.4±9.2 pcm, B/A = 79.0±0.6 s{sup -1} and A 89.7±1.4 micros. Predictions of the same kinetic parameters were obtained with MCNP5-v1.6 and the JEFF-3.1 and ENDF/B-VII.1 nuclear data libraries. On average the predictions for B and B/A overestimate the experimental results by 5% and 11%, respectively. The discrepancy is suspected to come from either a corruption of the data or from the inadequacy of the point kinetic equations to interpret the measurements in the Minerve driven system. (authors)
International Nuclear Information System (INIS)
Kobayashi, T.; Sakurai, Y.; Kanda, K.
2001-01-01
The neutron irradiation characteristics of the Heavy Water Neutron Irradiation Facility (HWNIF) at the Kyoto University Research Reactor Institute (KIJRRI) for boron neutron capture therapy (BNCT), is described. The present method of dose measurement and its evaluation at the KURRI, is explained. Especially, the special feature and noticeable matters were expounded for the BNCT with craniotomy, which has been applied at present only in Japan. (author)
Neutronic characteristics simulation of LMFBR of great size
International Nuclear Information System (INIS)
Kim, Y.C.
1987-09-01
The CONRAD experimental program to be executed on the critical mockup MASURCA in Cadarache and use all the european plutonium stock. The objectives of this program are to reduce the uncertainties on important project parameters such as the reactivity value of control rods, the flux distribution to valid calcul methods and data to use for new LMFBR conception (heterogeneous axial core by example) and to resolve the neutronic control problems for a LMFBR of great size. The present study has permitted to define this program and its physical characteristics [fr
International Nuclear Information System (INIS)
Klapdor, H.V.
1976-01-01
Recent results in β-delayed neutron emission are interpreted by structure of the Gamow-Teller giant resonance not included in the 'gross-theory' of β-decay. Inclusion of this structure of the β-decay function is important for calculations of β-decay production rates for heavy nuclides by astrophysical processes and thermonuclear explosions. (Auth.)
Delayed effects of neutron irradiation on central nervous system microvasculature in the rat
International Nuclear Information System (INIS)
Goodman, J.H.; McGregor, J.M.; Clendenon, N.R.; Gordon, W.A.; Yates, A.J.; Gahbauer, R.A.; Barth, R.F.; Fairchild, R.G.
1988-01-01
Pathologic examination of a series of 14 patients with malignant gliomas treated with BNCT showed well demarcated zones of radiation damage characterized by coagulation necrosis. Beam attenuation was correlated with edema, loss of parenchymal elements, demyelination, leukocytosis, and peripheral gliosis. Vascular disturbances consisted of endothelial swelling, medial and adventitial proliferation, fibrin impregnation, frequent thrombosis, and perivascular inflammation. Radiation changes appeared to be acute and delayed. The outcome of the patients in this series was not significantly different from the natural course of the disease, even though two of the patients had no residual tumor detected at the time of autopsy. The intensity of the vascular changes raised a suspicion that boron may have sequestered in vessel walls, resulting in selectively high doses of radiation to these structures (Asbury et al., 1972), or that there may have been high blood concentrations of boron at the time of treatment. The potential limiting effects of a vascular ischemic reaction in Boron Neutron Capture Therapy (BNCT) prompted the following study to investigate the delayed response of microvascular structures in a rat model currently being used for pre-clinical investigations. 8 refs., 3 figs., 1 tab
Noise characteristics of neutron images obtained by cooled CCD device
International Nuclear Information System (INIS)
Taniguchi, Ryoichi; Sasaki, Ryoya; Okuda, Shuichi; Okamoto, Ken-Ichi; Ogawa, Yoshihiro; Tsujimoto, Tadashi
2009-01-01
The noise characteristics of a cooled CCD device induced by neutron and gamma ray irradiation have been investigated. In the cooled CCD images, characteristic white spot noises (CCD noise) frequently appeared, which have a shape like a pixel in most cases and their brightness is extremely high compared with that of the image pattern. They could be divided into the two groups, fixed pattern noise (FPN) and random noise. The former always appeared in the same position in the image and the latter appeared at any position. In the background image, nearly all of the CCD noises were found to be the FPN, while many of them were the random noise during the irradiation. The random CCD noises increased with irradiation and decreased soon after the irradiation. In the case of large irradiation, a part of the CCD noise remained as the FPN. These facts suggest that the CCD noise is a phenomenon strongly relating to radiation damage of the CCD device.
Energy Technology Data Exchange (ETDEWEB)
Francisco, J L. de; Perez-Navarro, A; Rodriguez-Mayquez, E
1973-07-01
The effective delayed neutron fraction, {beta} eff, has been determined from the measurement of E / {beta}{sup 2}, by means of reactor noise analysis in the time domain, and the neutron detector efficiency, {epsilon}. For the {epsilon} measurement it is necessary to determine the fission rate in the reactor. This value can be obtained from the absolute measurement of the fission rate per cm{sup 3}, at a certain point of the reactor, and the determination of these two values ratio, which has been calculated by the Monte Cario method and also measured with results in good agreement. (Author)
International Nuclear Information System (INIS)
Irvine, J.M.
1978-01-01
The subject is covered in chapters entitled: introduction (resume of stellar evolution, gross characteristics of neutron stars); pulsars (pulsar characteristics, pulsars as neutron stars); neutron star temperatures (neutron star cooling, superfluidity and superconductivity in neutron stars); the exterior of neutron stars (the magnetosphere, the neutron star 'atmosphere', pulses); neutron star structure; neutron star equations of state. (U.K.)
Characteristics of polyethylene-moderated 252Cf neutron sources
International Nuclear Information System (INIS)
Alejnikov, V.E.; Beskrovnaya, L.G.; Florko, B.V.
2000-01-01
Polyethylene-moderated 252 Cf neutron sources were designed to produce neutron reference fields' spectra that simulate the spectra observed in the workplaces within nuclear reactors and accelerators. The paper describes the neutron sources and fields. Neutron spectra were calculated by Monte Carlo method and compared with experimental data
The characteristic calibration of the plastic scintillation detector for neutron diagnostic
Chen Hong Su
2002-01-01
The author presents the characteristic of the plastic scintillation detector used for pulse neutron diagnostic. The detection efficiency and sensitivity of the detector to DT neutron have been calibrated by the K-400 accelerator and by the pulse neutron tube, separately. The detection efficiency from the experiment is in agreement with that from calculation in the range of experimental errors
Energy Technology Data Exchange (ETDEWEB)
Martin, Thierry [Lab. de Physique Corpusculaire, Caen Univ., 14 - Caen (France)
1998-11-09
This work has been undertaken within a French-Romanian collaboration in order to build a high efficiency detector array for delayed neutrons: barrel-shaped TONNERRE. Some neutron-rich nuclei decay through 1, 2 or 3 neutron emission after {beta}{sup -} decay. More exotic nuclei will be produced by SPIRAL at GANIL. An array with high efficiency and good resolution is then required. Thirty two BC400 plastic scintillators (160 x 20 x 4 cm{sup 3}) allow us to get the time of flight neutron spectra. They are bent for uniform flight path and viewed by a photomultiplier tube at both ends. Simulations have allowed to establish scintillator size and to minimize light attenuation. Intrinsic efficiency and crosstalk have been measured with {sup 252}Cf and compared to GEANT. 1 to 5 MeV neutrons are detected with good timing and position properties. Other counters will be built for neutrons from 300 keV to 1 MeV. Planned to run at several particle accelerators (GANIL, CERN, and others), TONNERRE is modular and many geometries are possible. (author) 48 refs., 78 figs., 20 tabs.
International Nuclear Information System (INIS)
Onega, R.J.; Florian, R.J.
1983-01-01
The delayed-neutron energy spectra for LMFBRs are not as well known as those for LWRs. These spectra are necessary for kinetics calculations which play an important role in safety and accident analyses. A sensitivity analysis was performed to study the response of the reactor power and power density to uncertainties in the delayed-neutron spectra during a rod-ejection accident. The accidents studied were central control-rod-ejections with ejection times of 2,10 and 30s. A two-energy group and two-precursor group model was formulated for the International Nuclear Fuel Cycle Evaluation (INFCE) reference design MOX-fueled LMFBR. The sensitivity analysis is based on the use of adjoints so that it is not necessary to repeatedly solve the governing (kinetics) equations to obtain the sensitivity derivatives. This is of particular importance when large systems of equations are used. The power and power-density responses were found to be most sensitive to uncertainties in the spectrum of the second delayed-neutron precursor group, resulting from the fission of 238 U, producing neutrons in the first energy group. It was found, for example, that for a rod-ejection time of 30s, and uncertainty of 7.2% in the fast components of the spectra resulted in a 24% uncertainty in the predicted power and power density. These responses were recalculated by repeatedly solving the kinetics equations. The maximum discrepancy between the recalculated and the sensitivity analysis response was only 1.6%. The results of the sensitivity analysis indicate the need for improved delayed-neutron spectral data in order to reduce the uncertainties in accident analyses. (author)
Energy Technology Data Exchange (ETDEWEB)
Vega C, H.R.; Manzanares A, E.; Hernandez D, V.M.; Mercado S, G.A. [Universidad Autonoma de Zacatecas, A.P. 336, 98000 Zacatecas (Mexico); Gallego D, E.; Lorente F, A. [Universidad Politecnica de Madrid, C/Jose Gutierrez Abascal 2, E-28006 Madrid (Spain)
2005-07-01
By means of Monte Carlo methods the spectra of the produced neutrons {sup 252} Cf, {sup 252} Cf/D{sub 2}O, {sup 241} Am Be, {sup 239} Pu Be, {sup 140} La Be, {sup 239} Pu{sup 18}O{sub 2} and {sup 226} Ra Be have been calculated. With the information of the spectrum it was calculated the average energy of the neutrons of each source. By means of the fluence coefficients to dose it was determined, for each one of the studied sources, the fluence factors to dose. The calculated doses were H, H{sup *}(10), H{sub p,sIab} (10, 0{sup 0}), E{sub AP} and E{sub ISO}. During the phase of the calculations the sources were modeled as punctual and their characteristics were determined to 100 cm in the hole. Also, for the case of the sources of {sup 239} Pu Be and {sup 241} Am Be, were carried out calculations modeling the sources with their respective characteristics and the dosimetric properties were determined in a space full with air. The results of this last phase of the calculations were compared with the experimental results obtained for both sources. (Author)
International Nuclear Information System (INIS)
Roschenko, V.A.; Piksaikin, V.M.; Kazakov, L.E.; Isaev, S.G.; Korolev, G.G.; Tarasko, M.Z.; Tertychnyi, R.G.
2001-01-01
The fundamental role of delayed neutrons in behavior, control and safety of reactors is well known today. Delayed neutron data are of great interest not only for reactor physics but also for nuclear fission physics and astrophysics. The purpose of the present work was the measurement of energy dependence of delayed neutrons (DN) group parameters at fission of nuclei 239 Pu in a range of energies of primary neutrons from 0.37 up to 5 MeV. The measurements were executed on installation designed on the basis of the electrostatic accelerator of KG - 2.5 SSC RF IPPE. The data are obtained in 6-group representation. It is shown, that there is a significant energy dependence of DN group parameters in a range of primary neutrons energies from thermal meanings up to 5 MeV, which is expressed in reduction of the average half-life of nuclei of the DN precursors on 10 %. The data, received in the present work, can be used at creation of a set of group constants for reactors with an intermediate spectrum of neutrons. (authors)
Liamsuwan, T.; Wonglee, S.; Channuie, J.; Esoa, J.; Monthonwattana, S.
2017-06-01
The objective of this work was to systematically investigate the response characteristics of optically stimulated luminescence Albedo neutron (OSLN) dosimeters to ensure reliable personal dosimetry service provided by Thailand Institute of Nuclear Technology (TINT). Several batches of InLight® OSLN dosimeters were irradiated in a reference neutron field generated by the in-house 241AmBe neutron irradiator. The OSL signals were typically measured 24 hours after irradiation using the InLight® Auto 200 Reader. Based on known values of delivered neutron dose equivalent, the reading correction factor to be used by the reader was evaluated. Subsequently, batch homogeneity, dose linearity, lower limit of detection and fading of the OSLN dosimeters were examined. Batch homogeneity was evaluated to be 0.12 ± 0.05. The neutron dose response exhibited a linear relationship (R2=0.9974) within the detectable neutron dose equivalent range under test (0.4-3 mSv). For this neutron field, the lower limit of detection was between 0.2 and 0.4 mSv. Over different post-irradiation storage times of up to 180 days, the readings fluctuated within ±5%. Personal dosimetry based on the investigated OSLN dosimeter is considered to be reliable under similar neutron exposure conditions, i.e. similar neutron energy spectra and dose equivalent values.
International Nuclear Information System (INIS)
Perry, R.T.; Wilson, W.B.; Charlton, W.S.
1998-04-01
In many systems, it is imperative to have accurate knowledge of all significant sources of neutrons due to the decay of radionuclides. These sources can include neutrons resulting from the spontaneous fission of actinides, the interaction of actinide decay α-particles in (α,n) reactions with low- or medium-Z nuclides, and/or delayed neutrons from the fission products of actinides. Numerous systems exist in which these neutron sources could be important. These include, but are not limited to, clean and spent nuclear fuel (UO 2 , ThO 2 , MOX, etc.), enrichment plant operations (UF 6 , PuF 4 , etc.), waste tank studies, waste products in borosilicate glass or glass-ceramic mixtures, and weapons-grade plutonium in storage containers. SOURCES-3A is a computer code that determines neutron production rates and spectra from (α,n) reactions, spontaneous fission, and delayed neutron emission due to the decay of radionuclides in homogeneous media (i.e., a mixture of α-emitting source material and low-Z target material) and in interface problems (i.e., a slab of α-emitting source material in contact with a slab of low-Z target material). The code is also capable of calculating the neutron production rates due to (α,n) reactions induced by a monoenergetic beam of α-particles incident on a slab of target material. Spontaneous fission spectra are calculated with evaluated half-life, spontaneous fission branching, and Watt spectrum parameters for 43 actinides. The (α,n) spectra are calculated using an assumed isotropic angular distribution in the center-of-mass system with a library of 89 nuclide decay α-particle spectra, 24 sets of measured and/or evaluated (α,n) cross sections and product nuclide level branching fractions, and functional α-particle stopping cross sections for Z < 106. The delayed neutron spectra are taken from an evaluated library of 105 precursors. The code outputs the magnitude and spectra of the resultant neutron source. It also provides an
Measured and calculated effective delayed neutron fraction of the IPR-R1 Triga reactor
Energy Technology Data Exchange (ETDEWEB)
Souza, Rose Mary G.P.; Dalle, Hugo M.; Campolina, Daniel A.M., E-mail: souzarm@cdtn.b, E-mail: dallehm@cdtn.b, E-mail: campolina@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)
2011-07-01
The effective delayed neutron fraction, {beta}{sub eff}, one of the most important parameter in reactor kinetics, was measured for the 100 kW IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil. The current reactor core has 63 fuel elements, containing about 8.5% and 8% by weight of uranium enriched to 20% in U{sup 235}. The core has cylindrical configuration with an annular graphite reflector. Since the first criticality of the reactor in November 1960, the core configuration and the number of fuel elements have been changed several times. At that time, the reactor power was 30 kW, there were 56 fuel elements in the core, and the {beta}{sub eff} value for the reactor recommended by General Atomic (manufacturer of TRIGA) was 790 pcm. The current {beta}{sub eff} parameter was determined from experimental methods based on inhour equation and on the control rod drops. The estimated values obtained were (774 {+-} 38) pcm and (744 {+-} 20) pcm, respectively. The {beta}{sub eff} was calculated by Monte Carlo transport code MCNP5 and it was obtained 747 pcm. The calculated and measured values are in good agreement, and the relative percentage error is -3.6% for the first case, and 0.4% for the second one. (author)
Reactor kinetics calculated in the summation method and key delayed-neutron data
International Nuclear Information System (INIS)
Oyamatsu, Kazuhiro
2001-01-01
The point-reactor kinetics after a step reactivity insertion to a critical condition is solved directly form fission-product (FP) data (fission yields and decay data) for the first time. Numerical calculations are performed with the FP data in ENDF/B-VI. The inhour equation obtained directly from the FP data shows a different behavior at long periods from the one obtained from Tuttle's six-group parameter sets. The behavior is quite similar to the one obtained from the six-group parameter sets in ENDF/B-VI, that were obtained from FP data in a preliminary version of ENDF/B-VI. To identify the erroneous FP data, we examine the asymptotic form of the inhour equation at an infinitely long period. It is found that the most important precursors for long reactor periods are found 137 I, 88 Br and 87 Br. They cover more than 60% of the reactivity. It is remarkable that 137 I alone covers 30-50% depending on the fissioning system. In addition to the three precursors, 136 Te is found a candidate precursor for the peculiarity from the time dependence of the delayed neutron activity. It is recommended that the precision of their Pn values should be improved experimentally. For 137 I, 88 Br, and 87 Br, the relative uncertainty, dPn/Pn, should be decreased down to 2% and for 136 Te to 5%. (author)
Determination of uranium in urine: Comparison of ICP-mass spectrometry and delayed neutron assay
International Nuclear Information System (INIS)
Gladney, E.S.; Moss, W.D.; Gautier, M.A.; Bell, M.G.
1986-01-01
Los Alamos analytical chemistry group acquired a VG-Plasmaquad ICP-MS in January, 1986 and have applied the technique to a variety of environmental and bioassay analytical problems. The authors report on their experience with the determination of uranium and its isotopics in urine and compare this new method with their current uranium procedure, delayed neutron activation analysis (DNA) at the Los Alamos Omega West Reactor. The authors have utilized DNA for bioassay samples since 1978. They currently analyze approximately 2000 urine samples annually. Quantitative data on uranium concentrations are obtained by concurrent measurement of urine standards of known uranium content and isotopic ratio. Detection of 0.03 μg of normal U in a 25 mL sample (1 μg/L) can be achieved by the DNA system. The NRC has proposed new urine bioassay standards that might require at least an order of magnitude reduction in the authors current DNA detection limits. The authors have fully optimized the reactor, and can forsee no instrumental improvement. They may be forced to resort to time-consuming chemical separations at greatly increased costs. DNA is a mature technology with little prospect for radical change. ICPMS is still in its infancy, and there are several ideas for obtaining drastic improvements in detection limits. Costs and time per analysis for both methods are equal
Man/machine interface algorithm for advanced delayed-neutron signal characterization system
International Nuclear Information System (INIS)
Gross, K.C.
1985-01-01
The present failed-element rupture detector (FERD) at Experimental Breeder Reactor II (EBR-II) consists of a single bank of delayed-neutron (DN) detectors at a fixed transit time from the core. Plans are currently under way to upgrade the FERD in 1986 and provide advanced DN signal characterization capability that is embodied in an equivalent-recoil-area (ERA) meter. The new configuration will make available to the operator a wealth of quantitative diagnostic information related to the condition and dynamic evolution of a fuel breach. The diagnostic parameters will include a continuous reading of the ERA value for the breach; the transit time, T/sub tr/, for DN emitters traveling from the core to the FERD; and the isotopic holdup time, T/sub h/, for the source. To enhance the processing, interpretation, and display of these parameters to the reactor operator, a man/machine interface (MMI) algorithm has been developed to run in the background on EBR-II's data acquisition system (DAS). The purpose of this paper is to describe the features and implementation of this newly developed MMI algorithm
TASK, 1-D Multigroup Diffusion or Transport Theory Reactor Kinetics with Delayed Neutron
International Nuclear Information System (INIS)
Buhl, A.R.; Hermann, O.W.; Hinton, R.J.; Dodds, H.L. Jr.; Robinson, J.C.; Lillie, R.A.
1975-01-01
1 - Description of problem or function: TASK solves the one-dimensional multigroup form of the reactor kinetics equations, using either transport or diffusion theory and allowing an arbitrary number of delayed neutron groups. The program can also be used to solve standard static problems efficiently such as eigenvalue problems, distributed source problems, and boundary source problems. Convergence problems associated with sources in highly multiplicative media are circumvented, and such problems are readily calculable. 2 - Method of solution: TASK employs a combination scattering and transfer matrix method to eliminate certain difficulties that arise in classical finite difference approximations. As such, within-group (inner) iterations are eliminated and solution convergence is independent of spatial mesh size. The time variable is removed by Laplace transformation. (A later version will permit direct time solutions.) The code can be run either in an outer iteration mode or in closed (non-iterative) form. The running mode is dictated by the number of groups times the number of angles, consistent with available storage. 3 - Restrictions on the complexity of the problem: The principal restrictions are available storage and computation time. Since the code is flexibly-dimensioned and has an outer iteration option there are no internal restrictions on group structure, quadrature, and number of ordinates. The flexible-dimensioning scheme allows optional use of core storage. The generalized cylindrical geometry option is not complete in Version I of the code. The feedback options and omega-mode search options are not included in Version I
Differential and integral characteristics of prompt fission neutrons in the statistical theory
International Nuclear Information System (INIS)
Gerasimenko, B.F.; Rubchenya, V.A.
1989-01-01
Hauser-Feshbach statistical theory is the most consistent approach to the calculation of both spectra and prompt fission neutrons characteristics. On the basis of this approach a statistical model for calculation of differential prompt fission neutrons characteristics of low energy fission has been proposed and improved in order to take into account the anisotropy effects arising at prompt fission neutrons emission from fragments. 37 refs, 6 figs
Intercomparison of delayed neutron summation calculations among JEF2.2, ENDF/B-VI and JNDC-V2
Energy Technology Data Exchange (ETDEWEB)
Sagisaka, Mitsuyuki [Nagoya Univ. (Japan); Oyamatsu, K.; Kukita, Y.
1998-03-01
We perform intercomparison of delayed neutron activities calculated with JEF2.2, ENDF/B-VI and JNDC-V2 with a simple new method. Significant differences are found at t < 20 (s) for major fissioning systems. The differences are found to stem from fission yields or decay data of several nuclides. The list of these nuclides are also given for the future experimental determination of these nuclear data. (author)
International Nuclear Information System (INIS)
Dillmann, Iris; Dimitriou, Paraskevi; Singh, Balraj
2014-03-01
A summary is given of the 1st Research Coordination Meeting of the new IAEA Coordinated Research Project (CRP) on Development of a Reference Database for Beta-delayed neutron emission data. Participants presented their work, reviewed the current status of the field with regards to individual precursors and aggregate data, and discussed the scope of the work to be undertaken. A list of priorities and task assignments was produced. (author)
International Nuclear Information System (INIS)
El-Mongy, S.A.
2000-01-01
This paper introduces, describes and initiates a very sensitive and rapid non-destructive technique to be used for analysis of the safeguarded nuclear materials 235 U and 239 Pu. The technique is based on fission of the nuclear material by neutrons and then measuring the delayed neutrons produced from the neutron rich fission products. By this technique, fissile isotope content ( 235 U) can be determined in the presence of the other fissile (e.g. 239 Pu) or fertile isotopes (e.g. 238 U) in fresh and spent fuel. The time consumed for analysis of bulk materials by this technique is only 4 minutes. The method is also used for analysis of uranium in rock, sediment, soil, meteorites, lunar, biological, urine, archaeological, zircon sand and seawater samples. The method enables uranium in a sample to be measured without respect to its oxidation state, organic and inorganic elements
International Nuclear Information System (INIS)
Ansarifar, G.R.; Nasrabadi, M.N.; Hassanvand, R.
2016-01-01
Highlights: • We present a S.M.C. system based on the S.M.O for control of a fast reactor power. • A S.M.O has been developed to estimate the density of delayed neutron precursor. • The stability analysis has been given by means Lyapunov approach. • The control system is guaranteed to be stable within a large range. • The comparison between S.M.C. and the conventional PID controller has been done. - Abstract: In this paper, a nonlinear controller using sliding mode method which is a robust nonlinear controller is designed to control a fast nuclear reactor. The reactor core is simulated based on the point kinetics equations and one delayed neutron group. Considering the limitations of the delayed neutron precursor density measurement, a sliding mode observer is designed to estimate it and finally a sliding mode control based on the sliding mode observer is presented. The stability analysis is given by means Lyapunov approach, thus the control system is guaranteed to be stable within a large range. Sliding Mode Control (SMC) is one of the robust and nonlinear methods which have several advantages such as robustness against matched external disturbances and parameter uncertainties. The employed method is easy to implement in practical applications and moreover, the sliding mode control exhibits the desired dynamic properties during the entire output-tracking process independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness and stability.
Recovery characteristics of neutron-irradiated V-Ti alloys
International Nuclear Information System (INIS)
Leguey, T.; Pareja, R.
2000-01-01
The recovery characteristics of neutron-irradiated pure V and V-Ti alloys with 1.0 and 4.5 at.% Ti have been investigated by positron annihilation spectroscopy. Microvoid formation during irradiation at 320 K is produced in pure V and V-1Ti but not in V-4.5Ti. The results are consistent with a model of swelling inhibition induced by vacancy trapping by solute Ti during irradiation. The temperature dependencies of the parameter S in the range 8-300 K indicate a large dislocation bias for vacancies and solute Ti. This dislocation bias prevents the microvoid nucleation in V-4.5Ti, and the microvoid growth in V-1Ti, when vacancies become mobile during post-irradiation annealing treatments. A characteristic increase of the positron lifetime is found during recovery induced by isochronal annealing. It is attributed to a vacancy accumulation into the lattice of Ti oxides precipitated during cooling down, or at their matrix/precipitate interfaces. These precipitates could be produced by the decomposition of metastable phases of Ti oxides formed during post-irradiation annealing above 1000 K
International Nuclear Information System (INIS)
Kodeli, I.; Diop, C.M.; Nimal, J.C.
1994-01-01
In the framework of future Boron Neutron Capture Therapy (BNCT) experiments, where cells and animals irradiations are planned at the research reactor of Strasbourg University, the feasibility to obtain a suitable epithermal neutron beam is investigated. The neutron fluence and spectra calculations in the reactor are performed using the 3D Monte Carlo code TRIPOLI-3 and the 2D SN code TWODANT. The preliminary analysis of Al 2 O 3 and Al-Al 2 O 3 filters configurations are carried out in an attempt to optimize the flux characteristics in the beam tube facility. 7 figs., 7 refs
Characteristics of thermal neutron calibration fields using a graphite pile
International Nuclear Information System (INIS)
Uchita, Yoshiaki; Saegusa, Jun; Kajimoto, Yoichi; Tanimura, Yoshihiko; Shimizu, Shigeru; Yoshizawa, Michio
2005-03-01
The Facility of Radiation Standards of Japan Atomic Energy Research Institute is equipped with thermal neutron fields for calibrating area and personal neutron dosemeters. The fields use moderated neutrons leaked from a graphite pile in which radionuclide sources are placed. In January 2003, we have renewed the pile with some modifications in its size. In accordance with the renewal, we measured and calculated thermal neutron fluence rates, neutron energy distributions and angular distributions of the fields. The thermal neutron fluence rates of the ''inside-pile fields'' and the outside-pile fields'' were determined by the gold foil activation method. The neutron energy distributions of the outside-pile fields were also measured with the Bonner multi-sphere spectrometer system. The contributions of epithermal and fast neutrons to the total dose-equivalents were 9% in the southern outside-pile field and 12% in the western outside-pile field. The personal dose-equivalents, H p,slab (10, α), in the outside-pile fields are evaluated by considering the calculated angular distributions of incoming neutrons. The H p,slab (10, α) was found to be about 40% higher than the value in assuming the unidirectional neutron between the pile and the test point. (author)
International Nuclear Information System (INIS)
Flip, A.; Pang, H.F.; D'Angelo, A.
1995-01-01
Due to the persistent uncertainties: ∼ 5 % (the uncertainty, here and there after, is at 1σ) in the prediction of the 'reactivity scale' (β eff ) for a fast power reactor, an international project was recently initiated in the framework of the OECD/NEA activities for reevaluation, new measurements and integral benchmarking of delayed neutron (DN) data and related kinetic parameters (principally β eff ). Considering that the major part of this uncertainty is due to uncertainties in the DN yields (v d ) and the difficulty for further improvement of the precision in differential (e.g. Keepin's method) measurements, an international cooperative strategy was adopted aiming at extracting and consistently interpreting information from both differential (nuclear) and integral (in reactor) measurements. The main problem arises from the integral side; thus the idea was to realize β eff like measurements (both deterministic and noise) in 'clean' assemblies. The 'clean' calculational context permitted the authors to develop a theory allowing to link explicitly this integral experimental level with the differential one, via a unified 'Master Model' which relates v d and measurables quantities (on both levels) linearly. The combined error analysis is consequently largely simplified and the final uncertainty drastically reduced (theoretically, by a factor √3). On the other hand the same theoretical development leading to the 'Master Model', also resulted in a structured scheme of approximations of the general (stochastic) Boltzmann equation allowing a consistent analysis of the large range of measurements concerned (stochastic, dynamic, static ... ). This paper is focused on the main results of this theoretical development and its application to the analysis of the Preliminary results of the BERENICE program (β eff measurements in MASURCA, the first assembly in CADARACHE-FRANCE)
Pebble bed modular reactor fuel enrichment discrimination using delayed neutrons - HTR2008-58133
International Nuclear Information System (INIS)
Skoda, R.; Rataj, J.; Uhera, J.
2008-01-01
The Pebble Bed Modular Reactor (PBMR) is a helium-cooled, graphite-moderated high temperature nuclear power reactor which utilise fuel in form of spheres that are randomly loaded and continuously circulated through the core until they reach their prescribed end-of-life burn-up limit. When the reactor is started up for the first time, the lower-enriched start-up fuel is used, mixed with graphite spheres, to bring the core to criticality. As the core criticality is established and the start-up fuel is burned-in, the graphite spheres are progressively removed and replaced with more start-up fuel. Once it becomes necessary for maintaining power output, the higher enriched equilibrium fuel is introduced to the reactor and the start-up fuel is removed. During the initial run of the reactor it is important to discriminate between the irradiated startup fuel and the irradiated equilibrium fuel to ensure that only the equilibrium fuel is returned to the reactor. There is therefore a need for an on-line enrichment discrimination device that can discriminate between irradiated start-up fuel spheres and irradiated equilibrium fuel spheres. The device must also not be confused by the presence of any remaining graphite spheres. Due to it's on-line nature the device must accomplish the discrimination within tight time limits. Theoretical calculations and experiments show that Fuel Enrichment Discrimination based on delayed neutrons detection is possible. The paper presents calculations and experiments showing viability of the method. (authors)
Directory of Open Access Journals (Sweden)
Sushenok E.O.
2018-01-01
Full Text Available The neutron emission of the β-decay of 74;76;78;80Ni are studied within the quasiparticle random phase approximation with the Skyrme interaction. The coupling between one- and two-phonon terms in the wave functions of the low-energy 1+ states of the daughter nuclei is taken into account. It is shown that the strength decrease of the neutronproton tensor interaction leads to the increase of the half-life and the neutron-emission probability.
Nuclear characteristics of epoxy resin as a space environment neutron shielding
Energy Technology Data Exchange (ETDEWEB)
Adeli, Ruhollah [Nuclear Science and Technology Research Institute, Yazd (Iran, Islamic Republic of). Central Iran Research Complex; Shirmardi, Seyed Pezhman [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of). Radiation Application Research School; Mazinani, Saideh [Amirkabir Nanotechnology Research Institute, Tehran (Iran, Islamic Republic of); Ahmadi, Seyed Javad [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of). Nuclear Fuel Cycle Research School
2017-03-15
In recent years many investigations have been done for choosing applicable light neutron shielding in space environmental applications. In this study, we have considered the neutron radiation-protective characteristics of neat epoxy resin, a thermoplastic polymer material and have compared it with various candidate materials in neutron radiation protection such as Al 6061 alloy and Polyethylene. The aim of this investigation is the effect of type of moderator for fast neutron, notwithstanding neutron absorbers fillers. The nuclear interactions and the effective dose at shields have been studied with the Monte Carlo N-Particle transport code (MCNP), using variance reductions to reduce the relative error. Among the candidates, polymer matrix showed a better performance in attenuating fast neutrons and caused a lower neutron and secondary photon effective dose.
How the RRR neutronic characteristics impact on the mechanical design
International Nuclear Information System (INIS)
Villarino, E.; Coquibus, K.
2005-01-01
This paper describes how the neutronic characteristics of a very demanding research reactor facility impact on the mechanical design of the reactor core. The Replacement Research Reactor (RRR) for the Australian Nuclear Science and Technology Organization is described making emphasis in the mechanical solutions to improve the core performance. The compact core is located inside a chimney, surrounded by heavy water contained in the Reflector Vessel. The whole assembly is at the bottom of the Reactor Pool, which is full of de-mineralized light water acting as coolant and moderator and biological shielding. The core is an array of sixteen plate-type Fuel Assemblies (FAs) and five absorber plates, which are called Control Plates (CP). The coolant is light water, which flows upwards. The final design of the core layout and control guide boxes was adopted to minimize the flux and PPF perturbation during the normal operation. The four lateral control plates are used mainly to shutdown the reactor and to compensate large reactivity transients. The central and cruciform regulating plate is used to compensate the reactivity change during the cycle operation. The regulating plate does minimize perturbation on PPF and irradiation fluxes. The design of the reflector tank fulfills all the flux requirements for the irradiation facilities and also the flux perturbation between irradiation facilities. (authors)
Influence of core model parameters on the characteristics of neutron beams of the research reactor
Directory of Open Access Journals (Sweden)
N. A. Khafizova
2013-12-01
Full Text Available IRT MEPhI reactor is equipped with a number of facilities at horizontal experimental channels (HEC. Knowing of parameters influencing spatio-angular distribution of irradiation fields is essential for each application area. The research for neutron capture therapy (NCT facility at HEC of the reactor was made. Calculation methods have been used to estimate how the reactor core parameters influence neutron beam characteristics at the HEC output. The impact of neutron source model in Monte Carlo calculations by MCNP code on the parameters of neutron and secondary photon field at the output of irradiation beam tubes of research reactor is estimated. The study shows that specifying neutron source with fission reaction rate distribution in SDEF option gives almost the same results as criticality calculation considered the most accurate. Our calculations show that changes of the core operational parameters have insignificant influence on characteristics of neutron beams at HEC output.
Study on the neutron dosimetric characteristics of Tissue Equivalent Proportional Counter
Energy Technology Data Exchange (ETDEWEB)
Nunomiya, T.; Kim, E.; Kurosawa, T.; Taniguchi, S.; Nakamura, T. [Tohoku Univ., Sendai (Japan). Cyclotron and Radioisotope Center; Tsujimura, N.; Momose, T.; Shinohara, K. [Japan Nuclear Cycle Development Inst., Environment and Safety Division, Tokai Works, Tokai, Ibaraki (Japan)
1999-03-01
The neutron dosimetric characteristics of TEPC (Tissue Equivalent Proportional Counter) has been investigated under a cooperative study between Tohoku University and JNC since 1997. This TEPC is a spherical, large volume, single-wire proportional counter (the model LETSW-5, manufactured by Far West Technology, Inc.) and filled with a tissue equivalent gas in a spherical detector of the A-150 tissue equivalent plastic. The TEPC can measure the spectra of absorbed dose in LET and easily estimate the tissue equivalent dose to neutron. This report summarizes the dosimetric characteristics of TEPC to the monoenergetic neutrons with energy from 8 keV to 15 MeV. It is found that TEPC can estimate the ambient dose equivalent, H*(10), with an accuracy from 0.9 to 2 to the neutron above 0.25 MeV and TEPC has a good counting efficiency enough to measure neutron doses with low dose rate at the stray neutron fields. (author)
Characteristic analysis on moderating material for obtaining epithermal neutron beam
International Nuclear Information System (INIS)
Jiang Xinbiao; Chen Da; Zhang Ying
2000-01-01
The one dimension discrete coordinates transport code ANISN was used to calculate three-group constants of 11 elements which could be used to consist moderating epithermal neutron material of beam. Moderating character of simple substances, compounds and mixtures consisted of the optimized elements analyzed three kinds of moderating materials were optimized for epithermal neutron beam
Refractometry characteristics of α-quartz after neutron irradiation
International Nuclear Information System (INIS)
Abdkadyrova, I.Kh.
1997-01-01
Lattice structure distortions in irradiated crystalline quartz were studied by refractometry methods. The refractometry constants of α-quartz for the flux of fast neutrons 10 18 - 10 21 neutron/cm 2 were calculated. The critical kinetics of this constants at the phase transformation is observed.(author). 5 refs., 1 fig
Calculated characteristics of subcritical assembly with anisotropic transport of neutrons
International Nuclear Information System (INIS)
Gorin, N.V.; Lipilina, E.N.; Lyutov, V.D.; Saukov, A.I.
2003-01-01
There was considered possibility of creating enough sub-critical system that multiply neutron fluence from a primary source by many orders. For assemblies with high neutron tie between parts, it is impossible. That is why there was developed a construction consisting of many units (cascades) having weak feedback with preceding cascades. The feedback attenuation was obtained placing layers of slow neutron absorber and moderators between the cascades of fission material. Anisotropy of fast neutron transport through the layers was used. The system consisted of many identical cascades aligning one by another. Each cascade consists of layers of moderator, fissile material and absorber of slow neutrons. The calculations were carried out using the code MCNP.4a with nuclear data library ENDF/B5. In this construction neutrons spread predominantly in one direction multiplying in each next fissile layer, and they attenuate considerably in the opposite direction. In a calculated construction, multiplication factor of one cascade is about 1.5 and multiplication factor of whole construction composed of n cascades is 1.5 n . Calculated keff value is 0.9 for one cascade and does not exceed 0.98 for a system containing any number of cascades. Therefore the assembly is always sub-critical and therefore it is safe in respect of criticality. There was considered using such a sub-critical assembly to create a powerful neutron fluence for neutron boron-capturing therapy. The system merits and demerits were discussed. (authors)
Characteristics of the WNR: a pulsed spallation neutron source
International Nuclear Information System (INIS)
Russell, G.J.; Lisowski, P.W.; Howe, S.D.; King, N.S.P.; Meier, M.M.
1982-01-01
The Weapons Neutron Research facility (WNR) is a pulsed spallation neutron source in operation at the Los Alamos National Laboratory. The WNR uses part of the 800-MeV proton beam from the Clinton P. Anderson Meson Physics Facility accelerator. By choosing different target and moderator configurations and varying the proton pulse structure, the WNR can provide a white neutron source spanning the energy range from a few MeV to 800 MeV. The neutron spectrum from a bare target has been measured and is compared with predictions using an Intranuclear Cascade model coupled to a Monte Carlo transport code. Calculations and measurements of the neutronics of WNR target-moderator assemblies are presented
A two-solar-mass neutron star measured using Shapiro delay
Demorest, P.B.; Pennucci, T.; Ransom, S.M.; Roberts, M.S.E.; Hessels, J.W.T.
2010-01-01
Neutron stars are composed of the densest form of matter known to exist in our Universe, the composition and properties of which are still theoretically uncertain. Measurements of the masses or radii of these objects can strongly constrain the neutron star matter equation of state and rule out
International Nuclear Information System (INIS)
Van Rooijen, W. F. G.; Lathouwers, D.
2007-01-01
In advanced Generation IV (fast) reactors an integral fuel cycle is envisaged, where all Heavy Metal is recycled in the reactor. This leads to a nuclear fuel with a considerable content of Minor Actinides. For many of these isotopes the nuclear data is not very well known. In this paper the sensitivity of the kinetic behaviour of the reactor to the dynamic parameters λ k , β k and the delayed spectrum χ d,k is studied using first order perturbation theory. In the current study, feedback due to Doppler and/or thermohydraulic effects are not treated. The theoretical framework is applied to a Generation IV Gas Cooled Fast Reactor. The results indicate that the first-order approach is satisfactory for small variations of the data. Sensitivities to delayed neutron data are similar for increasing and decreasing transients. Sensitivities generally increase with reactivity for increasing transients. For decreasing transients, there are less clearly defined trends, although the sensitivity to the delayed neutron spectrum decreases with larger sub-criticality, as expected. For this research, an adjoint capable version of the time-dependent diffusion code DALTON is under development. (authors)
Characteristics of Pyrolytic Graphite as a Neutron Monochromator
International Nuclear Information System (INIS)
Adib, M.; Habib, N.; El-Mesiry, M.S.; Fathallah, M.
2011-01-01
Pyrolytic graphite (PG) has become nearly indispensable in neutron spectroscopy. Since the integrated reflectivity of the monochromatic neutrons from PG crystals cut along its c-axis is high within a wavelength band from 0.1 nm up to .65 nm. The monochromatic features of PG crystal is detailed in terms of the optimum mosaic spread, crystal thickness and reactor moderating temperature for efficient integrated neutron reflectivity within the wavelength band. A computer code Mono-PG has been developed to carry out the required calculations for the PG hexagonal close-packed structure. Calculation shows that, 2 mm thick of PG crystal having 0.30 FWHM on mosaic spread are the optimum parameters of PG crystal as a monochromator at selected neutron wavelength shorter than 2 nm. However, the integrated neutron intensity of 2nd and 3rd orders from thermal reactor flux is even higher than that of the 1st order one at neutron wavelengths longer than 2 nm. While, from cold reactor flux, integrated neutron intensity of the 1st order within the wavelength band from 0.25 up to 0.5 nm is higher than the 2nd and 3rd ones
Characteristics of rotating target neutron source and its use in radiation effects studies
International Nuclear Information System (INIS)
Van Konynenburg, R.A.; Barschall, H.H.; Booth, R.; Wong, C.
1975-07-01
The Rotating Target Neutron Source (RTNS) at Lawrence Livermore Laboratory is currently the most intense source of DT fusion neutrons available for the study of radiation effects in materials. This paper will present a brief description of the machine, outline the history of its development and discuss its performance characteristics and its application to CTR materials research. (U.S.)
Characteristics of Fabricated SiC Neutron Detectors for Neutron Flux Monitoring
Energy Technology Data Exchange (ETDEWEB)
Kim, Han Soo; Ha, Jang Ho; Park, Se Hwan; Lee, Kyu Hong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Cheol Ho [Hanyang University, Seoul (Korea, Republic of)
2011-05-15
An SPND (Self-powered Neutron Detector) is commonly used for neutron detection in NPP (Nuclear Power Plant) by virtue of un-reactivity for gamma-rays. But it has a drawback, which is that it cannot detect neutrons in real time due to beta emissions (about > 48 s) after reactions between neutrons and {sup 103}Rh in an SPND. And Generation IV reactors such as MSR (Molten-salt reactor), SFR (Sodium-cooled fast reactor), and GFR (Gas-cooled fast reactor) are designed to compact size and integration type. For GEN IV reactor, neutron monitor also must be compact-sized to apply such reactor easily and much more reliable. The wide band-gap semiconductors such as SiC, AlN, and diamond make them an attractive alternative in applications in harsh environments by virtue of the lower operating voltage, faster charge-collection times compared with gas-filled detectors, and compact size.1) In this study, two PIN-type SiC semiconductor neutron detectors, which are for fast neutron detection by elastic and inelastic scattering SiC atoms and for thermal neutron detection by charged particle emissions of 6LiF reaction, were designed and fabricated for NPP-related applications. Preliminary tests such as I-V and alpha response were performed and neutron responses at ENF in HANARO research reactor were also addressed. The application feasibility of the fabricated SiC neutron detector as an in-core neutron monitor was discussed
International Nuclear Information System (INIS)
Blanc, Pauline; Tobin, Stephen J.; Croft, Stephen; Menlove, Howard O.; Swinhoe, M.; Lee, T.
2010-01-01
The Next Generation Safeguards Initiative (NGSI) of the U.S. Department of Energy (DOE) has funded multiple laboratories and universities to develop a means to accurately quantify the Plutonium (Pu) mass in spent nuclear fuel assemblies and ways to also detect potential diversion of fuel pins. Delayed Neutron (DN) counting provides a signature somewhat more sensitive to 235 U than Pu while Differential Die-Away (DDA) is complementary in that it has greater sensitivity to Pu. The two methods can, with care, be combined into a single instrument which also provides passive neutron information. Individually the techniques cannot robustly quantify the Pu content but coupled together the information content in the signatures enables Pu quantification separate to the total fissile content. The challenge of merging DN and DDA, prompt neutron (PN) signal, capabilities in the same design is the focus of this paper. Other possibilities also suggest themselves, such as a direct measurement of the reactivity (multiplication) by either the boost in signal obtained during the active interrogation itself or by the extension of the die-away profile. In an early study, conceptual designs have been modeled using a neutron detector comprising fission chambers or 3He proportional counters and a ∼14 MeV neutron Deuterium-Tritium (DT) generator as the interrogation source. Modeling was performed using the radiation transport code Monte Carlo N-Particles eXtended (MCNPX). Building on this foundation, the present paper quantifies the capability of a new design using an array of 3 He detectors together with fission chambers to optimize both DN and PN detections and active characterization, respectively. This new design was created in order to minimize fission in 238 U (a nuisance DN emitter), to use a realistic neutron generator, to reduce the cost and to achieve near spatial interrogation and detection of the DN and PN, important for detection of diversion, all within the constraints of
Calculations of accelerator-based neutron sources characteristics
International Nuclear Information System (INIS)
Tertytchnyi, R.G.; Shorin, V.S.
2000-01-01
Accelerator-based quasi-monoenergetic neutron sources (T(p,n), D(d;n), T(d;n) and Li (p,n)-reactions) are widely used in experiments on measuring the interaction cross-sections of fast neutrons with nuclei. The present work represents the code for calculation of the yields and spectra of neutrons generated in (p, n)- and ( d; n)-reactions on some targets of light nuclei (D, T; 7 Li). The peculiarities of the stopping processes of charged particles (with incident energy up to 15 MeV) in multilayer and multicomponent targets are taken into account. The code version is made in terms of the 'SOURCE,' a subroutine for the well-known MCNP code. Some calculation results for the most popular accelerator- based neutron sources are given. (authors)
Neutron characteristic and spectroscopy logging methods and apparatus
International Nuclear Information System (INIS)
Antkiw, S.
1977-01-01
Earth formations surrounding a well bore are irradiated with pulses of fast neutrons, and gamma rays resulting from the ensuring thermal neutron capture interactions with nuclei of the formations are detected, from which measurements of the thermal neutron decay times characterizing the respective formations are derived. The gamma ray energy spectra of the respective formations are analyzed. Gating of the gamma ray detection periods is automatically controlled, both for the decay time and the spectroscopy functions, in accrdance with the measured values of the decay times. The duration and repetition rate of the neutron pulses are also controlled as a function of the measured decay times to provide an overall optimized decay time-spectroscopy operating cycle. spectroscopy outputs representative of formation lithology, salinity, porosity and shaliness are developed to supplement and improve decay time log interpretation
International Nuclear Information System (INIS)
Agazzi, A.; Gavazzi, C.; Vincenti, E.; Monterosso, R.
1964-01-01
1 - Nature of physical problem solved: The programme studies the spatial dynamics of reactor TESI, in the two group and one space dimension approximation. Only one group of delayed neutrons is considered. The programme simulates the vertical movement of the control rods according to any given movement law. The programme calculates the evolution of the fluxes and temperature and precursor concentration in space and time during the power excursion. 2 - Restrictions on the complexity of the problem: The maximum number of lattice points is 100
Generalized Aharonov-Bohm and wheeler-type delayed choice experiments with neutrons
International Nuclear Information System (INIS)
Zeilinger, A.
1984-01-01
Novel time-dependent neutron interferometry experiments are proposed. These would elucidate the peculiar role of potential energy in quantum mechanics on the one hand and the complementarity in quantum interference on the other hand
Atlantic Richfield Hanford Company californium multiplier/delayed neutron counter safety analysis
International Nuclear Information System (INIS)
Zimmer, W.H.
1976-08-01
The Californium Multiplier (CFX) is a subcritical assembly of uranium surrounding 252 Cf spontaneously fissioning neutron sources; its function is to multiply the neutron flux to a level useful for activation analysis. This document summarizes the safety analysis aspects of the CFX, DNC, pneumatic transfer system, and instrumentation and to detail all the aspects of the total facility as a starting point for the ARHCO Safety Analysis Review. Recognized hazards and steps already taken to neutralize them are itemized
Neutron radiation characteristics of the IVth generation reactor spent fuel
Bedenko, Sergey; Shamanin, Igor; Grachev, Victor; Knyshev, Vladimir; Ukrainets, Olesya; Zorkin, Andrey
2018-03-01
Exploitation of nuclear power plants as well as construction of new generation reactors lead to great accumulation of spent fuel in interim storage facilities at nuclear power plants, and in spent fuel «wet» and «dry» long-term storages. Consequently, handling the fuel needs more attention. The paper is focused on the creation of an efficient computational model used for developing the procedures and regulations of spent nuclear fuel handling in nuclear fuel cycle of the new generation reactor. A Thorium High-temperature Gas-Cooled Reactor Unit (HGTRU, Russia) was used as an object for numerical research. Fuel isotopic composition of HGTRU was calculated using the verified code of the MCU-5 program. The analysis of alpha emitters and neutron radiation sources was made. The neutron yield resulting from (α,n)-reactions and at spontaneous fission was calculated. In this work it has been shown that contribution of (α,n)-neutrons is insignificant in case of such (Th,Pu)-fuel composition and HGTRU operation mode, and integral neutron yield can be approximated by the Watt spectral function. Spectral and standardized neutron distributions were achieved by approximation of the list of high-precision nuclear data. The distribution functions were prepared in group and continuous form for further use in calculations according to MNCP, MCU, and SCALE.
Design characteristics of a three-component AEOI Neutriran Albedo Neutron Personnel Dosimeter
International Nuclear Information System (INIS)
Sohrabi, M.; Katouzi, M.
1991-01-01
In establishing a national personnel neutron dosimetry service in Iran, different parameters of the AEOI Neutriran Albedo Neutron Personnel Dosimeter (NANPD) have been optimized. A NANPD was designed with three dosimetry components to measure (a) direct thermal neutrons, (b) direct fast neutrons and (C) direct neutrons by the detection of the albedo neutrons reflected from the body. The dosimeter consists of one or more Lexan polycarbonate and/or CR-39 foils and two 10 B (n,α) 7 Li converters in a cadmium cover so arranged as to efficiently measure the three neutron dose components separately. The boron converter thickness, its position relative to the beam direction and its distance from the PC foil were studied and the results were incorporated into the design. The dose response of the dosimeter, its lower detection limit as well as the correction factors related to the field neutrons and albedo neutrons were also determined for a 238 Pu-Be, an 241 Am-Be and a 252 Cf sources. In this paper, the dosimeter design and its dosimetric characteristics are presented and discussed. (author)
Energy Technology Data Exchange (ETDEWEB)
Acosta, G. [INFN Laboratori Nazionali di Legnaro, 35020 Legnaro (Italy); Andre, T. [GANIL, Caen (France); Bermudez, J. [INFN Laboratori Nazionali di Legnaro, 35020 Legnaro (Italy); Universidad Simon Bolivar, Caracas (Venezuela, Bolivarian Republic of); Blinov, M.F. [Budker Institute of Nuclear Physics, 630090 Novosibirsk (Russian Federation); Jamet, C. [GANIL, Caen (France); Logatchev, P.V.; Semenov, Y.I.; Starostenko, A.A. [Budker Institute of Nuclear Physics, 630090 Novosibirsk (Russian Federation); Tecchio, L.B., E-mail: tecchio@lnl.infn.it [INFN Laboratori Nazionali di Legnaro, 35020 Legnaro (Italy); Tsyganov, A.S. [Budker Institute of Nuclear Physics, 630090 Novosibirsk (Russian Federation); Udup, E. [INFN Laboratori Nazionali di Legnaro, 35020 Legnaro (Italy); Horia Hulubei National Institute of Physics and Engineering, Bucharest (Romania); University Polytechnic of Bucharest (Romania); Vasquez, J. [INFN Laboratori Nazionali di Legnaro, 35020 Legnaro (Italy); Universidad Simon Bolivar, Caracas (Venezuela, Bolivarian Republic of)
2014-09-11
Research and development of a safety system for the SPIRAL2 facility has been conceived to protect the UCx target from a possible interaction with the 200 kW deuteron beam. The system called “delay window” (DW) is designed as an integral part of the neutron converter module and is located in between the neutron converter and the fission target. The device has been designed as a barrier, located directly behind the neutron converter on the axis of the deuteron beam, with the purpose of “delaying” the eventual interaction of the deuteron beam with the UCx target in case of a failure of the neutron converter. The “delay” must be long enough to allow the interlock to react and safely stop the beam operation, before the beam will reach the UCx target. The working concept of the DW is based on the principle of the electrical fuse. Electrically insulated wires placed on the surface of a Tantalum disk assure a so called “free contact”, normally closed to an electronic circuit located on the HV platform, far from the radioactive environment. The melting temperature of the wires is much less than Tantalum. Once the beam is impinging on the disk, one or more wires are melted and the “free contact” is open. A solid state relay is changing its state and a signal is sent to the interlock device. A prototype of the DW has been constructed and tested with an electron beam of power density equivalent to the SPIRAL2 beam. The measured “delay” is 682.5 ms (σ=116 ms), that is rather long in comparison to the intrinsic delays introduced by the detectors itself (2 ms) and by the associated electronic devices (120 ns). The experimental results confirm that, in the case of a failure of the neutron converter, the DW as conceived is enable to withstand the beam power for a period of time sufficiently long to safely shut down the SPIRAL2 accelerator.
Characteristics of a delay-line readout in a cylindrical drift chamber system
International Nuclear Information System (INIS)
Barber, R.; Ahmed, M.W.; Dzemidzic, M.; Empl, A.; Hungerford, E.V.; Lan, K.J.; Wilson, J.; Cooper, M.D.; Gagliardi, C.A.; Haim, D.; Kim, G.J.; Koetke, D.D.; Tribble, R.E.; Van Ausdeln, L.A.
2002-01-01
This paper reports on the design, construction, and operational characteristics of a delay-line readout implemented on the cathode foils of a cylindrical drift chamber system. The readout was used to determine the position of an event along the length of the 1.74 m drift wires in the MEGA detectors used at the Los Alamos Meson Physics Facility. The performance of the system is interpreted by comparison to a PSPICE simulation, and to simple analytical models
Applications and characteristics of imaging plates as detector in neutron radiography at SINQ
Kolbe, H; Gunia, W; Körner, S
1999-01-01
Imaging plate technique is a commonly accepted method in many fields as in medicine, biology and physics for detection of the distribution of beta- and gamma-radiation or X-rays on large areas. Recently a new type of imaging plate sensitive to neutrons has been developed. The storage layer is doped with gadolinium, which, after absorption of neutrons, produces radiation detectable by the same sensitive crystals used in conventional imaging plates. At the spallation neutron source, SINQ, at the Paul Scherrer Institut (CH) some of the characteristics of the neutron radiography station in combination with the imaging plate technique were investigated. The intensity distribution of the source was measured to check the accuracy for quantification of the image data. Also, the reproducibility of results obtained by this detection system was stated. For a test object, the high selectivity for different neutron absorption is demonstrated at details with low contrast. The obtainable spatial resolution was determined re...
Energy Technology Data Exchange (ETDEWEB)
Perlini, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires
1961-07-01
This paper describes a theoretical and experimental study on the detection of neutrons present in the primary cooling circuit of a reactor cooled by heavy or light water, with a view to the installation of a canning burst detection unit. The concentration of background neutrons is first calculated, taking into account the neutrons from nitrogen 17 decay, and the photoneutrons produced by the decay of nitrogen 16 and sodium 24. The emission of delayed fission neutrons, originating at a given crack in the canning, has been estimated. Using the D{sub 2}O circuit of the pile EL-3, three units have been developed by means of which the following three types of detector may be compared: 1) BF{sub 3} proportional counter 2) Boron scintillator 3) Fission chamber Under the present experimental conditions the BF{sub 3} counter gave the best results. The influence on these detectors of the {gamma} flux, which in certain cases reaches 200 R/h, is analysed. Finally a calibration is carried out with an experimental crack of 30 mm{sup 2} of uranium exposed to a flux of 5.8 x 10{sup 13} n.cm{sup -2}.s{sup -1}. The sensitivity obtained with the BF{sub 3} counter during this test is 2 counts/s per mm{sup 2} of exposed uranium. (author) [French] Le present rapport est une etude theorique et experimentale sur la detection des neutrons presents dans le circuit primaire de refroidissement d'un reacteur refrigere par l'eau lourde ou l'eau legere, en vue d'une installation de detection de ruptures de gaines. On fait d'abord un calcul sur la concentration des neutrons de bruit de fond en tenant compte: des neutrons de decroissance de l'azote 17 et des photoneutrons produits par les decroissances de l'azote 16 et du sodium 24. L'emission des neutrons differes de fission, qui ont pour origine une fissure de gaine donnee, a ete evaluee. Utilisant le circuit D{sub 2}O de la pile EL3, trois installations ont ete mises au point permettant de comparer les trois types de detecteurs suivants: 1
Energy Technology Data Exchange (ETDEWEB)
Ryu, Eun Hyun; Song, Yong Mann; Park, Joo Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2013-05-15
If time-dependent equation is solved with the FEM, the limitation of the input geometry will disappear. It has often been pointed out that the numerical methods implemented in the RFSP code are not state-of-the-art. Although an acceleration method such as the Coarse Mesh Finite Difference (CMFD) for Finite Difference Method (FDM) does not exist for the FEM, one should keep in mind that the number of time steps for the transient simulation is not large. The rigorous formulation in this study will richen the theoretical basis of the FEM and lead to an extension of the dynamics code to deal with a more complicated problem. In this study, the formulation for the 1-D, 1-G Time Dependent Neutron Diffusion Equation (TDNDE) without consideration of the delay neutron will first be done. A problem including one multiplying medium will be solved. Also several conclusions from a comparison between the numerical and analytic solutions, a comparison between solutions with various element orders, and a comparison between solutions with different time differencing will be made to be certain about the formulation and FEM solution. By investigating various cases with different values of albedo, theta, and the order of elements, it can be concluded that the finite element solution is agree well with the analytic solution. The higher the element order used, the higher the accuracy improvements are obtained.
International Nuclear Information System (INIS)
Taran, Yu.V.
1985-01-01
A new method has been developed for measuring the polarization characteristics of ferromagnetic films on ultracold neutrons (UCN) by single-, double- and triple-transmission of UCN beam through one and the same film. To realize the method an installation has been proposed consisting of the two UCN storage traps connected with a mirror neutron guide. An investigated film is placed in the slit in the middle of the neutron guide. On both sides of the film a spin-flipper is installed bottle is equiped with three neutron values which permit filling in the bottle with UCN and allow oneto let UCN out to the neutron guide or detector. The neutrons once let out from one bottle into the neutron guide are caught by the other. The film can be moved out of the neutron guide or rotated. By manipulating with spin-flippers and the film one may take the integral polarization parameters of the film: transmission, polarizing and analysing efficiencies, so-called S-factor, which is the fourth independent linear combination of the elements of the square 2x2 transmission matrix of the film. The measurement parameters help to restore the film transmission matrix. Then a comparison is drawn with the theoretical models of UCN depolarization on transmission through a ferromagnetic film
Self-powered in-core neutron detector assembly with uniform perturbation characteristics
International Nuclear Information System (INIS)
Todt, W.H.; Playfoot, K.C.
1979-01-01
Disclosed is a self-powered in-core neutron detector assembly in which a plurality of longitudinally extending self-powered detectors have neutron responsive active portions spaced along a longitudinal path. A low neutron absorptive extension extends from the active portions of the spaced active portions of the detectors in symmetrical longitudinal relationship with the spaced active detector portions of each succeeding detector. The detector extension terminates with the detector assembly to provide a uniform perturbation characteristic over the entire assembly length
International Nuclear Information System (INIS)
Poskas, P.; Bartkus, G.
1999-01-01
The overview of the activities in the Division of Thermo hydro-mechanics related with the assessment of thermal, neutronic and radiation characteristics in spent nuclear fuel and radwaste facilities are performed. Also some new data about radiation characteristics of the RBMK-1500 spent nuclear fuel are presented. (author)
High energy resolution characteristics on 14MeV neutron spectrometer for fusion experimental reactor
Energy Technology Data Exchange (ETDEWEB)
Iguchi, Tetsuo [Tokyo Univ., Tokai, Ibaraki (Japan). Nuclear Engineering Research Lab.; Takada, Eiji; Nakazawa, Masaharu
1996-10-01
A 14MeV neutron spectrometer suitable for an ITER-like fusion experimental reactor is now under development on the basis of a recoil proton counter telescope principle in oblique scattering geometry. To verify its high energy resolution characteristics, preliminary experiments are made for a prototypical detector system. The comparison results show reasonably good agreement and demonstrate the possibility of energy resolution of 2.5% in full width at half maximum for 14MeV neutron spectrometry. (author)
Investigation and analysis of neutron emission characteristics in Denaplasma focus facility
International Nuclear Information System (INIS)
Goodarzi, Sh.; Amrollahi, R.; Babazadeh, A.; Nasiri, A.
2003-01-01
Since the first experiments with plasma focus facilities in 1960' s. These devices are known as intense sources of neutron when the working gas contains deuterium with a proper density. Most of the emitted neutrons are produced by D-D reactions, but the mechanism of these reactions in not still clear completely. In this paper, the results of experimental investigations of neutron emission characteristics in D ena p lasma focus facility (Filipov type, 90 kJ, 25 kV) over a range of discharge voltages and pressures are presented. Out working gases are D 2 and D 2+%1 Kr, two different conic and flat insert anodes were employed. We have simultaneously measured the total emission in our experiments for analyzing the neutron generation mechanism in this device. We have found the upper and lower pressure limits and the optimum pressure for neutron generation, and we have observed the double pluses structure of neutron signal for the first time in this device. Form the experimental results, it seems that both thermonuclear and no thermonuclear mechanisms are always present in neutron generation, but their contribution in the total yield is strongly dependent on experimental conditions (initial pressure, discharge voltage, gas admixture, etc.). It was found that the range of variation of total neutron yield and neutron emission anisotropy factor for experiments with D + %1 Kr is wider than experiments with D 2, and the best neutron emission results belongs to discharges in D 2 + %1 Kr with a conic insert anode. By employing D 2 + %1 Kr with a conic insert anode, and varying pressure between 0.3-2 torr at a discharge voltage of 16 kV, it can be deduced that in low pressures ( n ∝ I α ρ ∝E α / 2 was found about 3.62 for D 2 + %1 Kr and 3 for D 2
International Nuclear Information System (INIS)
Diniz, Ricardo
2005-01-01
A reactor noise approach has been employed at the IPEN/MB-01 research reactor facility in order to determine experimentally the effective delayed neutron parameters β i and λ i in a six group model and assuming the point reactor. The method can be considered a novice one because exploits the very low frequency domain of the spectral densities. The proposed method has some advantages to other in-pile methods since it does not disturb the reactor system and consequently does not 'excite' any sort of harmonic modes. As a byproduct and a consistency check, the β eff parameter was obtained without the need of the Diven factor and the power normalization and it is in excellent agreement with independent measurements. The theory/experiment comparison shows that for the abundances the JENDL 3.3 presents the best performance while for the decay constants the revised version of ENDF/B-VI.8 shows the best agreement. The best performance for the β eff determination is obtained with JENDL3.3. In contrast, ENDF/B-VI.8 and its revised version performed at LANL overestimate β eff by as much as 4%. The β eff results of this work support totally the proposal of reducing the thermal delayed neutron number for 235 U fission as made by Sakurai and Okajima. A new observed effect related to the correlation between the fluctuations of both measurement channels is also presented and discussed. This effect can be considered as an indirect evidence for the use of the point reactor model in this work as well as a possible useful tool in the understanding of reactor dynamics. (author)
International Nuclear Information System (INIS)
Kawai, Takeshi; Yoshino, Hiroshi; Kawabata, Yuji; Hino, Masahiro
2000-01-01
Shapes of moderator baths of ORPHEE and NIST without bottom of inner cylinder, entering liquid from downward and push down the liquid by steam formed nuclear exotherm to fill inner part of the inner cylinder with steam, require to determine a number of parameters to be optimum to realize a state storing steam in inner cylinder and liquid in shell portion. Then, for a modulator bath with a structure shielding the inner cylinder from shell portion by preparing bottom without any pore and supplying steam into the cylinder through a steam return pipe mounted with pores at its upper portion. By such structure, a cold neutron source with self-balance-ability and capable of following output without time delaying. And, its liquid volume can also be controlled by system pressure. And that, as its structure is simple, it has another characteristic that its connection structure of transmission pipe portion with moderator bath portion. (G.K.)
Directory of Open Access Journals (Sweden)
Minato Futoshi
2016-01-01
Full Text Available Nuclear β-decay and delayed neutron (DN emission is important for the r-process nucleosynthesis after the freeze-out, and stable and safe operation of nuclear reactors. Even though radioactive beam facilities have enabled us to measure β-decay and branching ratio of neutron-rich nuclei apart from the stability line in the nuclear chart, there are still a lot of nuclei which one cannot investigate experimentally. In particular, information on DN is rather scarce than that of T1/2. To predict T1/2 and the branching ratios of DN for next JENDL decay data, we have developed a method which comprises the quasiparticle-random-phase-approximation (QRPA and the Hauser-Feshbach statistical model (HFSM. In this work, we calculate fission fragments with T1/2 ≤ 50 sec. We obtain the rms deviation from experimental half-life of 3:71. Although the result is still worse than GT2 which has been adopted in JENDL decay data, DN spectra are newly calculated. We also discuss further subjects to be done in future for improving the present approach and making next generation of JENDL decay data.
Beta-delayed proton emission in neutron-deficient lanthanide isotopes
International Nuclear Information System (INIS)
Wilmarth, P.A.
1988-01-01
Forty-two β-delayed proton precursors with 56≤Z≤71 and 63≤N≤83 were produced in heavy-ion reactions at the Lawrence Berkeley Laboratory SuperHILAC and their radioactive decay properties studied at the on-line mass separation facility OASIS. Twenty-five isotopes and eight delayed proton branches were identified for the first time. Delayed proton energy spectra and proton coincident γ-ray and x-ray spectra were measured for all precursors. In a few cases, proton branching ratios were also determined. The precursor mass numbers were determined by the separator, while the proton coincident x-ray energies provided unambiguous Z identifications. The proton coincident γ-ray intensities were used to extract final state branching ratios. Proton emission from ground and isomeric states was observed in many cases. The majority of the delayed proton spectra exhibited the smooth bell-shaped distribution expected for heavy mass precursors. The experimental results were compared to statistical model calculations using standard parameter sets. Calculations using Nilsson model/RPA β-strength functions were found to reproduce the spectral shapes and branching ratios better than calculations using either constant or gross theory β-strength functions. Precursor half-life predictions from the Nilsson model/RPA β-strength functions were also in better agreement with the measured half-lives than were gross theory predictions. The ratios of positron coincident proton intensities to total proton intensities were used to determine Q/sub EC/-B/sub p/ values for several precursors near N=82. The statistical model calculations were not able to reproduce the experimental results for N=81 precursors. 154 refs., 82 figs., 19 tabs
Footprint Characteristics of Cosmic-Ray Neutron Sensors for Soil Moisture Monitoring
Schrön, Martin; Köhli, Markus; Zreda, Marek; Dietrich, Peter; Zacharias, Steffen
2015-04-01
Cosmic-ray neutron sensing is a unique and an increasingly accepted method to monitor the effective soil water content at the field scale. The technology is famous for its low maintenance, non-invasiveness, continuous measurement, and most importantly, for its large footprint. Being more representative than point data and finer resolved than remote-sensing products, cosmic-ray neutron derived soil moisture products provide unrivaled advantage for mesoscale hydrologic and land surface models. The method takes advantage of neutrons induced by cosmic radiation which are extraordinarily sensitive to hydrogen and behave like a hot gas. Information about nearby water sources are quickly mixed in a domain of tens of hectares in air. Since experimental determination of the actual spatial extent is hardly possible, scientists have applied numerical models to address the footprint characteristics. We have revisited previous neutron transport simulations and present a modified conceptual design and refined physical assumptions. Our revised study reveals new insights into probing distance and water sensitivity of detected neutrons under various environmental conditions. These results sharpen the range of interpretation concerning the spatial extent of integral soil moisture products derived from cosmic-ray neutron counts. Our findings will have important impact on calibration strategies, on scales for data assimilation and on the interpolation of soil moisture data derived from mobile cosmic-ray neutron surveys.
International Nuclear Information System (INIS)
Mills, R.W.
1990-07-01
A review of fission product yields and delayed neutron data for Np-237, Pu-242, Am-242m, Am-243, Cm-243 and Cm-245 has been undertaken. Gaps in understanding and inconsistencies in existing data were identified and priority areas for further experimental, theoretical and evaluation investigation detailed
Neutron-neutron probe for uranium exploration
International Nuclear Information System (INIS)
Smith, R.C.
1979-01-01
A neutron activation probe for assaying the amount of fissionable isotopes in an ore body is described which comprises a casing which is movable through a borehole in the ore body, a neutron source and a number of delayed neutron detectors arranged colinearly in the casing below the neutron source for detecting delayed neutrons
Characteristics of old neutron stars in dense interstellar clouds
International Nuclear Information System (INIS)
Boehringer, H.; Morfill, G.E.; Zimmermann, H.U.
1987-01-01
The forms observable radiation will assume as old neutron stars pass through interstellar clouds and accrete material are examined theoretically. The radiation, mainly X-rays and gamma rays, will be partially absorbed by the surrounding dust and gas, which in turn produces far-IR radiation from warm dust and line radiation from the gas. Adiabatic compression of the accretion flow and the accretion shock are expected to produce cosmic rays, while gamma rays will be emitted by interaction of the energetic particles with the cloud material. The calculations indicate that the stars will then be identified as X-ray sources, some of which may be unidentified sources in the COS-B database. 37 references
DEFF Research Database (Denmark)
Leutscher, Peter; Madsen, Gitte; Erlandsen, Mogens
2012-01-01
Background: Delays in the diagnosis and treatment of tuberculosis (TB) are commonly encountered. Methods: A study was undertaken among pulmonary tuberculosis (PTB) and extrapulmonary tuberculosis (EPTB) patients in a Danish university hospital to describe demographic and clinical characteristics...
Austin, David; Yan, Andrew T; Spratt, James C; Kunadian, Vijay; Edwards, Richard J; Egred, Mohaned; Bagnall, Alan J
2014-09-01
Delayed arrival to a primary percutaneous coronary intervention (PPCI)-capable hospital following ST-elevation myocardial infarction (STEMI) is associated with poorer outcome. The influence of patient characteristics on delayed presentation during STEMI is unknown. This was a retrospective observational study. Patients presenting for PPCI from March 2008 to November 2011 in the north of England (Northumbria, Tyne and Wear) were included. The outcomes were self-presentation to a non-PPCI-capable hospital, symptom to first medical contact (STFMC) time, total ischaemic time and mortality during follow-up. STEMI patients included numbered 2297; 619 (26.9%) patients self-presented to a non-PPCI-capable hospital. STFMC of >30 min and total ischaemic time of >180 min was present in 1521 (70.7%) and 999 (44.9%) cases, respectively. Self-presentation was the strongest predictor of prolonged total ischaemic time (odds ratio, OR (95% confidence interval, CI): 5.05 (3.99-6.39)). Married patients (OR 1.38 (1.10-1.74)) and patients living closest to an Emergency Room self-presented more commonly (driving time (vs. ≤10 min) 11-20 min OR 0.66 (0.52-0.83), >20 minutes OR 0.46 (0.33-0.64). Unmarried females waited longest to call for help (OR vs. married males 1.89 (1.29-2.78) and experienced longer total ischaemic times (OR 1.51 (1.10-2.07)). Married patients had a borderline association with lower mortality (hazard ratio 0.75 (0.53-1.05), p=0.09). Unmarried female patients had the longest treatment delays. Married patients and those living closer to an Emergency Room self-present more frequently. Early and exclusive use of the ambulance service may reduce treatment delay and improve STEMI outcome. © The European Society of Cardiology 2014.
Caballero-Folch, R; Cortès, G; Taín, J L; Agramunt, J; Algora, A; Ameil, F; Ayyad, Y; Benlliure, J; Bowry, M; Calviño, F; Cano-Ott, D; Davinson, T; Dillmann, I; Estrade, A; Evdokimov, A; Faestermann, T; Farinon, F; Galaviz, D; García-Ríos, A; Geissel, H; Gelletly, W; Gernhäuser, R; Gómez-Hornillos, M B; Guerrero, C; Heil, M; Hinke, C; Knöbel, R; Kojouharov, I; Kurcewicz, J; Kurz, N; Litvinov, Y; Maier, L; Marganiec, J; Marta, M; Martínez, T; Montes, F; Mukha, I; Napoli, D R; Nociforo, C; Paradela, C; Pietri, S; Podolyák, Zs; Prochazka, A; Rice, S; Riego, A; Rubio, B; Schaffner, H; Scheidenberger, C; Smith, K; Sokol, E; Steiger, K; Sun, B; Takechi, M; Testov, D; Weick, H; Wilson, E; Winfield, J S; Wood, R; Woods, P J; Yeremin, A
2014-01-01
New measurements of very exotic nuclei in the neutron-rich region beyond N=126 have been performed at the GSI facility with the fragment separator (FRS). The aim of the experiment is to determine half-lives and beta-delayed neutron emission branching ratios of isotopes of Hg, Tl and Pb in this region. This contribution summarizes final counting statistics for identification and for implantation, as well as the present status of the data analysis of the half-lives. In summary, isotopes of Pt, Au, Hg, Ti, Pb, Bi, Po, At, Rn and Fr were clearly identified and several of them (Hg208-211, Tl211-215, Pb214-218) were implanted with enough statistics to determine their half-lives. About half of them are expected to be neutron emitters, in such cases it will become possible to obtain the neutron emission probabilities, P-n.
International Nuclear Information System (INIS)
Volkov, N.G.; Kryanev, A.V.
1984-01-01
Technique for obtaining estimations of neutron-physical characteristics of nuclides on the basis of stable estimation methods is set forth. The technique presupposes correction of incorrectly determined errors of measurements and disclosure of systematic errors with their succeeding accountancy. A system of orthogonal polynomials is used as approximating functional dependence. The technique is also generalized at the presence of correlation between measurements
Isomer-delayed gamma-ray spectroscopy of neutron-rich 166Tb
Directory of Open Access Journals (Sweden)
Gurgi L.A.
2017-01-01
Full Text Available This short paper presents the identification of a metastable, isomeric-state decay in the neutron-rich odd-odd, prolate-deformed nucleus 166Tb. The nucleus of interest was formed using the in-flight fission of a 345 MeV per nucleon 238U primary beam at the RIBF facility, RIKEN, Japan. Gamma-ray transitions decaying from the observed isomeric states in 166Tb were identified using the EURICA gamma-ray spectrometer, positioned at the final focus of the BigRIPS fragments separator. The current work identifies a single discrete gamma-ray transition of energy 119 keV which de-excites an isomeric state in 166Tb with a measured half-life of 3.5(4 μs. The multipolarity assignment for this transition is an electric dipole and is made on the basis internal conversion and decay lifetime arguments. Possible two quasi-particle Nilsson configurations for the initial and final states which are linked by this transition in 166Tb are made on the basis of comparison with Blocked BCS Nilsson calculations, with the predicted ground state configuration for this nucleus arising from the coupling of the v(1-/2[521] and π(3+/2 Nilsson orbitals.
Characteristics of the NE-213 large-volume neutron counters for muon catalyzed fusion investigation
International Nuclear Information System (INIS)
Bystritsky, V.M.; Wozniak, J.; Zinov, V.G.
1984-01-01
The Monte-Carlo method was used to establish the properties and feasibility of a large-volume NE-213 scin illator as an efficient neutron detector. The recoil proton spectra, calculated efficiencies for different detection thresholds and scintillator sizes are presented for the neutron energy up to 15 MeV. The time characteristics, e. g., time resolution, are discussed. It is also shown that no strong influence of light attenuation by the scintilla or itself on calculated efficiencies is observed, when gamma-calibration technique is used. The detector vol me of approximately 100 l is suggested for application in investigations of μ-atom and μ-molecular processes
Energy Technology Data Exchange (ETDEWEB)
Nicol, T. [CEA, DEN, Cadarache, Nuclear Measurement Laboratory, F-13108 Saint-Paul-lez-Durance (France); FZJ, Institute of Energy and Climate Research – Nuclear Waste Management and Reactor Safety, Wilhelm-Johnen-Straße, d-52425 Jülich (Germany); Pérot, B., E-mail: bertrand.perot@cea.fr [CEA, DEN, Cadarache, Nuclear Measurement Laboratory, F-13108 Saint-Paul-lez-Durance (France); Carasco, C. [CEA, DEN, Cadarache, Nuclear Measurement Laboratory, F-13108 Saint-Paul-lez-Durance (France); Brackx, E. [CEA, DEN, Marcoule, Metallography and Chemical Analysis Laboratory, F-30207 Bagnols-sur-Cèze (France); Mariani, A.; Passard, C. [CEA, DEN, Cadarache, Nuclear Measurement Laboratory, F-13108 Saint-Paul-lez-Durance (France); Mauerhofer, E. [FZJ, Institute of Energy and Climate Research – Nuclear Waste Management and Reactor Safety, Wilhelm-Johnen-Straße, d-52425 Jülich (Germany); Collot, J. [Laboratoire de Physique Subatomique et de Cosmologie, Université Grenoble Alpes, CNRS/IN2P3 Grenoble (France)
2016-10-01
This paper reports a feasibility study of {sup 235}U and {sup 239}Pu characterization in 225 L bituminized waste drums or 200 L concrete waste drums, by detecting delayed fission gamma rays between the pulses of a deuterium-tritium neutron generator. The delayed gamma yields were first measured with bare samples of {sup 235}U and {sup 239}Pu in REGAIN, a facility dedicated to the assay of 118 L waste drums by Prompt Gamma Neutron Activation Analysis (PGNAA) at CEA Cadarache, France. Detectability in the waste drums is then assessed using the MCNPX model of MEDINA (Multi Element Detection based on Instrumental Neutron Activation), another PGNAA cell dedicated to 200 L drums at FZJ, Germany. For the bituminized waste drum, performances are severely hampered by the high gamma background due to {sup 137}Cs, which requires the use of collimator and shield to avoid electronics saturation, these elements being very penalizing for the detection of the weak delayed gamma signal. However, for lower activity concrete drums, detection limits range from 10 to 290 g of {sup 235}U or {sup 239}Pu, depending on the delayed gamma rays of interest. These detection limits have been determined by using MCNPX to calculate the delayed gamma useful signal, and by measuring the experimental gamma background in MEDINA with a 200 L concrete drum mock-up. The performances could be significantly improved by using a higher interrogating neutron emission and an optimized experimental setup, which would allow characterizing nuclear materials in a wide range of low and medium activity waste packages.
Neutron, gamma ray, and temperature effects on the electrical characteristics of thyristors
Frasca, A. J.; Schwarze, G. E.
1992-01-01
Experimental data showing the effects of neutrons, gamma rays, and temperature on the electrical and switching characteristics of phase-control and inverter-type SCR's are presented. The special test fixture built for mounting, heating, and instrumenting the test devices is described. Four SCR's were neutron irradiated at 300 K and four at 365 K for fluences up to 3.2 x 10 exp 13 pn/sq. cm, and eight were gamma irradiated at 300 K only for gamma doses up to 5.1 Mrads. The electrical measurements were made during irradiation and the switching measurements were made only before and after irradiation. Radiation induced crystal defects, resulting primarily from fast neutrons, caused the reduction of minority carrier lifetime through the generation of R-G centers. The reduction in lifetime caused increases in the on-state voltage drop and in the reverse and forward leakage currents, and decreases in the turn-off time.
Neutron, gamma ray, and temperature effects on the electrical characteristics of thyristors
International Nuclear Information System (INIS)
Schwarze, G.E.; Frasca, A.J.
1992-01-01
In this paper, experimental data showing the effects of neutrons, gamma rays, and temperature on the electrical and switching characteristics of phase-control and inverter-type SCRs are presented. The special test fixture built for mounting, heating, and instrumenting the test devices is described. Four SCRs were neutron irradiated at 300 K and four at 365 K for fluences up to 3.2 x 10 13 n/cm 2 , and eight were gamma irradiated at 300 K only for gamma doses up to 5.1 Mrads. The electrical measurements were made during irradiation and the switching measurements were made only before and after irradiation. Radiation induced crystal defects, resulting primarily from fast neutrons, caused the reduction of minority carrier lifetime through the generation of R-G centers. The reduction in lifetime caused increases in the on-state voltage drop and in the reverse and forward leakage currents, and decreases in the turn-off time
Energy Technology Data Exchange (ETDEWEB)
Giacri-Mauborgne, M.L
2005-11-15
This thesis work consists in two parts. The first part is the description of the creation of a photonuclear activation file which will be used to calculated photonuclear activation. To build this file we have used different data sources: evaluations but also calculations done using several cross sections codes (HMS-ALICE, GNASH, ABLA). This file contains photonuclear activation cross sections for more than 600 nuclides and fission fragments distributions for 30 actinides at tree different Bremsstrahlung energies and the delay neutron spectrum associated. These spectra are not in good agreement with experimental data. That is why we decided to launch measurement of delayed neutrons spectra from photofission. The second part of this thesis consists in demonstrating the possibility to do such measurements at the ELSA accelerator facility. To that purpose, we have developed the detection, the acquisition system and the analysis method of such spectra. These were tested for the measurement of the delayed neutron spectrum of uranium-238 after irradiation in a 2 MeV neutron flux. Finally, we have measured the delayed neutron spectrum of uranium-238 after irradiation in a 15 MeV Bremsstrahlung flux. We compare our results with experimental data. The experiment has allowed us to improve the value of {nu}{sub p}-bar with an absolute uncertainty below 7%, we propose {nu}{sub p}-bar = (3.03 {+-} 0.02) n/100 fissions, and to correct the Nikotin's parameters for the six group representation. Particularly, we have improved the data concerning the sixth group by taking into account results from different irradiation times.
International Nuclear Information System (INIS)
Hill, T. R.; Reed, W. H.
1980-01-01
1 - Description of problem or function: TIMEX solves the time- dependent, one-dimensional multigroup transport equation with delayed neutrons in plane, cylindrical, spherical, and two-angle plane geometries. Both regular and adjoint, inhomogeneous and homogeneous problems subject to vacuum, reflective, periodic, white, albedo or inhomogeneous boundary flux conditions are solved. General anisotropic scattering is allowed and anisotropic inhomogeneous sources are permitted. 2 - Method of solution: The discrete ordinates approximation for the angular variable is used with the diamond (central) difference approximation for the angular extrapolation in curved geometries. A linear discontinuous finite element representation for the angular flux in each spatial mesh cell is used. Negative fluxes are eliminated by a local set-to-zero and correct algorithm. The time variable is differenced by an explicit technique that is unconditionally stable so that arbitrarily large time-steps can be taken. Two acceleration methods, exponential extrapolation and re-balance, are utilized to improve the accuracy of the time differencing scheme. 3 - Restrictions on the complexity of the problem: Variable dimensioning is used so that any combination of problem parameters leading to a container array less than MAXCOR can be accommodated. In addition, the CDC version permits the use of extended core storage less than MAXECS
Thermal-hydraulic and neutron-physical characteristics of a new SCWR fuel assembly
International Nuclear Information System (INIS)
Liu, X.J.; Cheng, X.
2009-01-01
A new fuel assembly design for a thermal supercritical water cooled reactor (SCWR) core is proposed. Compared to the existing fuel assemblies, the present fuel assembly has two-rows of fuel rods between the moderator channels, to achieve a more uniform moderation for all fuel rod cells, and subsequently, a more uniform radial power distribution. In addition, a neutron-kinetics/thermal-hydraulics coupling method is developed, to analyze the neutron-physical and thermal-hydraulic behavior of the fuel assembly designs. This coupling method is based on the sub-channel analysis code COBRA-IV for thermal-hydraulics and the neutron-kinetics code SKETCH-N for neutron-physics. Both the COBRA-IV code and the SKETCH-N code are accordingly modified. An interface is established for the data transfer between these two codes. This coupling method is applied to both the one-row fuel assemblies (previous design) and the two-row fuel assemblies (present design). The performance of the two types of fuel assemblies is compared. The results show clearly that the two-row fuel assembly has more favorable neutron-physical and thermal-hydraulic characteristics than the one-row fuel assembly. The effect of various parameters on the fuel assembly performance is discussed. The coupling method is proven to be well suitable for further applications to SCWR fuel assembly design analysis
Characteristics of SiC neutron sensor spectrum unfolding process based on Bayesian inference
Energy Technology Data Exchange (ETDEWEB)
Cetnar, Jerzy; Krolikowski, Igor [Faculty of Energy and Fuels AGH - University of Science and Technology, Al. Mickiewicza 30, 30-059 Krakow (Poland); Ottaviani, L. [IM2NP, UMR CNRS 7334, Aix-Marseille University, Case 231 -13397 Marseille Cedex 20 (France); Lyoussi, A. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance (France)
2015-07-01
This paper deals with SiC detector signal interpretation in neutron radiation measurements in mixed neutron gamma radiation fields, which is called the detector inverse problem or the spectrum unfolding, and it aims in finding a representation of the primary radiation, based on the measured detector signals. In our novel methodology we resort to Bayesian inference approach. In the developed procedure the resultant spectra is unfolded form detector channels reading, where the estimated neutron fluence in a group structure is obtained with its statistical characteristic comprising of standard deviation and correlation matrix. In the paper we present results of unfolding process for case of D-T neutron source in neutron moderating environment. Discussions of statistical properties of obtained results are presented as well as of the physical meaning of obtained correlation matrix of estimated group fluence. The presented works has been carried out within the I-SMART project, which is part of the KIC InnoEnergy R and D program. (authors)
Energy Technology Data Exchange (ETDEWEB)
Beliard, L; Janot, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires
1967-07-01
A preliminary study was conducted to determine the amount of fissile material present in a sample. The method used consisted in irradiating the sample by means of a pulsed neutron generator and delayed neutron counting. Results show the validity of this method provided some experimental precautions are taken. Checking on the residual proportion of fissile material in leached hulls seems possible. (authors) [French] Ce rapport rend compte d'une etude preliminaire effectuee en vue de determiner la quantite de matiere fissile presente dans un echantillon. La methode utilisee consiste a irradier l'echantillon considere au moyen d'une source puisee de neutrons et a compter les neutrons retardes produits. Les resultats obtenus permettent de conclure a la validite de la methode moyennant certaines precautions. Un controle de la teneur residuelle en matiere fissile des gaines apres traitement semble possible. (auteurs)
International Nuclear Information System (INIS)
Sellers, M.T.; Corcoran, E.C.; Kelly, D.G.
2011-01-01
A delayed neutron counting (DNC) system for the analysis of special nuclear materials (SNM) has been constructed and calibrated at the Royal Military College of Canada. The polyethylene vials used to transport SNM samples have been found to contribute a time-dependent count rate, B(t), far above the system background. B(t) has been found to be independent of polyethylene mass and shows a dependence on irradiation position in the SLOWPOKE-2 reactor and irradiation time. A comparison of B(t) and the theoretical delayed neutron production from the fission of small amounts of 235 U has indicated that trace amounts of uranium may be present in the DNC system tubing. (author)
International Nuclear Information System (INIS)
Orso, J A
2012-01-01
The critical state of a nuclear reactor is an unstable equilibrium. The nuclear reactor can go from critical to subcritical state or can go from critical to hypercritical state. Although the evolution of the system in these cases is slow, it requires the intervention of an operator to correct deviations. For this reason an automatic control technique was designed, based on the kinetic point to a group of delayed neutrons, which corrects deviations automatically. In this paper we study the point kinetics models in a group and six groups of delayed neutrons for different values of reactivity using the simulations software MATLAB, Simulink. A comparison of two models with the reactor kinetic behavior is made (author)
International Nuclear Information System (INIS)
Rachidi, J.
1983-04-01
This work is dedicated to the study of the emission of delayed neutrons observed in the decay of 49 K, 50 K and 51 K. Spectroscopic data are non-existent for these 3 isotopes, so we have had to design a specific detection system based on a large-surface scintillation counter. A series of n-γ coincidence measurement has allowed us to determine the energy levels of the non-bound states of 49 Ca, 50 Ca and 51 Ca and to establish the nature of the beta transitions (K → Ca). We have measured the energy of the delayed neutrons through the time-of-flight method. Our results are consistent with the model of the p-n states based on the Bansac-French's works. This model shows that the non-bound states of the calcium isotopes discovered in the experiment are represented by simple configurations of the (sd) -1 (fp) n type. (A.C.)
Energy Technology Data Exchange (ETDEWEB)
Andola, Sanjay; Niranjan, Ram [Applied Physics Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Kaushik, T.C., E-mail: tckk@barc.gov.in [Applied Physics Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Rout, R.K. [Applied Physics Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Kumar, Ashwani; Paranjape, D.B.; Kumar, Pradeep; Tomar, B.S.; Ramakumar, K.L. [Radioanalytical Chemistry Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Gupta, S.C. [Applied Physics Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)
2014-07-01
A pulsed neutron source based on plasma focus device has been used for active interrogation and assay of {sup 235}U by monitoring its delayed high energy γ-rays. The method involves irradiation of fissile material by thermal neutrons obtained after moderation of a burst of neutrons emitted upon fusion of deuterium in plasma focus (PF) device. The delayed gamma rays emitted from the fissile material as a consequence of induced fission were detected by a large volume sodium iodide (NaI(Tl)) detector. The detector is coupled to a data acquisition system of 2k input size with 2k ADC conversion gain. Counting was carried out in pulse height analysis mode for time integrated counts up to 100 s while the temporal profile of delayed gamma has been obtained by counting in multichannel scaling mode with dwell time of 50 ms. To avoid the effect of passive (natural) and active (from surrounding materials) backgrounds, counts have been acquired for gamma energy between 3 and 10 MeV. The lower limit of detection of {sup 235}U in the oxide samples with this set-up is estimated to be 14 mg.
International Nuclear Information System (INIS)
Unesaki, H.; Isaka, S.; Nakagome, Y.
2006-01-01
Impact of nuclear library difference on neutronic characteristics of thorium-loaded light water reactor fuel is investigated through cell burnup calculations using SRAC code system. Comparison of k ∞ and nuclide composition was made between the results obtained by JENDL-3.3, ENDF/B-VI.8 and JEFF3.0 for (U, Th)O 2 fuels as well as UO 2 fuels, with special interest on the burnup dependence of the neutronic characteristics. The impact of nuclear data library difference on k ∞ of (U, Th)O 2 fuels was found to be significantly large compared to that of UO 2 fuels. Notable difference was also found in nuclide concentration of TRU nuclides. (authors)
Energy Technology Data Exchange (ETDEWEB)
Qiu, Suizheng; Zhang, Dalin; Liu, Minghao; Liu, Limin; Xu, Rongshuan; Gong, Cheng; Su, Guanghui [Xi' an Jiaotong Univ. (China). State Key Laboratory of Multiphase Flow in Power Engineering
2016-05-15
Molten salt reactor (MSR) as one candidate of the Generation IV advanced nuclear power systems is attracted more attention in China due to its top ranked fuel cycle and thorium utilization. The MSRs are characterized by using liquid-fuel, which offers complicated coupling problem of neutronics and thermal hydraulics. In this paper, the fundamental model and numerical method are established to calculate and analyze the safety characteristics for liquid-fuel MSRs. The theories and methodologies are applied to the MOSART concept. The liquid-fuel flow effects on neutronics, reactivity coefficients and three operation parameters' influences at steady state are obtained, which provide the basic information for safety analysis. The unprotected loss of flow transient is calculated, the results of which shows the inherent safety characteristics of MOSART due to its strong negative reactivity feedbacks.
Improvement of radiation response characteristic on CdTe detectors using fast neutron irradiation
International Nuclear Information System (INIS)
Miyamaru, Hiroyuki; Takahashi, Akito; Iida, Toshiyuki
1999-01-01
The treatment of fast neutron pre-irradiation was applied to a CdTe radiation detector in order to improve radiation response characteristic. Electron transport property of the detector was changed by the irradiation effect to suppress pulse amplitude fluctuation in risetime. Spectroscopic performance of the pre-irradiated detector was compared with the original. Additionally, the pre-irradiated detector was employed with a detection system using electrical signal processing of risetime discrimination (RTD). Pulse height spectra of 241 Am, 133 Ba, and 137 Cs gamma rays were measured to examine the change of the detector performance. The experimental results indicated that response characteristic for high-energy photons was improved by the pre-irradiation. The combination of the pre-irradiated detector and the RTD processing was found to provide further enhancement of the energy resolution. Application of fast neutron irradiation effect to the CdTe detector was demonstrated. (author)
Ion beam characteristics of the controlatron/zetatron family of the gas filled neutron tubes
Energy Technology Data Exchange (ETDEWEB)
Berg, R.S.; Shope, L.A.; O' Neal, M.L.; Boers, J.E.; Bickes, R.W. Jr.
1981-03-01
A gas filled tube used to produce a neutron flux with the D(T,He/sup 4/)n reaction is described. Deuterium and tritium ions generated in a reflex discharge are extracted and accelerated to 100 keV by means of an accelerator electrode onto a deutero-tritide target electrode. The electrodes are designed to focus the ion beam onto the target. Total tube currents consisting of extracted ions, unsuppressed secondary electrons, and ions generated by interactions with the background gas are typically 100 mA. The characteristics of the extracted ion beam are discussed. Accelerating voltages greater than 50 kV are required to focus the beam through the accelerator aperture for configurations that give beams with the proper energy density onto the target. The perveance of the beam is discussed. Maximum perveance values are 2 to 20 nanopervs. Tube focusing and neutron production characteristics are described.
International Nuclear Information System (INIS)
Qiu, Suizheng; Zhang, Dalin; Liu, Minghao; Liu, Limin; Xu, Rongshuan; Gong, Cheng; Su, Guanghui
2016-01-01
Molten salt reactor (MSR) as one candidate of the Generation IV advanced nuclear power systems is attracted more attention in China due to its top ranked fuel cycle and thorium utilization. The MSRs are characterized by using liquid-fuel, which offers complicated coupling problem of neutronics and thermal hydraulics. In this paper, the fundamental model and numerical method are established to calculate and analyze the safety characteristics for liquid-fuel MSRs. The theories and methodologies are applied to the MOSART concept. The liquid-fuel flow effects on neutronics, reactivity coefficients and three operation parameters' influences at steady state are obtained, which provide the basic information for safety analysis. The unprotected loss of flow transient is calculated, the results of which shows the inherent safety characteristics of MOSART due to its strong negative reactivity feedbacks.
International Nuclear Information System (INIS)
Hanelt, E.
1992-02-01
In this thesis it was shown that the projectile fragmentation at relativistic projectile velocities is a production mechanism for exotic nuclei, which is because of its advantageous kinematics especially suited for the fast and efficient separation of the reaction product in an ion optical system. An essential result of these studies is that projectile fragments can be separated in a wide energy range from about 100 MeV/nucleon to 1 GeV/nucleon and over the whole mass range by means of a momentum-loss achromate. In the experiment described in this thesis this method was for the first time applied to the measurement of the β-deLayed neutron radioactivity. The studied isotopes - 1 - 4Be, - 1 - 7B, and - 1 - 9C were produced by the fragmentation of a - 2 - 2Ne beam at 60 MeV/nucleon. A measurement of β half-lifes and neutron branching ratios was performed, the accuracy of which was in other experiments with similarly exotic nuclei hitherto hardly reached. In - 1 - 7B thereby for the first time a β-delayed 4-neutron radioactivity could be detected. The results of these measurements were compared with calculations from different theoretical models. The observed multiplicities of the β-delayed neutrons are consistent with the multiplicities, which are expected by means of a comparison of the Q - β values and the neutron binding energies. The measured neutron branching ratios yield indirect information on distribution of the β strength in the daugther nuclei. At time none of the theories is yet able to reproduce these experimental values in sufficient way. (orig./HSI) [de
Energy Technology Data Exchange (ETDEWEB)
Coelho, Paulo Rogerio Pinto
1979-07-01
This work presents results of non destructive mass analysis of natural uranium by the pulsed source technique. Fissioning is produced by irradiating the test sample with pulses of 14 MeV neutrons and the uranium mass is calculated on a relative scale from the measured emission of delayed neutrons. Individual measurements were normalised against the integral counts of a scintillation detector measuring the 14 MeV neutron intensity. Delayed neutrons were measured using a specially constructed slab detector operated in anti synchronism with the fast pulsed source. The 14 MeV neutrons were produced via the T(d,n) {sup 4}He reaction using a 400 kV Van de Graaff accelerated operated at 200 kV in the pulsed source mode. Three types of sample were analysed, namely: discs of metallic uranium, pellets of sintered uranium oxide and plates of uranium aluminium alloy sandwiched between aluminium. These plates simulated those of Material Testing Reactor fuel elements. Results of measurements were reproducible to within an overall error in the range 1.6 to 3.9%; the specific error depending on the shape, size and mass of the sample. (author)
International Nuclear Information System (INIS)
Hill, T.R.; Reed, W.H.
1976-01-01
TIMEX solves the time-dependent, one-dimensional multigroup transport equation with delayed neutrons in plane, cylindrical, spherical, and two-angle plane geometries. Both regular and adjoint, inhomogeneous and homogeneous problems subject to vacuum, reflective, periodic, white, albedo or inhomogeneous boundary flux conditions are solved. General anisotropic scattering is allowed and anisotropic inhomogeneous sources are permitted. The discrete ordinates approximation for the angular variable is used with the diamond (central) difference approximation for the angular extrapolation in curved geometries. A linear discontinuous finite element representation for the angular flux in each spatial mesh cell is used. The time variable is differenced by an explicit technique that is unconditionally stable so that arbitrarily large time steps can be taken. Because no iteration is performed the method is exceptionally fast in terms of computing time per time step. Two acceleration methods, exponential extrapolation and rebalance, are utilized to improve the accuracy of the time differencing scheme. Variable dimensioning is used so that any combination of problem parameters leading to a container array less than MAXCOR can be accommodated. The running time for TIMEX is highly problem-dependent, but varies almost linearly with the total number of unknowns and time steps. Provision is made for creation of standard interface output files for angular fluxes and angle-integrated fluxes. Five interface units (use of interface units is optional), five output units, and two system input/output units are required. A large bulk memory is desirable, but may be replaced by disk, drum, or tape storage. 13 tables, 9 figures
International Nuclear Information System (INIS)
Galantucci, R.; Haas, R.; Janssens, A.; Agostini, P.; Becker, L.; Bernede, M.; Caldon, L.; Cordani, G.; Guardini, S.
1989-01-01
SIGMA is a device designed for the Safeguards verification measurements of the 235 U content of THTR (or AVR) pebble fuel elements; it has been in operation for more than 12 years at the HOBEG fuel fabrication plant in HANAU (FRG). The THTR fuel element is a 60 mm diameter sphere composed of particles of a U-Th mixture in a graphite matrix. The device is schematically composed of an irradiation facility where the pebble is irradiated by 252 Cf and a counting facility where the delayed fission neutrons are monitored. The system has recently been thoroughly updated, the modified instrument being called SIGMA-R. SIGMA-R automatically governs the whole charge-discharge, irradiation and counting sequence, collects the measured counts in the microprocessor memory and performs the relevant calculations, including the statistical evaluation. The instructions are given interactively to the inspector through an IRIS terminal. SIGMA-R provides extensive self diagnosis features. The inspector is also informed explicity of malfunctions during routine operation. The instrument is equipped with a permanent memory which facilitates the inspector's work considerably. SIGMA is complementary to the so-called fuel pebble sampling devices (FPSD) which are designed to take a random sample of 10 pebbles out of 1000 directly from the production line. With SIGMA-R it is now possible to measure the 235 U content of THTR pebbles very accurately. The uncertainty essentially is limited to Poisson statistics and allows us to evaluate the 235 U content in routine operation, with a random error down to 0.15% and a systematic uncertainty estimated at 0.25%, the latter coming from calibration and normalization procedures
International Nuclear Information System (INIS)
Rabinovich, E.; Golan, G.; Axelevich, A.; Inberg, A.; Oksman, M.; Rosenwaks, I.; Lubarsky, G.; Seidman, A.; Croitoru, N.; Rancoita, P.G.; Rattaggi, M.
1999-01-01
Neutron irradiation effects on (p-n) and Schottky-junction silicon detectors were studied. It was shown that neutron interactions with monocrystalline silicon create specific types of microstructure defects with morphology differing according to the level of neutron fluences (Φ). The isolated dislocation loops, formed by interstitial atoms were observed in microstructure images for 10 10 ≤ Φ ≤ 10 12 n/cm 2 . A strong change in the dislocation loops density and a cluster formation was observed for Φ ≥ 10 13 n/cm 2 . A drastic silicon damage was found for fluences over 10 14 n/cm 2 . These fluences created zones enriched with all types of dislocations, covering more than 50 % of the total surface area. A mechanical fragility appeared in that fluence range in a form of microcracks. 10 14 n/cm 2 appears to be a critical value of neutron irradiation because of the radiation damage described above and because the characteristics I f -V f of silicon detectors can be differentiated from those obtained at low fluences. (A.C.)
International Nuclear Information System (INIS)
Huang Chuming; Yin Zhao; Fang Yannan; Hong Weimin; Zeng Xianjie; Wang Tianwen
2008-01-01
Objective: To discuss the clinical and imaging characteristics of delayed encephalopathy after acute carbon monoxide poisoning (DEACMP). Methods: The clinical data of 14 patients with DEACMP were respectively analyzed. Results: Initial symptom included mentally falling (MMSE rate < 20 points) in 14 patients (100%), incontinence of urine in 11 patients (78%). Relatively significant increase of hematoma (HCT) (average level 0.503±0.027) was found in 6 patients. Three kinds of brain magnetic resonance imaging (MRI) appearance including diffuse white matter lesion, cortical lesion and basal nuclei lesion often existed in association. Therapy for improving circulation was given in 14 patients, and therapy with hormone was given in 9 patients; with 85% improvement rate. 5 patients without hormone therapy but with improving circulation therapy also received improvement. Conclusion: High level of HCT might be one of the causes of DECACMP due to acute disturbance of brain micro-circulation.. Close monitoring the change of HCT levels should be adopted for at least 3-6 months during therapy. Therapy for improving circulation and for anti-agglutination of platelets should be performed besides the routine high-tension oxygen therapy. Low FA value on DTI indicates the occurrence of a demyelization change in the brain long tract fibers, which indicated therapeutic effect of hormone treatment. And, therefore it would be better to perform DTI scan before therapy for individualized therapy. (authors)
[Analysis of characteristics and influence factors of diagnostic delay of endometriosis].
Han, X T; Guo, H Y; Kong, D L; Han, J S; Zhang, L F
2018-02-25
Objective: To access the influence factors of diagnostic delay of endometriosis. Methods: We designed a questionnaire of diagnostic delay of endometriosis. From February 2014 to February 2016, 400 patients who had dysmenorrhea and diagnosed with endometriosis by surgery in Peking University Third Hospital were surveyed retrospectively. Time and risk factors of diagnostic delay were analyzed. Results: The diagnostic delay of 400 patients was 13.0 years (0.2-43.0 years), 78.5%(314/400) patients thought pain was a normal phenomenon and didn't see the doctor. Patients who suffered dysmenorrhea at menarche experienced longer diagnostic delay than those who had dysmenorrhea after menarche (18.0 vs 4.5 years; Z= 191.800, Pendometriosis (DIE) , family history of dysmenorrhea or endometriosis, previous surgical history of endometriosis, high stage, with infertility, adenomyoma or other symptoms, could help to shorten diagnostic delay with no significant difference ( P> 0.05) . By multiple logistic regression analysis, the results shown that whether have dysmenorrhea at menarche and clinical diagnosis time were the independent factors affecting delayed diagnosis ( Pendometriosis is common and the mean delay time is 13.0 years mainly due to the unawareness of dysmenorrhea. Dysmenorrhea at menarche, clinical diagnosis time and dysmenorrhea intensity are the factors affecting time of diagnostic delay.
Evaluation of neutronic characteristics of in-pile test reactor for fast reactor safety research
Energy Technology Data Exchange (ETDEWEB)
Uto, N.; Ohno, S.; Kawata, N. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center
1996-09-01
An extensive research program has been carried out at the Power Reactor and Nuclear Fuel Development Corporation for the safety of future liquid-metal fast breeder reactors to be commercialized. A major part of this program is investigation and planning of advanced safety experiments conducted with a new in-pile safety test facility, which is larger and more advanced than any of the currently existing test reactors. Such a transient safety test reactor generally has unique neutronic characteristics that require various studies from the reactor physics point of view. In this paper, the outcome of the neutronics study is highlighted with presenting a reference core design concept and its performance in regard to the safety test objectives. (author)
Development of a CAD-based neutron transport code with the method of characteristics
International Nuclear Information System (INIS)
Chen Zhenping; Wang Dianxi; He Tao; Wang Guozhong; Zheng Huaqing
2012-01-01
The main problem determining whether the method of characteristics (MOC) can be used in complicated and highly heterogeneous geometry is how to combine an effective geometry processing method with MOC. In this study, a new idea making use of MCAM, which is a Mutlti-Calculation Automatic Modeling for Neutronics and Radiation Transport program developed by FDS Team, for geometry description and ray tracing of particle transport was brought forward to solve the geometry problem mentioned above. Based on the theory and approach as the foregoing statement, a two dimensional neutron transport code was developed which had been integrated into VisualBUS, developed by FDS Team. Several benchmarks were used to verify the validity of the code and the numerical results were coincident with the reference values very well, which indicated the accuracy and feasibility of the method and the MOC code. (authors)
Energy Technology Data Exchange (ETDEWEB)
Hummel, David William, E-mail: hummeld@mcmaster.ca; Novog, David Raymond
2016-03-15
Highlights: • A coupled spatial kinetics and thermalhydraulics model of the PT-SCWR was created. • Positive power excursions were demonstrated during accident-like transients. • The reactor will inherently self-shutdown in such transients with some delay. • A fast-acting shutdown system would limit the consequences of the power pulse. - Abstract: The Canadian Supercritical Water-cooled Reactor concept, as an evolution of the CANada Deuterium Uranium (CANDU) reactor, includes both pressure tubes and a low temperature heavy water moderator. The current Pressure Tube type SCWR (PT-SCWR) concept features 64-element fuel assemblies placed within High Efficiency Re-entrant Channels (HERCs) that connect to core inlet and outlet plena. Among current SCWR concepts the PT-SCWR is unique in that the HERC separates multiple coolant and moderator regions, giving rise to coupled neutronic-thermalhydraulic feedbacks beyond those present in CANDU or contemporary Light Water Reactors. The objective of this work was thus to model the coupled neutronic-thermal hydraulic properties of the PT-SCWR to establish the impact of these multiple regions on the core's transient behavior. To that end, the features of the PT-SCWR were first modeled with the neutron transport code DRAGON to create a database of homogenized and condensed cross-sections and thermalhydraulic feedback coefficients. These were used as input to a core-level neutron diffusion model created with the code DONJON. The behavior of the primary heat transport system was modeled with the thermalhydraulic system code CATHENA. A procedure was developed to couple the outputs of DONJON and CATHENA, facilitating three-dimensional spatial neutron kinetics and coupled thermalhydraulic analysis of the PT-SCWR core. Several postulated transients were initiated within the coupled model by changing the core inlet and outlet boundary conditions. Decreasing coolant density around the fuel was demonstrated to produce positive
Neutronic characteristics of linear-assembly breed-and-burn reactors
International Nuclear Information System (INIS)
Petroski, Robert; Forget, Benoit; Forsberg, Charles
2012-01-01
Highlights: ► Simple models used to characterize general behavior of linear-assembly B and B reactors. ► Diffusion theory model developed to explain axial distributions, height vs. reactivity. ► Neutron excess concept reformulated to include linear-assembly B and B reactors. ► Designed model of B and B reactor started using melt-refined B and B reactor used fuel. ► Computed doubling time of fuel cycle requiring no chemical separations. - Abstract: Linear-assembly breed-and-burn (B and B) reactors are B and B reactors that use axially connected assemblies similar to conventional LWR or fast reactor fuel assemblies. Methods for analyzing linear-assembly B and B reactors and their fuel cycles are developed and applied. General neutronic characteristics of linear-assembly B and B reactors are analyzed, including the effects that burnup, shuffling sequence, and radial and axial size have on equilibrium-cycle k-effective. The mechanisms that give rise to a highly peaked axial burnup distribution are explained, and a method for predicting peak burnup vs. k-effective based on infinite-medium depletion calculations is developed. Next, the neutron excess concept from previous studies of B and B reactors is extended to apply to linear-assembly B and B reactors, which allows the amount of starter fuel needed to establish a given equilibrium cycle to be calculated. Several example applications of the neutron excess formulation are given. First, an example model of a linear-assembly B and B reactor is analyzed to find the neutron excess cost of an equilibrium cycle. Second, simple one-dimensional models are used to predict the neutron excess value obtainable from different starter fuel configurations. Finally, these ideas are applied to design a fuel cycle consisting of linear-assembly B and B reactors and fuel recycling via a melt refining process. The neutron excess concept is used to design an appropriate starter fuel configuration made from melt refined fuel, which
Testing the characteristics of a neutron detector array by Monte-Carlo simulations
International Nuclear Information System (INIS)
Timis, C.; Cruceru, I.; Sandu, M.; Borcea, C.; Buta, A.; Negoita, F.; Angelique, J.C.; Martin, T.; Peter, J.; Grevy, S.; Lienard, E.; Orr, N.A.
1998-01-01
The characteristics of the neutron detector array TONNERRE have been determined experimentally via preliminary tests with a 252 Cf source and by means of simulation using a modified version of the Monte-Carlo program of Cecil et al. Of particular interest is the intrinsic detection efficiency. As it is well known, the neutron detection efficiency for one element of the detector array, depends on the threshold for the light collection (bias) expressed in energy electron equivalent. The experimental efficiencies for five neutron energies and for a bias of 80 KeV ee are presented. The efficiencies for three thresholds and neutron energies between 1-10 MeV are simulated. The neutron energy is determined by TOF over a flight path, s, and the relative energy resolution is given as a function of σ s and σ t (the uncertainties in the flight path), s (uniform as a function of depth) and flight time, t. The mean time resolution was 1.13 ns which gives a TOF resolution of 1.48 ns. That gives a relative energy resolution which increases slowly from 2% at E n =1 MeV to 3.5% at 5 MeV. Position resolution along one module is 12 cm. To help boosting the efficiency, the elements can be arranged in two layers, but that complicates the analysis by enhancing the effects of cross-talk and out-scattering. Cross-talk is the familiar problem of one neutron creating signals in two separate detectors. In out-scattering, a neutron scatters from the non-active part of a detector and is then detected in a different detector with incorrect position and TOF. While methods exist for identifying and eliminating cross-talk events, there are no methods available for identifying out-scattered events. For the case of two layers and a bias of 80 KeV ee, simulated efficiency of two superposed elements versus neutron energy, the out-scattering probability and the probability of cross-talk are presented. The out-scattering probability comes mainly from events when neutrons scatter first on carbon nuclei
β-delayed charged particle decays of neutron-deficient nuclei 20Mg and 23Si and 22Si
International Nuclear Information System (INIS)
Babo, Mathieu
2016-01-01
The neutron-deficient nuclei 20 Mg, 23 Si and 22 Si were produced by fragmentation at NSCL, at MSU (USA), and implanted into an array of 3 double sided stripped Si detectors, surrounded by 16 high-purity Ge detectors. This novel arrangement allowed the detection of the charged particles emitted by the unbound excited states in coincidence with the γ rays emitted by the de-excitation of the daughter. The βp decay of 20 Mg is very well-known and therefore was used to test and optimize the analysis program. The β-delayed proton transitions to the first 3 excited states in 19 Ne were identified and compared to previous measurements. The half-life, the branching ratio of the transitions and the excitation energies, including the IAS, were measured and are in good agreement with the adopted values. The study of the β+ decay of 23 Si allowed the identification of 14 excited states in 23 Al. The emission of 2 protons from the IAS was unambiguously identified. The measurement of the IAS energy allowed a better determination of the mass excess of 23 Si, giving 23.27 (7) MeV. A possible β3p decay channel was also tentatively identified. Most of the theoretical predictions are in favor of a 2-proton radioactive 22 Si. The β2p decays to the first excited state and the ground state of 20 Na were identified. The branching ratio of the decay to the IAS is 2.05 (44) %, and the IAS excitation energy was measured to be 9040 (54) keV. The additional measurement of the half-life gives T 1/2 = 30.38 (45) ms, and allowed the determination of the partial half-life. In this study, we propose a parameterization of the statistical rate function f for the superallowed Fermi β decays. This allow the first indirect mass measurement of 22 Si ground state, 31.49 (14) MeV. The two-proton threshold is then S2p = 645 (100) keV and does not allow 2p radioactivity. (author) [fr
International Nuclear Information System (INIS)
Haire, M.J.; Schryver, J.C.
1999-01-01
Apportionment is the assignment of top-level requirements to lower tier elements of the overall facility. A method for apportioning overall facility availability requirements among systems and subsystems is presented. Characteristics that influence equipment reliability and maintainability are discussed. Experts, using engineering judgment, scored each characteristic for each system whose availability design goal is to be established. The Analytic Hierarchy Process (AHP) method is used to produce a set of weighted rankings for each characteristic for each alternative system. A mathematical model is derived which incorporates these weighting factors. The method imposes higher availability requirements on those systems in which an incremental increase in availability is easier to achieve, and lower availability requirements where greater availability is more difficult and costly. An example is given of applying this top-down apportionment methodology to the Spallation Neutron Source (SNS) facility
Neutron irradiation characteristic tests of oxygen sensors using zirconia solid electrolyte
International Nuclear Information System (INIS)
Hiura, Nobuo; Endou, Yasuichi; Yamaura, Takayuki; Niimi, Motoji; Hoshiya, Taiji; Saito, Junichi; Souzawa, Shizuo; Ooka, Norikazu; Kobiyama, Mamoru.
1997-03-01
In the Department of JMTR of Japan Atomic Energy Research Institute (JAERI), the in-situ measuring technique of oxygen potential has been being developed to study the chemical behavior of high burn-up fuel base-irradiated in the Light Water Reactor. In this test for development of the technique, oxygen sensors using zirconia solid electrolyte stabilized by MgO, CaO and Y 2 O 3 , named MSZ, CSZ and YSZ, respectively, were irradiated by neutrons in the Japan Materials Testing Reactor (JMTR) of JAERI and the characteristics of electromotive force of these sensors under and after irradiation were discussed. From the experimental results, the electromotive force of YSZ sample under irradiation decreased with an increase in irradiation fluence within a range of neutron fluence (E>1 MeV) up to 1 x 10 23 m -2 . The electromotive force of MSZ sensor irradiated with neutron fluences (E>1 MeV) up to 9 x 10 21 m -2 was almost equal to the theoretical value of the electromotive force. It was shown that after irradiation, a decrease in the electromotive force of CSZ sensor was smaller than those of MSZ and YSZ sensors, although the electromotive forces of MSZ, CSZ and YSZ sensors were smaller than the theoretical value. (author)
Synergism of the method of characteristics and CAD technology for neutron transport calculation
International Nuclear Information System (INIS)
Chen, Z.; Wang, D.; He, T.; Wang, G.; Zheng, H.
2013-01-01
The method of characteristics (MOC) is a very popular methodology in neutron transport calculation and numerical simulation in recent decades for its unique advantages. One of the key problems determining whether the MOC can be applied in complicated and highly heterogeneous geometry is how to combine an effective geometry processing method with MOC. Most of the existing MOC codes describe the geometry by lines and arcs with extensive input data, such as circles, ellipses, regular polygons and combination of them. Thus they have difficulty in geometry modeling, background meshing and ray tracing for complicated geometry domains. In this study, a new idea making use of a CAD solid modeler MCAM which is a CAD/Image-based Automatic Modeling Program for Neutronics and Radiation Transport developed by FDS Team in China was introduced for geometry modeling and ray tracing of particle transport to remove these geometrical limitations mentioned above. The diamond-difference scheme was applied to MOC to reduce the spatial discretization error of the flat flux approximation in theory. Based on MCAM and MOC, a new MOC code was developed and integrated into SuperMC system, which is a Super Multi-function Computational system for neutronics and radiation simulation. The numerical testing results demonstrated the feasibility and effectiveness of the new idea for geometry treatment in SuperMC. (authors)
International Nuclear Information System (INIS)
Glushkov, E.S.; Grebennik, V.N.; Davidenko, V.G.; Kosovskij, V.G.; Smirnov, O.N.; Tsibul'skij, V.F.
1991-01-01
The paper is dedicated to the consideration of uncertainties is neutron-physical characteristics (NPC) of high-temperature gas-cooled reactors (HTGR) with a core as spherical fuel element bed, which are caused by calculations from HTGR parameters mean values affecting NPC. Among NPC are: effective multiplication factor, burnup depth, reactivity effect, control element worth, distribution of neutrons and heat release over a reactor core, etc. The short description of calculated methods and codes used for HTGR calculations in the USSR is given and evaluations of NPC uncertainties of the methodical character are presented. Besides, the analysis of the effect technological deviations in parameters of reactor main elements such as uranium amount in the spherical fuel element, number of neutron-absorbing impurities in the reactor core and reflector, etc, upon the NPC is carried out. Results of some experimental studies of NPC of critical assemblies with graphite moderator are given as applied to HTGR. The comparison of calculations results and experiments on critical assemblies has made it possible to evaluate uncertainties of calculated description of HTGR NPC. (author). 8 refs, 8 figs, 6 tabs
Influence of production technology and design on characteristics neutron-sensitive P-I-N diodes
International Nuclear Information System (INIS)
Perevertaylo, V.L.; Kovrygin, V.I.
2012-01-01
This paper presents the results of tests on neutron-sensitive p-i-n diode with local p-n junction, which allows to measure not only the integral dose by nonionizing energy loss (NIEL), but also the real-time dose and dose rate because of ionizing energy losses (IEL). The influence of design and process parameters and the lifetime of minority carriers on the radiation characteristics of the device considered. Sensitivity at low doses (from one to ten rad) is limited due to a decrease in the lifetime because of influence of lateral sides of cut. The sensitivity and accuracy of dose can be increased by moving of p-n junction away from the cut surface. The dependence of the voltage drop across the diode on the neutron dose irradiation up to 5 krad received, and the sensitivity was 2 - 3 mV/rad. We have demonstrated that replacement of the bulk p-i-n diode with total p-n junction by new diodes with local p-n junction allow for increase sensitivity, accuracy of dose and application in NIEL and IEL measurements simultaneously. Explanation for the extinction of a direct current through the diode with increasing doses of neutron irradiation proposed
International Nuclear Information System (INIS)
Kawai, Takeshi; Utsuro, Masahiko; Ogino, Fumimaru.
1991-01-01
The present report describes that a cold neutron source (CNS) having a closed-thermosiphon cooling loop shows a self-regulating characteristic under thermal disturbances if the effect of the moderator transfer tube is negligible. Due to this property, the liquid level in the moderator cell is kept almost constant under thermal disturbances. The thermodynamic meaning of the self-regulating property in the idealized closed-thermosiphon and the effect of the moderator transfer tube to the self-regulation are described. (author)
International Nuclear Information System (INIS)
Perry, R.T.; Gorenflo, H.; Daenner, W.
1976-01-01
INDRA is a program system for calculating the neutronics and photonics characteristics of fusion reactor blankets. It incorporates a total of 19 different codes and 5 large data libraries. 10 of the codes are available from the code distribution organizations. Some of them, however, have been slightly modified in order to permit a convenient transfer of information from one program module to the next. The remaining 9 programs have been prepared by the authors to complete the system with respect to flexibility and to facilitate the handling of the results. (orig./WBU) [de
International Nuclear Information System (INIS)
Yanagisawa, Hiroshi; Sono, Hiroki
2003-06-01
For the examination of neutronic safety design of forthcoming experimental core configurations in the Static Experiment Critical Facility (STACY), neutronic characteristics of 80-cm-diameter cylindrical cores fueled with 6% enriched uranyl nitrate solution have been evaluated by computational analyses. In the analyses, the latest nuclear data library, JENDL-3.3, was used as neutron cross section data. The neutron diffusion and transport calculations were performed using a diffusion code, CITATION, in the SRAC code system and a continuous-energy Monte Carlo code, MVP. Critical level heights of the cores were obtained using such parameters as uranium concentration (up to 500 gU/l), free nitric acid concentration (up to 8 mol/l), and concentration of soluble neutron poisons, gadolinium and boron. It has been confirmed from the evaluation that all critical cores comply with safety criteria required in the STACY operation concerning excess reactivity, reactivity addition rates and shutdown margins by safety rods. (author)
International Nuclear Information System (INIS)
Matsubayashi, Masahito; Sudo, Yukio; Haga, Katsuhiro
1996-01-01
In order to make clear the flow mechanism and characteristics of falling water limitation under the countercurrent two-phase flow, that is, the countercurrent flow limitation (CCFL), in a vertical channel, a technique of neutron radiography (NRG) provided in the Research Nuclear Reactor JRR-3M was applied to an air-water system of vertical rectangular channels of 50 and 782 mm in length with 66 mm in channel width and 2.3 mm in channel gap under atmospheric pressure. The neutron radiography facility used in this study has a high thermal neutron flux that is suitable for visualization of fluid phenomena. A real-time electronic imaging method was used for capturing two-phase flow images in a vertical channel. It was found the technique applied was very potential to clarify the characteristics of instantaneous, local and average void fractions which were important to understand flow mechanism of the phenomena, while the measurements of void fraction had not been applied fully effectively to understanding of the flow mechanism of CCFL, because the differential pressure for determining void fraction is, in general, too small along the tested channel and is fluctuating too frequently to be measured accurately enough. From the void fraction measured by NRG as well as through direct flow observation, it was revealed that the shorter side walls of rectangular channel tested were predominantly wetted by water falling down with the longer side walls being rather dry by ascending air flow. It was strongly suggested that the analytical flow model thus obtained and proposed for the CCFL based on the flow observation was most effective
Studies of frequency dependent C-V characteristics of neutron irradiated p+-n silicon detectors
International Nuclear Information System (INIS)
Li, Zheng; Kraner, H.W.
1990-10-01
Frequency-dependent capacitance-voltage fluence (C-V) characteristics of neutron irradiated high resistivity silicon p + -n detectors have been observed up to a fluence of 8.0 x 10 12 n/cm 2 . It has been found that frequency dependence of the deviation of the C-V characteristic (from its normal V -1/2 dependence), is strongly dependent on the ratio of the defect density and the effective doping density N t /N' d . As the defect density approaches the effective dopant density, or N t /N' d → 1, the junction capacitance eventually assumes the value of the detector geometry capacitance at high frequencies (f ≤ 10 5 Hz), independent of voltage. A two-trap-level model using the concept of quasi-fermi levels has been developed, which predicts both the effects of C-V frequency dependence and dopant compensation observed in this study
International Nuclear Information System (INIS)
Chamberlain, M; Gräfe, J L; Aslam; Byun, S H; Chettle, D R; Egden, L M; Orchard, G M; Webber, C E; McNeill, F E
2012-01-01
Fluorine (F) plays an important role in dental health and bone formation. Many studies have shown that excess fluoride (F − ) can result in dental or skeletal fluorosis, while other studies have indicated that a proper dosage of fluoride may have a protective effect on bone fracture incidence. Fluorine is stored almost completely in the skeleton making bone an ideal site for measurement to assess long-term exposure. This paper outlines a feasibility study of a technique to measure bone-fluorine non-invasively in the human hand using in vivo neutron activation analysis (IVNAA) via the 19 F(n,γ) 20 F reaction. Irradiations were performed using the Tandetron accelerator at McMaster University. Eight NaI(Tl) detectors arranged in a 4π geometry were employed for delayed counting of the emitted 1.63 MeV gamma ray. The short 11 s half-life of 20 F presents a difficult and unique practical challenge in terms of patient irradiation and subsequent detection. We have employed two simultaneous timing methods to determine the fluorine sensitivity by eliminating the interference of the 1.64 MeV gamma ray from the 37 Cl(n,γ) 38 Cl reaction. The timing method consisted of three counting periods: an initial 30 s (sum of three 10 s periods) count period for F, followed by a 120 s decay period, and a subsequent 300 s count period to obtain information pertaining to Ca and Cl. The phantom minimum detectable limit (M DL ) determined by this method was 0.96 mg F/g Ca. The M DL was improved by dividing the initial timing period into three equal segments (10 s each) and combining the results using inverse variance weighting. This resulted in a phantom M DL of 0.66 mg F/g Ca. These detection limits are comparable to ex vivo results for various bones in the adult skeleton reported in the literature. Dosimetry was performed for these irradiation conditions. The equivalent dose for each phantom measurement was determined to be 30 mSv. The effective dose was however low, 35 µSv, which is
The stochastic characteristics stability in the problem of the dynamic filtration with delay
Chashnikov, Mikhail
2017-07-01
In this paper thesolution's calculation of linear differential equation with delay is considered. It is supposed that initial conditions are the set with random error and we have chance of updating the solution in periodic timepointswith the same error. The dependence between the estimate dispersion and the measuring error dispersion is recieved.
International Nuclear Information System (INIS)
Yamamoto, Kazuyoshi; Kumada, Hiroaki; Kishi, Toshiaki; Torii, Yoshiya; Horiguchi, Yoji
2001-03-01
To evaluate nitrogen dose, boron dose and gamma-ray dose occurred by neutron capture reaction of the hydrogen at the medical irradiation, two-dimensional distribution of the thermal neutron flux is very important because these doses are proportional to the thermal neutron distribution. This report describes the measurement of the two-dimensional thermal neutron distribution in a head water phantom by neutron beams of the JRR-4 and evaluation of the dose distribution characteristic. Thermal neutron flux in the phantom was measured by gold wire placed in the spokewise of every 30 degrees in order to avoid the interaction. Distribution of the thermal neutron flux was also calculated using two-dimensional Lagrange's interpolation program (radius, angle direction) developed this time. As a result of the analysis, it was confirmed to become distorted distribution which has annular peak at outside of the void, though improved dose profile of the deep direction was confirmed in the case which the radiation field in the phantom contains void. (author)
CACTUS, a characteristics solution to the neutron transport equations in complicated geometries
International Nuclear Information System (INIS)
Halsall, M.J.
1980-04-01
CACTUS has been written to solve the multigroup neutron transport equation in a general two-dimensional geometry. The method is based upon a characteristics formulation for the problem in which the transport equation is integrated explicitly along straight line tracks that are suitably distributed throughout the problem. Source distributions and scattering are assumed to be isotropic, but the only restriction on geometry is that the outer boundary should be rectangular. Within this rectangular boundary the user is free to build his problem geometry using any combination of intersecting straight lines and circular arcs. The theory of the method is described, followed by some details of a coding, a sensitivity study on the number of tracks required to integrate fluxes in a particular problem, a user's guide, and a few test cases. (author)
Neutron characteristics of the Super-Phenix 1 reactor at Creys-Malville
International Nuclear Information System (INIS)
Giacometti, C.; Bouget, Y.H.; Hammer, P.; Lyon, F.; Salvatores, M.; Sicard, B.; Pipaud, J.Y.
1980-01-01
The paper describes the method used to determine the critical enrichments for the first loading of the Super-Phenix reactor and the correction factors (together with their uncertainties) applied to the data calculated from the CARNAVAL IV code. These enrichments must be chosen so as to conform to the planned operating conditions of the reactor: nominal power of the pressure vessels, lifetime of the in-pile assemblies. Allowance for uncertainties of neutronic origin and those associated with the fabrication of the fuel pins calls for an over-enrichment of the first loading by approximately 4 per cent. An analysis is made of the effects of this over-enrichment on the core characteristics, which have to remain compatible with the established limits. (author)
Characteristics of self-powered neutron detectors used in power reactors
International Nuclear Information System (INIS)
Todt, William H. Sr.
1998-01-01
Self-powered neutron detectors have been used effectively as in-core flux monitors for over twenty-five years in nuclear power reactors worldwide. This paper describes the basic properties of these radiation sensors including their nuclear, electrical and mechanical characteristics. Recommendations are given for the proper choice of the self-powered detector emitter to provide the proper response time and radiation sensitivity desired for use in an effective in-core radiation monitoring system. Examples are shown of specific self-powered detector designs, which are being effectively, used in in-core instrumentation systems for pressurized water, heavy water and graphite moderated light water reactors. Also examples are shown of the mechanical configurations of in-core assemblies of self-powered detectors combined with in-core thermocouples presently used in pressurized water and heavy water reactors worldwide. (author)
International Nuclear Information System (INIS)
Hunt, Alan; Tobin, S. J.
2015-01-01
In this two year project, the research team investigated how delayed γ-rays from short-lived fission fragments detected in the short interval between irradiating pulses can be exploited for advanced safeguards technologies. This program contained experimental and modeling efforts. The experimental effort measured the emitted spectra, time histories and correlations of the delayed γ-rays from aqueous solutions and solid targets containing fissionable isotopes. The modeling effort first developed and benchmarked a hybrid Monte Carlo simulation technique based on these experiments. The benchmarked simulations were then extended to other safeguards scenarios, allowing comparisons to other advanced safeguards technologies and to investigate combined techniques. Ultimately, the experiments demonstrated the possible utility of actively induced delayed γ-ray spectroscopy for fissionable material assay.
Energy Technology Data Exchange (ETDEWEB)
Hunt, Alan [Idaho State Univ., Pocatello, ID (United States). Idaho Accelerator Center, Dept. of Physics; Reedy, E. T.E. [Idaho State Univ., Pocatello, ID (United States). Dept. of Phyics, Idaho Accelerator Center; Mozin, V. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Tobin, S. J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Nuclear Nonproliferation
2015-02-12
In this two year project, the research team investigated how delayed γ-rays from short-lived fission fragments detected in the short interval between irradiating pulses can be exploited for advanced safeguards technologies. This program contained experimental and modeling efforts. The experimental effort measured the emitted spectra, time histories and correlations of the delayed γ-rays from aqueous solutions and solid targets containing fissionable isotopes. The modeling effort first developed and benchmarked a hybrid Monte Carlo simulation technique based on these experiments. The benchmarked simulations were then extended to other safeguards scenarios, allowing comparisons to other advanced safeguards technologies and to investigate combined techniques. Ultimately, the experiments demonstrated the possible utility of actively induced delayed γ-ray spectroscopy for fissionable material assay.
International Nuclear Information System (INIS)
Hoffman, Adam J.; Lee, John C.
2016-01-01
A new time-dependent Method of Characteristics (MOC) formulation for nuclear reactor kinetics was developed utilizing angular flux time-derivative propagation. This method avoids the requirement of storing the angular flux at previous points in time to represent a discretized time derivative; instead, an equation for the angular flux time derivative along 1D spatial characteristics is derived and solved concurrently with the 1D transport characteristic equation. This approach allows the angular flux time derivative to be recast principally in terms of the neutron source time derivatives, which are approximated to high-order accuracy using the backward differentiation formula (BDF). This approach, called Source Derivative Propagation (SDP), drastically reduces the memory requirements of time-dependent MOC relative to methods that require storing the angular flux. An SDP method was developed for 2D and 3D applications and implemented in the computer code DeCART in 2D. DeCART was used to model two reactor transient benchmarks: a modified TWIGL problem and a C5G7 transient. The SDP method accurately and efficiently replicated the solution of the conventional time-dependent MOC method using two orders of magnitude less memory.
Energy Technology Data Exchange (ETDEWEB)
Hoffman, Adam J., E-mail: adamhoff@umich.edu; Lee, John C., E-mail: jcl@umich.edu
2016-02-15
A new time-dependent Method of Characteristics (MOC) formulation for nuclear reactor kinetics was developed utilizing angular flux time-derivative propagation. This method avoids the requirement of storing the angular flux at previous points in time to represent a discretized time derivative; instead, an equation for the angular flux time derivative along 1D spatial characteristics is derived and solved concurrently with the 1D transport characteristic equation. This approach allows the angular flux time derivative to be recast principally in terms of the neutron source time derivatives, which are approximated to high-order accuracy using the backward differentiation formula (BDF). This approach, called Source Derivative Propagation (SDP), drastically reduces the memory requirements of time-dependent MOC relative to methods that require storing the angular flux. An SDP method was developed for 2D and 3D applications and implemented in the computer code DeCART in 2D. DeCART was used to model two reactor transient benchmarks: a modified TWIGL problem and a C5G7 transient. The SDP method accurately and efficiently replicated the solution of the conventional time-dependent MOC method using two orders of magnitude less memory.
DEFF Research Database (Denmark)
Schönewolf-Greulich, Bitten; Ravn, Kirstine; Hamborg-Petersen, Bente
2013-01-01
delay/intellectual disability. In contrast small duplications of 4p are rare but with the advent of microarray techniques a few cases have been reported in recent years. Here we describe a 3 Mb duplication at 4p16.3 segregating with a characteristic phenotype, macrocephaly, speech delay and mild...
Neutron shielding characteristics of nano-B2O3 dispersed Poly Vinyl Alcohol
International Nuclear Information System (INIS)
Kim, Jae Woo; Uhm, Young Rang; Lee, Min Ku; Lee, Hee Min; Rhee, Chang Kyu
2008-01-01
Neutron is sometimes beneficiary to human beings while they are unwanted for most cases same as the other radiations such as gamma, beta, and alpha, etc. do. Shielding for neutrons therefore is extremely important to keep the radiation environment safe. Especially, it is critical to absorb (or shield) neutrons generated from the spent fuel in a container/storage, nuclear reactor, and cyclotron, etc. In this regard, light materials containing neutron absorbers such as borated-polymers are very useful to shield neutrons in those radiation environments. This investigation is focused on the development of borated polymer-based materials whose neutron shielding efficiency is greatly enhanced by using nano sized boron compounds. Boron is well known as a thermal neutron absorber due to its large thermal neutron absorption cross-section (σ th = 760 b, b = 10 -2 - 4 cm 2 ). Although absorption of neutrons in the medium is mainly dependent on the boron atomic weight concentration, we firstly observed the size of boron particles also has an important role in neutron shielding. Mean free path of neutrons colliding with the smaller particles dispersed in the medium might be decreased when it is compared to the larger particles at the same atomic weight concentration. This means that the neutron shielding efficiency of a polymer mixed with the smaller boron compounds is higher than that of a polymer mixed with the larger boron compounds at the same atomic weight boron concentration
Oral Health Characteristics and Dental Rehabilitation of Children with Global Developmental Delay
Directory of Open Access Journals (Sweden)
Saurabh Kumar
2017-01-01
Full Text Available Global developmental delay (GDD is a chronic neurological disturbance which includes defects in one or more developmental domains. The developmental domain can be motor, cognitive, daily activities, speech or language, and social or personal development. The etiology for GDD can be prenatal, perinatal, or postnatal. It can be diagnosed early in childhood as the delay or absence of one or more developmental milestones. Hence the role of pedodontist and pediatricians becomes more crucial in identifying this condition. The diagnosis of GDD requires a detailed history including family history and environmental risk factors followed by physical and neurological examinations. Investigations for GDD include diagnostic laboratory tests, brain imaging, and other evidence-based evaluations. GDD affects multiple developmental domains that not only have direct bearing on maintenance of oral health, but also require additional behavior management techniques to deliver optimal dental care. This paper describes two different spectra of children with GDD. Since the severity of GDD can vary, this paper also discusses the different behavior management techniques that were applied to provide dental treatment in such children.
Oral Health Characteristics and Dental Rehabilitation of Children with Global Developmental Delay.
Kumar, Saurabh; Pai, Deepika; Saran, Runki
2017-01-01
Global developmental delay (GDD) is a chronic neurological disturbance which includes defects in one or more developmental domains. The developmental domain can be motor, cognitive, daily activities, speech or language, and social or personal development. The etiology for GDD can be prenatal, perinatal, or postnatal. It can be diagnosed early in childhood as the delay or absence of one or more developmental milestones. Hence the role of pedodontist and pediatricians becomes more crucial in identifying this condition. The diagnosis of GDD requires a detailed history including family history and environmental risk factors followed by physical and neurological examinations. Investigations for GDD include diagnostic laboratory tests, brain imaging, and other evidence-based evaluations. GDD affects multiple developmental domains that not only have direct bearing on maintenance of oral health, but also require additional behavior management techniques to deliver optimal dental care. This paper describes two different spectra of children with GDD. Since the severity of GDD can vary, this paper also discusses the different behavior management techniques that were applied to provide dental treatment in such children.
International Nuclear Information System (INIS)
Lovchikova, G.N.; Polyakov, A.V.; Sal'nikov, O.A.; Simakov, S.P.; Sukhikh, S.Eh.; Trufanov, A.M.
1983-01-01
The possibility to use gas tritium target as neutron source with the energy 2 MeV for nuclear-physical studies has been considered. Characteristics of neutron flux crested in the reaction 3 H(d, n) 4 He to obtain neutrons are investigated. The study of inelastic scattering processes at the energies permits to expand the experiments conducted up to the present day on the study of spectra of inelastically scattered neutrons in a lower energy region and it is of interest for the clarification of appearance mechanism of high-energy neutrons in the spectra. Characteristics of neutron flux as a result of the reaction 3 (α, n) 4 He at the energy of falling deuterons Esub(d)=5.54 MeV are investigated. Measurements of spectra of scattered neutrons on carbon-12 at the angles 30, 45, 60, 90, 120, 150 degrees are made. Differential cross sections of elastic scattering are obtained
Özdaş, Talih; Şahlı, Ayşe Sanem; Özdemir, Behiye Sarıkaya; Belgin, Erol
2018-01-05
Speech delay in a child could be the cause and/or result of the emotional disorder. The child rearing attitude that the parents have accepted could have both positive and negative effects on the personality of the child. The current study aimed to investigate the sociodemographic features and the mothers' anxiety of children with speech delay. One hundred five mothers with children aged between 3 and 6 years with speech delays were included in the patient group, and 105 mothers who have children aged between 3 and 6 years with normal speech and language development were included in the control group. An information form questionnaire including demographic characteristics, the Family Life and Childrearing Attitude Scale (PARI - Parental Attitude Research Instrument) and beck anxiety scale were requested from all mothers in the patient and the control groups. In the current study, there was a significant difference between the groups in terms of gender (p=0.001). According to Parental Attitude Research Instrument, the mean of mothers of the children with speech delays was higher than the mean of mothers of normal children in terms of the answers to overprotective mother aspect (poverprotective motherhood attitudes; however, the difference in terms of the answers to the aspects of democratic attitude and provision of equality, refusal to be a housewife, husband-wife conflict, and suppression and discipline were not statistically significant. The beck anxiety scale, a significant difference was detected between the two groups (p<0.01). It was found that the mothers of children with speech delays had more severe levels of anxiety. The social structure of the family, the attitudes and the behaviors of the mother, and the anxiety levels of the mothers have important effects on child development. Thus, it is necessary to perform further studies related to speech delays, in which many factors play a role in the etiology. Copyright © 2017 Associação Brasileira de
Identification of the fast and thermal neutron characteristics of transuranic waste drums
Energy Technology Data Exchange (ETDEWEB)
Storm, B.H. Jr.; Bramblett, R.L. [Lockheed Martin Specialty Components, Largo, FL (United States); Hensley, C. [Oak Ridge National Lab., TN (United States)
1997-11-01
Fissile and spontaneously fissioning material in transuranic waste drums can be most sensitively assayed using an active and passive neutron assay system such as the Active Passive Neutron Examination and Assay. Both the active and the passive assays are distorted by the presence of the waste matrix and containerization. For accurate assaying, this distortion must be characterized and accounted for. An External Matrix Probe technique has been developed that accomplishes this task. Correlations between in-drum neutron flux measurements and monitors in the Active Passive Neutron Examination and Assay chamber with various matrix materials provide a non-invasive means of predicting the thermal neutron flux in waste drums. Similarly, measures of the transmission of fast neutrons emitted from sources in the drum. Results obtained using the Lockheed Martin Specialty Components Active Passive Neutron Examination and Assay system are discussed. 12 figs., 1 tab.
Self-regulating characteristics of cold neutron source with annular cylindrical moderator cell
International Nuclear Information System (INIS)
Kawai, Takeshi; Lee Chien-Hsiung; Chan Yea-Kuang; Guung Tai-Cheng; Yoshino, Hirofumi; Kawabata, Yuji; Hino, Masahiro
2001-01-01
The conditions, in which the ORPHEE type cold neutron source with an annular cylindrical moderator cell could have self-regulating characteristics, were obtained through thermodynamic considerations. From a viewpoint of engineering, it is not easy to establish these conditions because three parameters are involved even in an idealized system without the effect of the mass transfer resistance in the moderator transfer tube between the condenser and the moderator cell. The inner shell of the ORPHEE moderator cell is open in the bottom, but it is expected that only hydrogen vapor is contained in the inner shell and liquid hydrogen in the outer shell. The thermodynamic considerations show that such a state is maintained only when a liquefaction capacity of the condenser is large compared to heat lead and three parameters are optimized with a good balance. We proposed another type of a moderator cell, which has an inner cylindrical cavity with no hole in the bottom but a vapor inlet opening at the uppermost part of the cavity. In this structure, a self-regulating characteristic is established easily and the liquid level in the outer shell is maintained almost constant against thermal disturbances. Therefore it is enough to control one parameter, that is, the reservoir tank pressure corresponding to the liquefaction capacity of the condenser given by the refrigeration power of the helium refrigerator. (author)
Characteristics of self-powered neutron detectors used in power reactors
International Nuclear Information System (INIS)
Todt, W.H.
1997-01-01
Self-Powered Neutron Detectors have been used effectively as in-core flux monitors for over twenty-five years in nuclear power reactors world-wide. The basic properties of these radiation sensors are described including their nuclear, electrical and mechanical characteristics. Recommendations are given for the proper choice of the self-powered detector emitter to provide the proper response time and radiation sensitivity desired for use in an effective in-core radiation monitoring system. Examples are shown of specific self-powered detector designs which are being effectively used in in-core instrumentation systems for pressurised water, heavy water and graphite moderated light water reactors. Examples are also shown of the mechanical configurations of in-core assemblies of self-powered detectors combined with in-core thermocouples presently used in pressurised water and heavy water reactors worldwide. This paper is a summary of a new IEC standard to be issued in 1996 describing the characteristics and test methods of self-powered detectors used in nuclear power reactors. (author)
Studies on the characteristics of nuclear track spark counting for neutron monitoring
International Nuclear Information System (INIS)
Kawai, H.; Koga, T.; Morishima, H.; Niwa, T.; Nishiwaki, Y.
1979-01-01
The fission-track counting method using polycarbonate foil is known to be very effective for neutron monitoring. The detection system consists of a track detector foil placed in contact with a thin layer of fissile material. When the unit is irradiated with neutrons, fission fragments from the fissile material produce tracks in the foil. The foil is then etched chemically to enlarge the tracks to etch-pits, which are counted by an automatic spark counting method suggested by Cross and Tommasino. Spark counting characteristics of the above system using polycarbonate detector foils of different thicknesses and aluminized polyester sheets having different thicknesses of the aluminium which serves as an electrode, and etch-pit sizes in relation to spark counting have been studied. The results obtained are summarized as follows: 1. A growth in etch-pit sizes was observed for etching times of 10 to 50 minutes with Makrofol KG (10 μm thick) and Panlite foils (18 μm thick). As the etching time increased, the number of etch-pits which appeared as round pits increased and the number which appeared rod-like decreased; 2. Round etch-pits seemed to be easily spark-punched; 3. Major and minor diameters of etch-pits produced by fission fragments from the uranium fissile target were larger than those from the thorium target; 4. As the thickness of the Makrofol KG foil increased, the minor diameters of the etch-pits appeared to become smaller and major diameters larger; 5. The foils that appeared to be best suited for routine use were 10 μm thick Makrofol KG or 6 μm thick Lumirror foils. (author)
International Nuclear Information System (INIS)
Carrel, F.
2007-10-01
An accurate estimation of the alpha-activity of a nuclear waste package is necessary to select the best mode of storage. The main purpose of this work is to develop a non-destructive active method, based on the fission process and allowing the identification of actinides ( 235 U, 238 U, 239 Pu). These three elements are the main alpha emitters contained inside a package. Our technique is based on the detection of delayed gammas emitted by fission products. These latter are created by irradiation with the help of a neutron or photon beam. Performances of this method have been investigated after an Active Photon or Neutron Interrogation (INA or IPA). Three main objectives were fixed in the framework of this thesis. First, we measured many yields of photofission products to compensate the lack of data in the literature. Then, we studied experimental performances of this method to identify a given actinide ( 239 Pu in fission, 235 U in photofission) present in an irradiated mixture. Finally, we assessed the application of this technique on different mock-up packages for both types of interrogation (118 l mock-up package containing EVA in fission, 220 l mock-up package with a wall of concrete in photofission). (author)
International Nuclear Information System (INIS)
Maksimenko, A.N.; Basin, Ya.N.; Novgorodov, V.A.
1974-01-01
To isolate reservoirs, the formation and deformation penetration zone parameters are used. They are estimated according to the false oil saturation factor and the time of the penetration zone deformation which are determined from the complex exploration of cased wells using the pulse neutron logging, thermal neutron-neutron logging and gamma logging techniques
Psychosocial Characteristics of Women with a Delayed Diagnosis of Turner Syndrome.
Reimann, Gabrielle E; Bernad Perman, Martha M; Ho, Pei-Shu; Parks, Rebecca A; Comis, Leora E
2018-05-09
To characterize the psychosocial profiles of adult women diagnosed with Turner syndrome before (early diagnosis) and at or after (late diagnosis) 13 years of age. Women with Turner syndrome ages 22 and older at evaluation (n = 110) participated in a cross-sectional study at the National Institutes of Health. Researchers performed nonparametric and logistic regression analyses to assess early and late diagnosis cohorts on measures of depression, substance use, and perceptions of competence and identity. Of study participants, 47% received a Turner syndrome diagnosis at or after age 13 years. Median age at diagnosis was 12.0 years (range, 0-43). Covariate-adjusted models revealed that women with late diagnoses had an increased likelihood of developing mild to severe depressive symptoms (OR, 7.36) and a decreased likelihood of being perceived as competent (OR, 0.26). Women with a late diagnosis also exhibited more frequent substance use compared with women with early diagnoses. These data suggest that Turner syndrome diagnoses received at or after age 13 years may contribute to adverse outcomes related to depression, substance use, and perceptions of competence. Delayed Turner syndrome diagnoses may place women and girls at risk for negative psychosocial development extending into adulthood. These findings indicate it is important for pediatricians to evaluate psychosocial domains in girls with Turner syndrome regularly, particularly among those diagnosed at age 13 years or older. ClinicalTrials.gov: NCT00006334. Published by Elsevier Inc.
Some neutron and gamma radiation characteristics of plutonium cermet fuel for isotopic power sources
Neff, R. A.; Anderson, M. E.; Campbell, A. R.; Haas, F. X.
1972-01-01
Gamma and neutron measurements on various types of plutonium sources are presented in order to show the effects of O-17, O-18 F-19, Pu-236, age of the fuel, and size of the source on the gamma and neutron spectra. Analysis of the radiation measurements shows that fluorine is the main contributor to the neutron yields from present plutonium-molybdenum cermet fuel, while both fluorine and Pu-236 daughters contribute significantly to the gamma ray intensities.
International Nuclear Information System (INIS)
Vanhavere, F.
2001-01-01
The objective of SCK-CEN's R and D programme on neutron dosimetry is to improve the determination of neutron doses by studying neutron spectra, neutron dosemeters and shielding adaptations. In 2000, R and D focused on the contiued investigation of the bubble detectors type BD-PND and BDT, in particular their sensitivity and temperature dependence; the updating of SCK-CEN's criticality dosemeter, the investigation of the characteristics of new thermoluminescent materials and their use in neutron dosemetry; and the investigation of neutron shielding
Energy Technology Data Exchange (ETDEWEB)
Vanhavere, F
2001-04-01
The objective of SCK-CEN's R and D programme on neutron dosimetry is to improve the determination of neutron doses by studying neutron spectra, neutron dosemeters and shielding adaptations. In 2000, R and D focused on the contiued investigation of the bubble detectors type BD-PND and BDT, in particular their sensitivity and temperature dependence; the updating of SCK-CEN's criticality dosemeter, the investigation of the characteristics of new thermoluminescent materials and their use in neutron dosemetry; and the investigation of neutron shielding.
Effect of Fast Neutron to MA/PU Burning/Transmutation Characteristic Using a Fast Reactor
International Nuclear Information System (INIS)
Marsodi; Lasman, As Natio; Kimamoto, A.; Marsongkohadi; Zaki, S.
2003-01-01
MA/Pu burning/transmutation has been studied and evaluated using fast neutrons. Generally, neutron density at this fast burner reactor and transmutation has spectrum energy level around 0.2 MeV with wide enough variation, i.e. from low neutron spectrum to its peak is 0.2 MeV. This neutron spectrum energy level depends on the kind of cooler material or fuel used. Neutron spectrum higher than fast power reactor neutron spectrum is found by means of changing oxide fuel by metallic fuel and changing natrium cooler material by metallic or gas cooler material. This evaluation is conducted by various variations in accordance with the kind of fuel or cooler, MA/Pu fractions and fuel comparison fraction with respect to its cooler in order to get better neutron usage and MA/Pu burning speed. Reactor calculation evaluation in this paper was conducted with 26-group nuclear data cross section energy spectrum. The main purpose of the discussion is to know the effect of fast neutrons to burning/transmutation MA/Pu using fast neutrons
International Nuclear Information System (INIS)
Young, Ryong Park; Nam, Zin Cho
2005-01-01
As the nuclear reactor core becomes more complex, heterogeneous, and geometrically irregular, the method of characteristics (MOC) is gaining its wide use in the neutron transport calculations. However, the long computing times require good acceleration methods. In this paper, the concept of coarse-mesh angular dependent re-balance (CMADR) acceleration is described and applied to the MOC calculation in x-y-z (z-infinite, uniform) geometry. The method is based on the angular dependent re-balance factors defined only on the coarse-mesh boundaries; a coarse-mesh consists of several fine meshes that may be heterogeneous and of mixed geometries with irregular or unstructured mesh shapes. In addition, the coarse-mesh boundaries may not coincide with the structural interfaces of the problem and can be chosen artificially for convenience. CMADR acceleration is tested on several test problems and the results show that CMADR is very effective in reducing the number of iterations and computing times of MOC calculations. Fourier analysis is also provided to investigate convergence of the CMADR method analytically and the results show that CMADR acceleration is unconditionally stable. (authors)
Effect of burnup history by moderator density on neutron-physical characteristics of WWER-1000 core
International Nuclear Information System (INIS)
Ovdiienko, I.; Kuchin, A.; Khalimonchuk, V.; Ieremenko, M.
2011-01-01
Results of assessment of burnup history effect by moderator density on neutron physical characteristics of WWER-1000 core are presented on example of stationary fuel loading with Russian design fuel assembly TWSA and AER benchmark for Khmelnitsky NPP that was proposed by TUV and SSTC NRC at nineteenth symposium. Assessment was performed by DYN3D code and cross section library sets generated by HELIOS code. Burnup history was taken into account by preparing of numerous cross section sets with different isotopic composition each of which was obtained by burning under different moderator density. For analysis of history effect 20 cross section sets were prepared for each fuel assembly corresponded to each of 20 axial layers of reactor core model for DYN3D code. Four fuel cycles were modeled both for stationary fuel loading with TWSA and AER benchmark for Khmelnitsky NPP to obtain steady value of error due to neglect of burnup history effect. Main attention of study was paid to effect of burnup history by moderator density to axial power distribution. Results of study for AER benchmark were compared with experimental values of axial power distribution for fuel assemblies of first, second, third and fourth year operation. (Authors)
Alekseev, A G; Britvich, G I; Kosyanenko, E V; Pikalov, V A; Gomonov, I P
2003-01-01
n this work the observed data of neutron spectra on Rostov NEP, Kursk NEP and Smolensk NEP and on the reactor IRT MIPHI are submitted. For measurement of neutron spectra two types of spectrometer were used: SHANS (IHEP design ) and SDN-MS01 (FEI design). The comparison of the data measurements per-formed by those spectrometers above one-type cells on the reactor RBMK is submitted. On the basis of the 1-st horizontal experimental channel HEC-1 of the IRT reactor 4 reference fields of neutrons are investigated. It is shown, that spectra of neutrons of reference fields can be used for imitation of neutron spectra for conditions of NEP with VVER and RBMK type reactors.
Characteristics of Pb{sup 2+} doped CsI matrix under gamma and neutron excitations
Energy Technology Data Exchange (ETDEWEB)
Pereira, Maria da Conceição Costa; Madi Filho, Tufic; Berretta, José Roberto; Tomaz, Lucas Faustino; Pinto, Marlene Cristina, E-mail: macoper@ipen.br, E-mail: tmfilho@ipen.br, E-mail: jrretta@ipen.br, E-mail: lucasfaus@gmail.com, E-mail: tynapinto@gmail.com [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)
2017-07-01
In recent years, there has been an increasing interest in finding new fast scintillating material or improve the characteristics of known scintillators for the demand of high energy physics, industrial and nuclear medical applications. Ions divalent lead Pb{sup 2+} built in some crystal structures are efficient emission centers and their applications in scintillators was and still is the reason of an intensive study of emission properties of different compounds containing these ions. The aggregation of impurities in CsI is poorly studied. The problem of impurity aggregation in CsI is of interest from point of view of the luminescent properties modification of this scintillation material. In this context, the crystals of Pb{sup 2+} doped CsI matrix were grown by the vertical Bridgman technique and subjected to annealing in vacuum of 10{sup -6} mbar and continuous temperature of 350°C, for 24 hours, and then they were employed. To evaluate the response of the CsI:Pb scintillator crystal to gamma radiation, radioactive sources of {sup 137}Cs (662 keV), {sup 60}Co (1173 keV and 1333 keV), {sup 22}Na (511 keV and 1275 keV) and {sup 13}3Ba (355 keV) was used. The operating voltage of the photomultiplier was 2700 V for the detection of gamma rays and the accumulation time in the counting process was 600 s. The crystals used in gamma spectroscopy were cut with dimensions of 20 mm in diameter and 20 mm in height. The scintillator response to neutron radiation from a radioactive source of Am/Be with energy range of 1 MeV to 12 MeV was available. The activity of the AmBe source was 1Ci Am. The fluency was 2.6 x 10{sup 6} neutrons / second. The operating voltage of the photomultiplier tube was 1300 V. The accumulation time in the counting process was 600 s. The scintillator crystals were cut with dimensions of 20 mm diameter and 10 mm height. With the results obtained, it may be observed that the crystals are sensitive to these radiations. (author)
Prediction of in-phantom dose distribution using in-air neutron beam characteristics for BNCS
Energy Technology Data Exchange (ETDEWEB)
Verbeke, Jerome M.
1999-12-14
A monoenergetic neutron beam simulation study is carried out to determine the optimal neutron energy range for treatment of rheumatoid arthritis using radiation synovectomy. The goal of the treatment is the ablation of diseased synovial membranes in joints, such as knees and fingers. This study focuses on human knee joints. Two figures-of-merit are used to measure the neutron beam quality, the ratio of the synovium absorbed dose to the skin absorbed dose, and the ratio of the synovium absorbed dose to the bone absorbed dose. It was found that (a) thermal neutron beams are optimal for treatment, (b) similar absorbed dose rates and therapeutic ratios are obtained with monodirectional and isotropic neutron beams. Computation of the dose distribution in a human knee requires the simulation of particle transport from the neutron source to the knee phantom through the moderator. A method was developed to predict the dose distribution in a knee phantom from any neutron and photon beam spectra incident on the knee. This method was revealed to be reasonably accurate and enabled one to reduce by a factor of 10 the particle transport simulation time by modeling the moderator only.
Prediction of in-phantom dose distribution using in-air neutron beam characteristics for BNCS
International Nuclear Information System (INIS)
Verbeke, Jerome M.
1999-01-01
A monoenergetic neutron beam simulation study is carried out to determine the optimal neutron energy range for treatment of rheumatoid arthritis using radiation synovectomy. The goal of the treatment is the ablation of diseased synovial membranes in joints, such as knees and fingers. This study focuses on human knee joints. Two figures-of-merit are used to measure the neutron beam quality, the ratio of the synovium absorbed dose to the skin absorbed dose, and the ratio of the synovium absorbed dose to the bone absorbed dose. It was found that (a) thermal neutron beams are optimal for treatment, (b) similar absorbed dose rates and therapeutic ratios are obtained with monodirectional and isotropic neutron beams. Computation of the dose distribution in a human knee requires the simulation of particle transport from the neutron source to the knee phantom through the moderator. A method was developed to predict the dose distribution in a knee phantom from any neutron and photon beam spectra incident on the knee. This method was revealed to be reasonably accurate and enabled one to reduce by a factor of 10 the particle transport simulation time by modeling the moderator only
International Nuclear Information System (INIS)
Bessis, J.
1986-09-01
Methods are described for calculating the probabilities, p(m), of detection of m neutrons, inside a split millisecond counting gate, m varying from zero to some units. At the present stage, these methods suppose the source to be very small. Using the generating function concept, they concern both possible modes of the counting system, for opening gates, i.e.: 1) Trigger pulses randomly with regard to the emitted neutrons, 2) Trigger pulses from the detected neutrons themselves. Computed values are finally compared to the measured ones. This comparison seems to be very favourable, since the respective deviations are often lower than 1 % [fr
Method and apparatus for measuring neutron characteristics of material surrounding a borehole
International Nuclear Information System (INIS)
Hopkinson, E.C.
1983-01-01
This invention relates to methods and apparatus for determining the macroscopic thermal neutron absorption cross section of the formations surrounding a borehole as determined by radiation measurements using optimized measurement intervals. A measurement of the decline of the thermal neutron population in the formation is derived by counting the detected radiation within a first pair of measurement intervals occurring at a fixed time after the neutron burst. A ratio of the two counting rates provides the rate of change over the selected time interval. The counting ratio is converted into a natural logarithm representative of the Sigma calculation
Energy Technology Data Exchange (ETDEWEB)
Cho, Soo Haeng; Hong, Sun Seok; Oh, Seung Chul; Do, Jae Bum [Korea Atomic Energy Research Institute, Taejon (Korea)
1998-10-01
Because the exposure to radiation in the nuclear facilities can be fatal to human, it is important to reduce the radiation dose level to a tolerable level. The purpose of this study is to develop highly effective neutron shielding materials for the shipping and storage cask of radioactive materials or in the nuclear/radiation facilities. In this study, we developed modified and novolac type epoxy resin based neutron shielding materials and their various material properties, including neutron shielding ability, prolonged time heat resistance, thermal and mechanical properties were evaluated experimently. (author). 31 refs., 27 figs., 16 tabs.
International Nuclear Information System (INIS)
Somogyi, G.; Dajko, G.; Turek, K.; Spurny, F.
1982-01-01
The density of background spots revealed by chemical (CE) and electrochemical (ECE) etching and by their sequential application (CE + ECE) has been measured in several sorts of CR-39 material. The trends in the variation of sensitivity to Am-Be neutrons have been determined in CR-39 sheets covered by thick proton-radiator, when changing the field strength, frequency, etchant concentration, pre-etch duration and the fluence of neutrons. The results are analyzed in order to find out an optimum set of experimental parameters which may be proposed to attain high registration sensitivity to neutrons. (author)
International Nuclear Information System (INIS)
Garg, A.N.; Vivek Singh; Chutke, N.L.; Ambulkar, M.N.
1995-01-01
In order to determine hazardous effects of smoking five different brands of Indian cigarette tobacco including its ash and smoke, three brands of bi di (a typical of Indian subcontinent) and two of chewing tobacco have been analysed for 24 elements (As, Ba, Br, Ca, Ce, Co, Cr, Cs, Cu, Eu, Fe, Hg, K, Hf, La, Na, P, Rb, Sb, Sc, Se, Sr, Th and Zn) by instrumental neutron activation analysis (INAA). A simple and indigenous smoking device has been developed. A mean ash content of 16.5 ± 1.1% was obtained for different brands of cigarettes whereas smoke content was found to be 4.8±0.8%. It is observed that most elements are retained in ash and a few are transferred to smoke. Percent amount of elements retained in ash and transferred to smoke condensate have been calculated. Significant amounts of Sb, Hg, Co, Se and Zn are transferred to smoke but >90% of Ca, Fe, Sr and Cr are retained in ash. a comparison of elemental contents with the cigarette tobacco from other countries shows comparable amounts for most elements except few minor variations attributable to local soil characteristics. Most elemental contents in bi di and chewing tobacco are comparable except Br, Cr, Na, Se and Sr which are higher in chewing tobacco but its Ca content is lower. Standard Reference Materials Bowen's Kale, Kentucky Reference Cigarette and Citrus Leaves (SRM 1572) along with a newly developed tobacco CRM OTL-1 from Poland were also analysed. (author). 31 refs., 1 fig., 4 tabs
International Nuclear Information System (INIS)
Petrascu, M.; Isbasescu, Alina; Constantinescu, A.; Serban, S.; Stoica, I.V.
2004-01-01
The neutron multidetector consists of 81 detectors, made of 4x4x12 cmc BC-400 crystals mounted on XP2972 phototubes. This detector placed in the forward direction at 138 cm from the target, was used to detect the correlated neutrons in the fusion of Li11 halo nuclei with Si targets. To verify the criterion for selecting the true coincidences against cross-talk ( a spurious effect in which the same neutron is registered by two or more detectors) and to establish the optimal distance between adjacent detectors, the program MENATE ( written by P.Desesquelles, IPN - Orsay) was used to generate Monte Carlo neutrons and their interactions in multidetector. The results were analysed with PAW (from CERN Library). (authors)
International Nuclear Information System (INIS)
Kawai, Takeshi; Utsuro, Masahiko; Okamoto, Sunao
1989-01-01
A cold neutron source (CNS) having a closed-thermosiphon cooling loop shows a characteristic of self-regulation to the heat load fluctuations if the moderator transfer tube fulfills certain conditions. A dynamical equation of the closed-thermosiphon type CNS having such a property has been presented on the basis of the non-equilibrium thermodynamics. Kyoto University Reactor (KUR) CNS is investigated by numerical simulation of this equation. The numerical predictions for the self-regulating characteristics are in agreement with available experimental data. (author)
International Nuclear Information System (INIS)
Fevotte, F.
2008-01-01
At the various stages of a nuclear reactor's life, numerous studies are needed to guaranty the safety and efficiency of the design, analyse the fuel cycle, prepare the dismantlement, and so on. Due to the extreme difficulty to take extensive and accurate measurements in the reactor core, most of these studies are numerical simulations. The complete numerical simulation of a nuclear reactor involves many types of physics: neutronics, thermal hydraulics, materials, control engineering, Among these, the neutron transport simulation is one of the fundamental steps, since it allows computation - among other things - of various fundamental values such as the power density (used in thermal hydraulics computations) or fuel burn-up. The neutron transport simulation is based on the Boltzmann equation, which models the neutron population inside the reactor core. Among the various methods allowing its numerical solution, much interest has been devoted in the past few years to the Method of Characteristics in unstructured meshes (MOC), since it offers a good accuracy and operates in complicated geometries. The aim of this work is to propose improvements of the calculation scheme bound on the two dimensions MOC, in order to decrease the needed resources number. (A.L.B.)
International Nuclear Information System (INIS)
Chernikova, D.; Romodanov, V.L.; Belevitin, A.G.; Afanas'ev, V.V.; Sakharov, V.K.; Bogolubov, E.P.; Ryzhkov, V.I.; Khasaev, T.O.; Sladkov, A.A.; Bitulev, A.A.
2014-01-01
The present paper discusses results of full-scale experimental and numerical investigations of influence of construction materials of portable pulsed neutron generators ING-031, ING-07, ING-06 and ING-10-20-120 (VNIIA, Russia) to their radiation characteristics formed during and after an operation (shutdown period). In particular, it is shown that an original monoenergetic isotropic angular distribution of neutrons emitted by TiT target changes into the significantly anisotropic angular distribution with a broad energy spectrum stretching to the thermal region. Along with the low-energetic neutron part, a significant amount of photons appears during the operation of generators. In the pulse mode of operation of neutron generator, a presence of the construction materials leads to the “tailing” of the original neutron pulse and the appearance of an accompanying photon pulse at ∼3ns after the instant neutron pulse. In addition to that, reactions of neutron capture and inelastic scattering lead to the creation of radioactive nuclides, such as 58 Co, 62 Cu, 64 Cu and 18 F, which form the so-called activation radiation. Thus, the selection of a portable neutron generator for a particular type of application has to be done considering radiation characteristics of the generator itself. This paper will be of interest to users of neutron generators, providing them with valuable information about limitations of a specific generator and with recommendations for improving the design and performance of the generator as a whole
Characteristics and calibration of the transmission-type fast neutron moisture meter
International Nuclear Information System (INIS)
Banzai, K.
1984-01-01
With the Transmission-type Fast Neutron Moisture Meter, we did some experiments for calibration and the effective range of fast neutron scattering, and observed soil moisture process before and after making artificial rainfall at a lysimeter filled by decomposed granite. A fast neutron source of this meter is 252 Cf and capacity of 100 μ Ci. The neutron detector is NE-213 liquid scintilator which recovers a little flux of neutron source. For the customary thermal neutron meter, the effective range of neutron scattering is variable by soil moisture values surrounding the observation point, but this fast neutron, insert and transmission-type meter shows soil moisture in small capacity between a source and a detector. Experimental Results; 1) The calibration curve, calculated statistically from the relation of soil moisture and the count ratio in a 200 l drum packed with beads, gravel, sand and Kanto loam, became only one line. The correlation coefficient of this curve was 0.996 and the standard error was 1.94% with volumetric water content. 2) Count ratio started to decrease as observation point approached soil surface from the boundary of 6 cm depth in soil. Volumetric water content increased more than fact with the previous calibration curve. 3) We limited the detectable range to fast neutron, but a little scattering was seen surrounding the soil of a observation point. The effective range of horizontal scattering was a width of 20 cm with the center line connected between a source and a detector, with a circle of 5 cm diameter surrounding the source, and a circle of 10-15 cm diameter surrounding the detector. 4) Soil moisture before and after artificial rainfall was observed with this meter and by the measurement of a 100 cm 3 oven dried sampling vessel. Volumetric water content by the latter measurement, was more variable because sampling points were at a distance from the center of observation site and sampling technique was bad. Otherwise soil moisture values
Characteristics of the neutron and X-ray tomography system at the SANRAD facility in South Africa
International Nuclear Information System (INIS)
Beer, F.C. de
2005-01-01
Through collaboration with the NEUTRA-facility at the Paul Scherrer Institute (PSI), Switzerland, a turnkey tomography system was designed specifically for the beam geometry at the South African Neutron Radiography (SANRAD) facility, located on the beam port floor of the SAFARI-1 nuclear research reactor and operated by Necsa. The new system is currently being extensively utilized in both 2D and 3D mode for various applications in general industry and institutional activities. The basic performance characteristics of its 3D tomography capability in a neutron and X-ray configuration are presented with the aid of several case studies. An X-ray source has also been commissioned to further diversify the capabilities of the facility
International Nuclear Information System (INIS)
Kim, Bong Hwan
2006-01-01
Neutron field spectrometry using multi spheres such as Bonner Spheres (BS) has been almost essential in radiation protection dosimetry for a long time at workplace in spite of poor energy resolution because it is not asking the fine energy resolution but requiring easy operation and measurement performance over a wide range of energy interested. KAERI has developed and used extended BS system based on a LiI(Eu) scintillator as the representative neutron spectrometry system for workplace monitoring as well as for the quantification of neutron calibration fields such as those recommended by ISO 8529. Major topics in using BS are how close the unfolded spectra is the real one and to minimize the interference of gamma radiation in neutron/gamma mixed fields in case of active instrument such as a BS with a LiI(Eu) scintillator. The former is related with choosing a priori information when unfolding the measured data and the latter is depend on how to discriminate it in intense gamma radiation fields. Influence of a priori information in unfolding and effect of counting loss due to pile-up of signals for the KAERI BS system were investigated analyzing the spectral measurement results of Scattered Neutron Calibration Fields (SNCF)
Development of new multigrid schemes for the method of characteristics in neutron transport theory
International Nuclear Information System (INIS)
Grassi, G.
2006-01-01
This dissertation is based upon our doctoral research that dealt with the conception and development of new non-linear multigrid techniques for the Method of the Characteristics (MOC) within the TDT code. Here we focus upon a two-level scheme consisting of a fine level on which the neutron transport equation is iteratively solved using the MOC algorithm, and a coarse level defined by a more coarsely discretized phase space on which a low-order problem is considered. The solution of this problem is then used in order to correct the angular flux moments resulting from the previous transport iteration. A flux-volume homogenization procedure is employed to evaluate the coarse-level material properties after each transport iteration. This entails the non-linearity of the methods. According to the Generalised Equivalence Theory (GET), additional degrees of freedom are introduced for the low-order problem so that the convergence of the acceleration scheme can be ensured. We present two classes of non-linear methods: transport-like methods and discussion-like methods. Transport-like methods consider a homogenized low-order transport problem on the coarse level. This problem is iteratively solved using the same MOC algorithm as for the transport problem on the fine level. Discontinuity factors are then employed, per region or per surface, in order to reconstruct the currents evaluated by the low-order operator, which ensure the convergence of the acceleration scheme. On the other hand, discussion-like methods consider a low-order problem inspired by diffusion. We studied the non-linear Coarse Mesh Finite Difference (CMFD) method, already present in literature, in the perspective of integrating it into TDT code. Then, we developed a new non-linear method on the model of CMFD. From the latter, we borrowed the idea to establish a simple relation between currents and fluxes in order to obtain a problem involving only coarse fluxes. Finally, those non-linear methods have been
Energy Technology Data Exchange (ETDEWEB)
Glumac, B [Institut Jozef Stefan, Ljubljana (Yugoslavia)
1979-07-01
After the reconstruction of the exposure room of the TRIGA reactor a new measurement of the neutron spectrum behind the fission converter was necessary. The multiple foil method was applied. Following twelve reactions were chosen: 197-Au (n, {gamma}), 63-Cu (n, {gamma}), 59-Co (n, {gamma}), 115-In (n.n'), 64-Zn )n.p), 27-Al (n, {alpha}), 25 Mg (n,p), 56-Fe (n,p), 19 F (n,2n'), 237-Np (n,f), 239 Pu (n,f) and 235-U (n,f). The neutron spectrum was calculated from measured reaction rates using our unfolding code ITERAD. The neutron cross sections library DOCROS 77 was applied for the calculations. (author)
Neutronics of pulsed spallation neutron sources
Watanabe, N
2003-01-01
Various topics and issues on the neutronics of pulsed spallation neutron sources, mainly for neutron scattering experiments, are reviewed to give a wide circle of readers a better understanding of these sources in order to achieve a high neutronic performance. Starting from what neutrons are needed, what the spallation reaction is and how to produce slow-neutrons more efficiently, the outline of the target and moderator neutronics are explained. Various efforts with some new concepts or ideas have already been devoted to obtaining the highest possible slow-neutron intensity with desired pulse characteristics. This paper also reviews the recent progress of such efforts, mainly focused on moderator neutronics, since moderators are the final devices of a neutron source, which determine the source performance. Various governing parameters for neutron-pulse characteristics such as material issues, geometrical parameters (shape and dimensions), the target-moderator coupling scheme, the ortho-para-hydrogen ratio, po...
The analysis of neutron physical characteristics of fast reactors by means of pulsed experiments
International Nuclear Information System (INIS)
Stumbur, Eh.A.; Milyutina, Z.N.
1992-01-01
Possibility is considered for determination of macroscopic cross sections of homogeneous multiplicating media with fast neutrons. It is shown that by means of the critical size, laplaccian and neutron pulse damping decrement measurement results it is possible to obtain values of almost all cross sections of a medium. The method is tested with systems of metal 235 U and BFS-32 assemblies with the composition, typical for fast power reactors. A suitable algorithm is developed for solving nonstationary asymptotic transport problems. Calculation results are compared with experimental ones. 21 refs.; 2 figs.; 3 tabs
Energy Technology Data Exchange (ETDEWEB)
Ilic, R.; Rant, J.; Humar, M.; Somogyi, G.; Hunyadi, I.
1986-01-01
Neutron radiographic properties of recently developed new nuclear track detectors (MA-ND/..cap alpha.., MA-ND/p and MA-ND/p1), manufactured from allyl diglycol carbonate, were studied. It was found that the quality of radiographic image has an optimum at a removed layer thickness of about 0.8 ..mu..m. The image obtained under this condition is characterized by high detection sensitivity to neutrons (approx. 8.10/sup -3/ tracks/nsub(th) when using B converter) and by excellent inherent unsharpness (approx. 5 ..mu..m) as well as high image contrast (maximum value of net optical density is approx. 1.4).
Phonon characteristics of high Tc superconductors from neutron Doppler broadening measurements
International Nuclear Information System (INIS)
Trela, W.J.; Kwei, G.H.; Lynn, J.E.; Meggers, K.
1994-01-01
Statistical information on the phonon frequency spectrum of materials can be measured by neutron transmission techniques if they contain nuclei with low energy resonances, narrow enough to be Doppler-broadened, in their neutron cross sections. The authors have carried out some measurements using this technique for materials of the lanthanum barium cuprate class, La 2-x Ba x CuO 4 . Two samples with slightly different concentrations of oxygen, one being superconductive, the other not, were examined. Pure lanthanum cuprate was also measured. Lanthanum, barium and copper all have relatively low energy narrow resonances. Thus it should be possible to detect differences in the phonons carried by different kinds of atom in the lattice. Neutron cross section measurements have been made with high energy resolution and statistical precision on the 59m flight path of LANSCE, the pulsed spallation neutron source at Los Alamos National Laboratory. Measurements on all three materials were made over a range of temperatures from 15K to 300K, with small steps through the critical temperature region near 27K. No significant changes in the mean phonon energy of the lanthanum atoms were observed near the critical temperature of the super-conducting material. It appears however that the mean phonon energy of lanthanum in the superconductor is considerably higher than that in the non-superconductors. The samples used in this series of experiments were too thin in barium and copper to determine anything significant about their phonon spectra
Self-powered in-core neutron detector assembly with uniform perturbation characteristics
International Nuclear Information System (INIS)
1981-01-01
An in-core neutron detector assembly consisting of a number of longitudinally extending self-powered detectors is described. The uniform mechanical structures and materials are placed symmetrically at each active detector portion thus ensuring that local perturbation factors are uniform. (U.K.)
Radiation damages and electro-conductive characteristics of Neutron-Transmutation-Doped GaAs
Energy Technology Data Exchange (ETDEWEB)
Kuriyama, Kazuo; Sato, Masataka; Sakai, Kiyohiro [Hosei Univ., Koganei, Tokyo (Japan). Coll. of Engineering; Okada, Moritami
1996-04-01
Neutron Transmutation Doping (NTD) method made it possible to do homogeneous doping of impurities and to easily control the doping level. Thus, the method has been put into practice for some materials such as silicon. Here, the annealing behavior of anti-site defects generated in neutron-irradiated GaAs was studied. Electric activations of NTD-impurities were started around 550degC in P1 and P2 radiation fields, which were coincident with the beginning of extinction of electron trapping which was caused by anti-site defects due to fast neutron radiation. The electric resistivities of GaAs in neutron radiation fields; P1, P2 and P3 changed depending with the annealing temperature. The electric resistivities of GaAs in P1 and P2 fields indicate the presence of hopping conduction through radiation damages. The resistance of GaAs irradiated in P1 was smaller by nearly 2 orders than that of the untreated control. Further, the electric activation process for NTD-impurities was investigated using ESR and Raman spectroscopy. (M.N.)
Phonon characteristics of high {Tc} superconductors from neutron Doppler broadening measurements
Energy Technology Data Exchange (ETDEWEB)
Trela, W.J.; Kwei, G.H.; Lynn, J.E. [Los Alamos National Lab., NM (United States); Meggers, K. [Univ. of Kiel (Germany)
1994-12-01
Statistical information on the phonon frequency spectrum of materials can be measured by neutron transmission techniques if they contain nuclei with low energy resonances, narrow enough to be Doppler-broadened, in their neutron cross sections. The authors have carried out some measurements using this technique for materials of the lanthanum barium cuprate class, La{sub 2{minus}x}Ba{sub x}CuO{sub 4}. Two samples with slightly different concentrations of oxygen, one being superconductive, the other not, were examined. Pure lanthanum cuprate was also measured. Lanthanum, barium and copper all have relatively low energy narrow resonances. Thus it should be possible to detect differences in the phonons carried by different kinds of atom in the lattice. Neutron cross section measurements have been made with high energy resolution and statistical precision on the 59m flight path of LANSCE, the pulsed spallation neutron source at Los Alamos National Laboratory. Measurements on all three materials were made over a range of temperatures from 15K to 300K, with small steps through the critical temperature region near 27K. No significant changes in the mean phonon energy of the lanthanum atoms were observed near the critical temperature of the super-conducting material. It appears however that the mean phonon energy of lanthanum in the superconductor is considerably higher than that in the non-superconductors. The samples used in this series of experiments were too thin in barium and copper to determine anything significant about their phonon spectra.
Energy Technology Data Exchange (ETDEWEB)
Galahom, Ahmed Abdelghafar [Higher Technological Institute, Ramadan (Egypt)
2016-06-15
The present work discusses two different models of boiling water reactor (BWR) bundle to compare the neutronic characteristics of uranium dioxide (UO{sub 2}) and uranium zirconium hydride (UZrH{sub 1.6}) fuel. Each bundle consists of four assemblies. The BWR assembly fueled with UO{sub 2} contains 8 × 8 fuel rods while that fueled with UZrH{sub 1.6} contains 9 × 9 fuel rods. The Monte Carlo N-Particle Transport code, based on the Mont Carlo method, is used to design three dimensional models for BWR fuel bundles at typical operating temperatures and pressure conditions. These models are used to determine the multiplication factor, pin-by-pin power distribution, axial power distribution, thermal neutron flux distribution, and axial thermal neutron flux. The moderator and coolant (water) are permitted to boil within the BWR core forming steam bubbles, so it is important to calculate the reactivity effect of voiding at different values. It is found that the hydride fuel bundle design can be simplified by eliminating water rods and replacing the control blade with control rods. UZrH{sub 1.6} fuel improves the performance of the BWR in different ways such as increasing the energy extracted per fuel assembly, reducing the uranium ore, and reducing the plutonium accumulated in the BWR through burnup.
International Nuclear Information System (INIS)
Zizin, M.N.; Savochkina, O.A.; Chukhlova, O.P.
1978-01-01
A structure of standard designations is described and semantics of a number of standard values used in a NF-6 program complex is given. Main source data and results of neutron-physical reactor calculation are standard values, the peculiarities of FORTRAN and ALGOL-GDR algorithm languages in the DUBNA monitoring system were taken account of. As a base of standard values list the FIHAR system list, supplemented with new standard designations for integral reactor characteristics, is used. Developed is also a list of standard values to organize the exchange with external memory in the process of task solution and long-range storage
International Nuclear Information System (INIS)
Nakashima, Hideki; Nakao, Yasuyuki; Ohta, Masao
1978-01-01
An examination is made of the dependence shown by the nuclear characteristics of the blanket and shield of D-D fusion reactors on S sub( d d)/S sub( d t), the ratio between the 2.45 MeV neutrons resulting from the D-D reaction and those of 14.06 MeV from the D-T reaction. Also, an estimate is presented of this neutron source ratio S sub( d d)/S sub( d t) for the case of D-D reactors, taken as an example. It is shown that an increase of S sub( d d)/S sub( d t) reduces the amount of nuclear heating per unit source neutron, while at the same time improving the shielding characteristics. This is accountable to lowering of the energy and penetrability of incident neutrons into the blanket brought about by the increase of S sub( d d)/S sub( d t). The value of S sub( d d)/S sub( d t) in a steady state D-D fusioning plasma core is estimated to be 1.46 -- 1.72 for an ion temperature ranging from 60 -- 180 keV. The reductions obtained on H sub( t)sup( b) (total heating in the blanket), H sub( t)sup( m g)/H sub( t)sup( b) (shielding indicator = ratio between total heating in superconducting magnet and that in the blanket) and phi sup( m g)/phi sup( w) (ratio of fast neutron fluxes between that at the magnet inner surface and that at the first wall inner surface) brought about by increasing S sub( d d)/S sub( d t) from unity to the value cited above do not differ to any appreciable extent, whichever is adopted among the design models considered here, the differences being at most about 10, 15 and 25%, respectively, for these three parameters. These results would broaden the validity of the conclusion derived in the previous paper for the case of S sub( d d)/S sub( d t) = 1.0, that the blanket-shield concept would appear to be the most suitable for D-D fusion reactors. (author)
Exciton luminescence characteristic of ZnO. Ga scintillator for neutron detection
International Nuclear Information System (INIS)
Kinoshita, A.; Fujiwara, A.; Koyama, S.; Takei, Y.; Nanto, H.; Katagiri, Masaki
2008-01-01
ZnO family phosphors as novel phosphor materials for neutron detector have prepared using Spark Plasma Sintering method. The optical properties of ZnO phosphor prepared are investigated. The following results were obtained. Two dominant PL emission peaks at 395 nm and 495 nm were observed. The lifetime of the PL emission peak at 395 nm (UV emission band) is about 20 ns, which is suitable for neutron detection. The Ga (30 mol%)-doped ZnO phosphor exhibited an intense UV emission band without the visible emission band. The Ga-doped ZnO phosphor can be prepared at the atmospheric pressure of about 8 Pa using SPS method. It was found that the PL intensity of UV emission band is increased with improving the crystallinity of the ZnO phosphor. (author)
Neutron-physical characteristics of the TVRM-100 reactor with ten ring fuel channels
International Nuclear Information System (INIS)
Mikhajlov, V.M.; Myrtsymova, L.A.
1988-01-01
Three-dimensional heterogeneous calculations of TVRM-100 reactor which is a research reactor using enriched fuel with heavy-water moderator, coolant and reflector, are conducted. Achievable burnup depths depending on the number of removable FAs are presented. The maximum non-perturbed thermal neutron flux in the reflector is (2-1.8)x10 15 cm -2 c -1 ; mean flux on the fuel is 2.9x10 14 cm -2 c -1 . Energy release radial non-uniformity is 0.67, maximum bending by FA is ∼3.7. Reactivity temperature effect is negative and is equal to - 0.9x10 -4 grad -1 without accounting for experimental channels. Control rod efficiency in the radial reflector is high, but their location dose to experimental devices in the high neutron flux area is undesirable. 4 refs.; 5 figs
International Nuclear Information System (INIS)
Kataoka, H.
1975-01-01
The 'not-linear' or '2nd-class-nonlinear' characteristics in nuclear reactor kinetics with the feedback effect in the high-power operation and induce the increase in the amplitude of the neutron flux noise, specially in the very low frequency region. The fundamental behaviour of 'not-linear' characteristics and its effect for the reactor noise was investigated. Application of the reactor noise analysis technique to power reactors has not been successful because of unknown large disagreement between the result of the conventional theoretical analysis and the experimental facts. When the cause of this discrepancy is clear, reactor noise analysis techniques can be effectively applied to instrumentation, control, monitoring and diagnosis of power reactors. (author)
Stephan, Andrew C [Knoxville, TN; Jardret,; Vincent, D [Powell, TN
2011-04-05
A neutron detector has a volume of neutron moderating material and a plurality of individual neutron sensing elements dispersed at selected locations throughout the moderator, and particularly arranged so that some of the detecting elements are closer to the surface of the moderator assembly and others are more deeply embedded. The arrangement captures some thermalized neutrons that might otherwise be scattered away from a single, centrally located detector element. Different geometrical arrangements may be used while preserving its fundamental characteristics. Different types of neutron sensing elements may be used, which may operate on any of a number of physical principles to perform the function of sensing a neutron, either by a capture or a scattering reaction, and converting that reaction to a detectable signal. High detection efficiency, an ability to acquire spectral information, and directional sensitivity may be obtained.
Wu, Yican
2017-01-01
This book provides a systematic and comprehensive introduction to fusion neutronics, covering all key topics from the fundamental theories and methodologies, as well as a wide range of fusion system designs and experiments. It is the first-ever book focusing on the subject of fusion neutronics research. Compared with other nuclear devices such as fission reactors and accelerators, fusion systems are normally characterized by their complex geometry and nuclear physics, which entail new challenges for neutronics such as complicated modeling, deep penetration, low simulation efficiency, multi-physics coupling, etc. The book focuses on the neutronics characteristics of fusion systems and introduces a series of theories and methodologies that were developed to address the challenges of fusion neutronics, and which have since been widely applied all over the world. Further, it introduces readers to neutronics design’s unique principles and procedures, experimental methodologies and technologies for fusion systems...
International Nuclear Information System (INIS)
Hiraoka, Eiichi
1988-01-01
The thermal neutron absorption coefficient is essentially different from the X-ray absorption coefficient. Each substance has a characteristic absorption coefficient regardless of its density. Neutron deams have the following features: (1) neutrons are not transmitted efficiently by low molecular weight substances, (2) they are transmitted efficiently by heavy metals, and (3) the transmittance differs among isotopes. Thus, neutron beams are suitable for cheking for foreign matters in heavy metals and testing of composites consisting of both heavy and light materials. A neutron source generates fast neutrons, which should be converted into thermal neutrons by reducing their energy. Major neutron souces include nuclear reactors, radioisotopes and particle accelerators. Photographic films and television systems are mainly used to observe neutron transmission images. Computers are employed for image processing, computerized tomography and three-dimensional analysis. The major applications of neutron radiography include inspection of neclear fuel; evaluation of material for airplane; observation of fuel in the engine and oil in the hydraulic systems in airplanes; testing of composite materials; etc. (Nogami, K.)
Method and apparatus for measuring neutron characteristics of material surrounding a borehole
International Nuclear Information System (INIS)
Randall, R.R.
1983-01-01
A method for logging the formations surrounding an earth borehole, comprising: pulsedly irradiating said formations with discrete bursts form a source of high energy neutrons; detecting radiations emanating from said irradiated formations at a detector spaced from said source; measuring said detected radiations during first and second time intervals following a burst from said source; generating time-separated electrical pulses indicative of said detected radiations during a third fixed time duration interval following a subsequent burst from said sources; and controlling automatically the starting time of said third measurement interval in accordance with the said measurement from said first and second time intervals
Directory of Open Access Journals (Sweden)
Susan Abadi
2016-09-01
Full Text Available Speech delay with an unknown cause is a problem among children. This diagnosis is the last differential diagnosis after observing normal findings in routine hearing tests. The present study was undertaken to determine whether auditory brainstem responses to click stimuli are different between normally developing children and children suffering from delayed speech with unknown causes. In this cross-sectional study, we compared click auditory brainstem responses between 261 children who were clinically diagnosed with delayed speech with unknown causes based on normal routine auditory test findings and neurological examinations and had >12 months of speech delay (case group and 261 age- and sex-matched normally developing children (control group. Our results indicated that the case group exhibited significantly higher wave amplitude responses to click stimuli (waves I, III, and V than did the control group (P=0.001. These amplitudes were significantly reduced after 1 year (P=0.001; however, they were still significantly higher than those of the control group (P=0.001. The significant differences were seen regardless of the age and the sex of the participants. There were no statistically significant differences between the 2 groups considering the latency of waves I, III, and V. In conclusion, the higher amplitudes of waves I, III, and V, which were observed in the auditory brainstem responses to click stimuli among the patients with speech delay with unknown causes, might be used as a diagnostic tool to track patients’ improvement after treatment.
Superplastic characteristics and microstructure of neutron irradiated 3Y-TZP
International Nuclear Information System (INIS)
Shibata, Taiju; Motohashi, Yoshinobu; Ishihara, Masahiro; Baba, Shinichi; Sawa, Kazuhiro
2006-01-01
Fast neutrons (energy > 1.6 x 10 -13 J) were irradiated to 3Y-TZP specimens, typical superplastic ceramics, at the fluence of 2.5 x 10 24 and 4.3 x 10 24 m -2 at JMTR of JAEA. The Vickers hardness with indentation load of 4.9 and 9.8 N at room temperature was seemed to be slightly increased by the irradiation. Through the superplastic tensile tests in a temperature range from 1623 to 1773 K with initial strain rates of 5.0 x 10 -4 and 1.0 x 10 -3 s -1 , it was found that the superplastic flow stress is decreased with increasing the neutron fluence. The microstructural features of the fractured specimens were observed by a SEM. It implies that the grain boundary microstructure of the irradiated specimens would be changed by annealing in the superplastic tests are elevated temperatures. It is quite probable that the irradiation-induced vacancy clusters might play an important role to weaken the grain boundary cohesion which may be an important factor to determine the superplastic properties, and hence they would decrease the superplastic flow stress. (author)
International Nuclear Information System (INIS)
Kim, M.G.
1980-01-01
The multigroup neutron diffusion equations with the constraint of specified power distributions are investigated by the application of the straight-line method which can be considered as the limiting case of zero mesh space in the finite difference method. The standard partial differential form of the diffusion equation is reduced to sets of ordinary differential equations and then converted into sets of integral equations by using Green's functions defined on the pseudo straight lines. Coupling of each straight line to the adjacent lines arises from the application of a three-point central difference formula. The interfaces encountered between two regions are taken into account by imposing the continuity conditions for the grown fluxes and net currents with Taylor expansions of internal fluxes at the interface positions. A few sample problems are selected to test the validity of the method. It is found that the proposed method of solution is similar to the finite Fourier sine transform. Numerical results for the solutions obtained by the method of straight lines are compared with the results of the exact analytical solutions for simple geometries. These comparisons indicate that the proposed method is compatible with the analytical method, and in some problems considered the straight-line solutions are much more efficient than the exact solutions. The method is also extended to the reactor kinetics problem by expressing the kinetics parameters in terms of the basis functions which are used to obtain the solutions of the steady-state neutron diffusion equations
Neutronics characteristics of micro-heterogeneous ThO2-UO2 PWR cores
International Nuclear Information System (INIS)
Zhao, X.; Driscoll, M.J.; Kazimi, S.
2001-01-01
A new fuel concept, axially-micro-heterogeneous ThO 2 -UO 2 fuel, where ThO 2 fuel pellets and UO 2 fuel pellets are stacked in separate layers in the fuel rods, is being studied at MIT as an option to reduce plutonium production in LWR fuel. Very interesting neutronic behavior is observed: (1) A reactivity increase of 3% to 4% at EOL for a given 235 U inventory which results in a 20-30% increase in average core discharge burnup; (2) For certain configurations, a ''burnable poison'' effect is observed. Analysis shows that these effects are achieved due to a combination of changes in self-shielding, local fissile worth, and conversion ratio, among which self-shielding is the dominant effect at the end of a reactivity-limited burnup. Other variations of micro-heterogeneous UO 2 -ThO 2 fuel including duplex pellets, checkerboard pin distribution, and checkerboard-axial combinations have also been investigated, and their neutronic performance compared. It is concluded that the axial fuel micro-heterogeneity provides the largest gain in reactivity-limited burnup. (author)
International Nuclear Information System (INIS)
Sohrabi, M.; Khoshnoodi, M.
1986-01-01
The effects of pre-etching time (PET) or duration of etching of fast-neutron-induced-recoil tracks in CR-39 in 6N KOH at 60 0 C on electrochemical etching neutron characteristic responses; i.e. sensitivity and mean recoil track diameter (MRTD) versus KOH normality up to 18N are investigated in this paper. Six sets of responses for PETs of 0, 1, 2, 3, 4, and 5 hours were obtained by using our new multi-chamber ECE (MCECE) system which reduced total operation time to about 6% of the time usually required when single-chamber ECE systems are used. The sensitivity response for zero PET showed a broad plateau and a high sensitivity low-LET peak around 16N. By increasing PET, another peak was also developed around 5N leading to 'double-humped' responses with two maximums around 5N and 16N, and a minimum around 11N. On the other hand, the MRTD responses for all PETs studied showed the same general trend with maximums around 11N. In this paper, shape of tracks under different conditions are also investigated, new optimum conditions such as KOH concentrations of 5, 11, and 15N at 25 0 C, with or without pre-etching, are recommended for tracks of lower-LET recoils including possibly protons, and alpha particle tracks over a broad energy range, and the efficiency of the MCECE system is also demonstrated. (author)
International Nuclear Information System (INIS)
Qin Xue; Han Jifeng; Yang Chaowen
2014-01-01
The characteristic γ-ray spectrum of TNT under soil induced by DT neutron is measured based on the PFTNA demining system. GEANT4 Monte Carlo simulation toolkit is used to simulate the whole experimental procedure. The simulative spectrum is compared with the experimental spectrum. The result shows that they are mainly consistent. It is for the first time to analyze the spectrum by Monte Carlo simulation, the share of the background sources such as neutron, gamma are obtained, the contribution that the experimental apparatus such as shielding, detector sleeve, moderator make to the background is analysed. The study found that the effective gamma signal (from soil and TNT) is only 29% of the full-spectrum signal, and the background signal is more than 68% of the full-spectrum signal, which is mainly produced in the shielding and the detector sleeve. The simulation result shows that by gradually improving the shielding and the cadmium of the detector sleeve, the share of the effective gamma signal can increase to 36% and the background signal can fell 7% eventually. (authors)
Energy Technology Data Exchange (ETDEWEB)
Ganich, P P; Parlag, O A; Sikora, D I; Sychev, S I
1989-03-01
Relation between yields and cross sections of photofission and photoproduction is studied in order to use them in the methods for analysis of fissile nuclides. Total yield of delayed neutrons from the {sup 232}Th target and ratios of total yields from {sup 238}U and {sup 232}Th targets were measured in the M=300 microtron in 6-18 MeV energy range. Efficiency of the suggested method for refining the {sup 238}U photofission cross sections in the range of E1-giant resonance is shown.
Characteristics of the quarry as shielding for "2"4"1AmBe neutrons and monoenergetic photons
International Nuclear Information System (INIS)
Vega C, H. R.; Hernandez D, V. M.; Letechipia de L, C.; Salas L, M. A.; Rodriguez R, J. A.; Juarez A, C. A.
2016-09-01
Shielding is an important element in radiation protection since allows the management of radiation sources. Currently there are different materials of natural or anthropogenic origin that are used as shielding for both photons and neutrons. The quarry is a material of natural origin and abundant in our country, which is used in construction or for the manufacture of sculptures, however its characteristics as shielding have not been reported. In this paper we report some of the properties of the quarry as shielding for monoenergetic photons and for neutrons produced by an isotopic neutron source of "2"4"1AmBe. A quarry piece was used to determine its density and its chemical composition, with the XCOM code the elemental composition was determined and the mass interaction and total attenuation coefficients of the quarry were determined with photons of 10"-"3 to 10"-"5 MeV; the interaction coefficients included coherent dispersion, photoelectric absorption, Compton dispersion and the production of pairs in the nuclear and electronic field. Using the MCNP5 code, a narrow geometry attenuation experiment was modeled and the photon fluence was estimated that reaches a point detector at a distance of 42 cm from a point source, isotropic and monoenergetic photon when the source and the point detector were added quarry pieces of different thicknesses. The reduction of the number of photons as a function of the thickness of the quarry was used to determine the coefficient of linear attenuation of the quarry before photons of 0.03, 0.07, 0.1, 0.3, 1, 2 and 3 MeV that were the same as those calculated with the XCOM code. With the MCNP, the K a and H(10) transmission curves were also calculated. This same model was used to determined the variation of the "2"4"1AmBe neutron spectrum as a function of quarry thickness, as well as the E_R_O_T and H(10) transmission curves. (Author)
International Nuclear Information System (INIS)
Akie, Hiroshi; Nakano, Yoshihiro; Okubo, Tsutomu
2009-01-01
Innovative Water Reactor for Flexible Fuel Cycle (FLWR) is a low moderation type boiling water reactor which can realize plutonium multiple recycling and breeding. For the introduction stage of FLWR, a high conversion (HC) type FLWR is proposed to keep technical continuity from current light water reactors. The HC core of FLWR has a less tight fuel lattice with lower coolant void fraction than the breeder (BR) type core. The HC type FLWR core is to be shifted to the BR core by only replacing the fuel assemblies of the same outer shape and size in the same reactor system. In the HC to BR transition phase of FLWR, there exist both types of fuel assemblies in the same core configuration. In the HC assembly, neutron spectrum is softer than in the BR assembly, and the axial fuel and blanket arrangement is different from the BR assembly. Due to these differences, there might appear a power peaking in the adjacent region between HC and BR assemblies. The power distribution in the HC + BR assemblies mixed core configuration is studied by performing assembly calculations and core calculations on a few assemblies local geometry and the whole core geometry. As a result, although a power peaking can be locally very large in the HC and BR assemblies adjacent regions, such local power peakings are shown to be effectively reduced by considering a rod-wise fuel enrichment distribution. In the whole core calculation, it seems possible to optimize the fuel assembly loading and shuffling pattern to avoid large power level mismatch between the assemblies. It is expected that FLWR can be shifted from HC type to BR type without major neutronic difficulties. (author)
Problems in the neutron dynamics of source-driven systems
International Nuclear Information System (INIS)
Ravetto, P.
2001-01-01
The present paper presents some neutronic features of source-driven neutron multiplying systems, with special regards to dynamics, discussing the validity and limitations of classical methods, developed for systems in the vicinity of criticality. Specific characteristics, such as source dominance and the role of delayed neutron emissions are illustrated. Some dynamic peculiarities of innovative concepts proposed for accelerator-driven systems, such as fluid-fuel, are also discussed. The second portion of the work formulates the quasi-static methods for source-driven systems, evidencing its novel features and presenting some numerical results. (author)
Gresnigt, Marco M.M.; Cune, Marco S.; de Roos, Joanne G.; Özcan, Mutlu
OBJECTIVES: Adhesion on dentin is less reliable than on enamel, which could affect the durability of laminate veneers (LV). Immediate dentin sealing (IDS) is suggested instead of delayed dentin sealing (DDS) to overcome hypersensitivity and prevent debonding from dentin. This study evaluated the
Bradley, D.; Kalghatgi, G.T.
2009-01-01
The functional relationship of autoignition delay time with temperature and pressure is employed to derive the propagation velocities of autoignitive reaction fronts for particular reactivity gradients, once autoignition has been initiated. In the present study of a variety of premixtures, with
Energy Technology Data Exchange (ETDEWEB)
Proskuryakov, K.N.; Bogomazov, D.N.; Poliakov, N. [Moscow Power Engineering Institute (Technical University), Moscow (Russian Federation)
2007-07-01
The new special module to neutron-physic and thermal-hydraulic computer codes for coolant acoustical characteristics calculation is worked out. The Russian computer code Rainbow has been selected for joint use with a developed module. This code system provides the possibility of EFOCP (Eigen Frequencies of Oscillations of the Coolant Pressure) calculations in any coolant acoustical elements of primary circuits of NPP. EFOCP values have been calculated for transient and for stationary operating. The calculated results for nominal operating were compared with results of measured EFOCP. For example, this comparison was provided for the system: 'pressurizer + surge line' of a WWER-1000 reactor. The calculated result 0.58 Hz practically coincides with the result of measurement (0.6 Hz). The EFOCP variations in transients are also shown. The presented results are intended to be useful for NPP vibration-acoustical certification. There are no serious difficulties for using this module with other computer codes.
International Nuclear Information System (INIS)
Proskuryakov, K.N.; Bogomazov, D.N.; Poliakov, N.
2007-01-01
The new special module to neutron-physic and thermal-hydraulic computer codes for coolant acoustical characteristics calculation is worked out. The Russian computer code Rainbow has been selected for joint use with a developed module. This code system provides the possibility of EFOCP (Eigen Frequencies of Oscillations of the Coolant Pressure) calculations in any coolant acoustical elements of primary circuits of NPP. EFOCP values have been calculated for transient and for stationary operating. The calculated results for nominal operating were compared with results of measured EFOCP. For example, this comparison was provided for the system: 'pressurizer + surge line' of a WWER-1000 reactor. The calculated result 0.58 Hz practically coincides with the result of measurement (0.6 Hz). The EFOCP variations in transients are also shown. The presented results are intended to be useful for NPP vibration-acoustical certification. There are no serious difficulties for using this module with other computer codes
International Nuclear Information System (INIS)
Tuseev, T.; Aksenova, T.I.; Berdauletov, A.K.
2000-01-01
In this work the results of comparative investigation of adsorption curves versus REM order number (La, Ce, Nd, Pr, Gd) and dose neutron irradiation are presented. It is discovered that the original REM oxides have high adsorption capacity both for donor gases and for acceptor ones. The adsorption capacity for oxygen and hydrogen becomes lower when the REM order number is higher, but it is contrary for water molecules. The obtained results showed that adsorption properties of REM oxides were changing on identical laws. But oxides of metals, having anomalous properties (variable valency, high cross-section capture) revealed especial adsorption properties under irradiation. In the time of consideration of possible mechanism of radiation - stimulated gas adsorption on oxide surface it is necessary to take in attention both the formation of radiation defects in crystal lattice and the characteristics of electron structure and presence of 4 f - cover in rare-earth metals
International Nuclear Information System (INIS)
Adamovich, L.A.; Azarov, V.V.
1999-01-01
Time dependent core power behavior in a nuclear reactor is described with well-known neutron kinetics equations. At the same time, two portions are distinguished in energy released from uranium nuclei fission; one released directly at fission and another delayed (residual) portion produced during radioactive decay of fission products. While prompt power is definitely described with kinetics equations, the delayed power presentation still remains outstanding. Since in operation the delayed power part is relatively small (about 6%) operation, it can be neglected for small reactivity disturbances assuming that entire power obeys neutron kinetics equations. In case of a high negative reactivity rapidly inserted in core (e.g. reactor scram initiation) the prompt and delayed components can be calculated separately with practically no impact on each other, employing kinetics equations for prompt power and known approximation formulas for delayed portion, named residual in this specific case. Under substantial disturbances the prompt component in the dynamic process becomes commensurable with delayed portion, thus making necessary to take into account their cross impact. A system of differential equations to describe time-dependent behavior of delayed power is presented. Specific NPP analysis shows a way to significantly simplify the task formulation. (author)
Energy Technology Data Exchange (ETDEWEB)
Fang, Jiayang; Yu, Dexin; Ma, Xiangxing; Xiang, Ying; Guo, Lijuan [Qilu Hospital of Shandong University, Department of Radiology, Ji' nan City, Shandong (China); Ma, Xiaoyuan [Qilu Hospital of Shandong University, Department of Orthopaedics, Ji' nan City (China)
2017-09-15
To obtain specific imaging findings of solitary necrotic nodule of the liver (SNNL) using longer delayed contrast-enhanced MRI and compare them with those of three mimic hepatic diseases. Sixteen patients with SNNL underwent plain and contrast-enhanced triphasic CT and multiphasic MRI with delayed time prolonged to 2 h after contrast bolus injection. Twenty-three patients with mimic lesions including seven with eight HCCs, five with five iCCs and 11 with metastatic lesions served as the control group. Those patients also received plain and multiphasic contrast-enhanced MRI. Imaging features of lesions such as peripheral wash-out time were evaluated. Among the 16 SNNLs, with a prolonged delayed MRI time, the enhancement degree of tumour periphery increased gradually. When it was up to 1 h, all lesions represented moderate/marked peripheral enhancement with internal hypointensity. However, the peripheral wash-out in seven HCCs (87.5%) and all metastatic lesions except three appeared at 10 or 15 min, one iCC (20%) at 30 min and the other lesions at 1 h. Longer MRI with a delayed time of 1-2 h may be useful in diagnosis SNNL, revealing the specific imaging characteristic of SNNL as pronounced peripheral enhancement with internal hypointensity. (orig.)
Characteristic Elemental Composition of Oil Pigments using Instrumental Neutron Activation Analysis
International Nuclear Information System (INIS)
Chung, Yong Sam; Kim, Sun Ha; Sun, Gwang Min; Lim, Jong Myung; Moon, Jong Hwa; Kim, Young Jin; Lim, Sung Jin; Song, Yu Na; Kim, Ken
2009-01-01
The principal aim of this study is to identify the applicability of instrumental neutron activation analysis as a non-destructive examination tool for the quantitative composition analysis associated with authentication, restoration, and conservation of art objects in the field of cultural heritage. Generally, the chemical composition of pigments are associated with the colors such as white, yellow, orange, red, green, blue and black, and it varies with raw materials of pigments. According to the colors of a different pigments, chemical compositions are as follows; for example, white pigments were used for a mixture of Pb(CO 3 ) 2 , PbSO 4 , PbO, Pb(OH) 2 , ZnO, ZnS, TiO 2 , BaSO 4 , CaCO 3 , Al 2 O 3 , As 2 S 3 , etc.; black pigments were series of carbon black, borne ash, MnO+Mn 2 O 3 , etc.; red pigments were Fe 2 O 3 , Pb 3 O 4 , HgS, PbMo 4 , CdS+CdSe, etc.; brown and yellow pigments were PbCrO 4 , ZnCrO 4 , CdS-ZnS, K 3 [Co(NO 2 ) 6 ], Pb(SbO 3 ) 2 , C 19 H 16 O 11 Mg, SrCrO 4 , etc.; green pigments were Cr 2 O 3 , Cr 2 O(OH) 4 , Cu(C 2 H 3 O 2 )-2Cu(OH) 2 ), Cr 2 O 3 -Al 2 O 3 -CoO, etc.; blue pigments were Fe 4 [Fe(CN) 6 ] 3 , CoO-Al 2 O 3 , Na 8 - 10 Al 6 Si 6 O 24 S 2-4 , etc. This first step is to obtain quantitative data on the concentrations of major, minor and trace elements in oil pigments and to explain pigment sources by statistical treatment as reported in many literatures. The determination of major, minor and micro elements in the subject materials are essential in many fields of basic science and technology as well as commercial and industrial fields. In particular, direct analysis of a sample offers a more effective investigation method in these fields. Instrumental neutron activation analysis (INAA) has an inherent advantage of being a non-destructive, simultaneously multi-elemental analysis with high accuracy and sensitivity. In order to characterize the elemental contents of art objects, the quantitative analysis of oil pigment series was
Neutron shielding and constructional characteristics of a new type concrete and from borated clinker
International Nuclear Information System (INIS)
Cakaloz, T.
1979-07-01
A boron containing cement, which can be used as nuclear shielding material, is produced at pilot plant scale applying two different methods. In the first method, the raw mixture of a normal portland cement is mixed with pre-calcined colemanite, a calcium borate mineral, and clinkerized in a rotary kiln (borated-clinker). In the second method, the colemanite is mixed with an admixture, which contains mainly limestone and marl, and burnt in the rotary kiln to obtain a borated-lime composite. The borated-lime composite is then added to the normal portland cement clinker up to 2% B 2 O 3 content for shielding purpose. The results have shown that the borated-clinker contained untolerable amount of free lime resulting in a decrease in compressive strength. The addition of the borated-lime composite to the normal portland cement clinker up to 1% B 2 O 3 content did not alter the setting time and the volume expansion properties. The reduction in the compressive strength was found to be tolerable, however, the decrease in the bending strength was 20% lower than that of permissible value. On the other hand, the increase in B 2 O 3 content of the mixture improved the neutron absorptivity resulting in an increase in total cross-section about 7 times for 1% B 2 O 3 without changing the gamma absorption value
International Nuclear Information System (INIS)
Chen Sen; Wu Yican; Jin Ming; Chen Zhibin; Bai Yunqing; Zhao Zhumin
2014-01-01
Accelerator Driven Sub-critical System (ADS) has particular neutronics behaviors compared with the critical system. Prompt jump approximation point reactor kinetics equations taken external source into account have been deduced using an approach of prompt jump approximation. And the relationship between injection reactivity and power ampliation has been achieved. In addition, based on the RELAP5 code the prolong development of point reactor kinetics code used into assessing sub-critical system have been promoted. Different sub-criticality (k eff = 0.90, 0.95, 0.97, 0.98 and 0.99) have been assessed in preliminary design of a type of natural circulation cooling sub-critical reactor under conditions of reactivity injection +1 β in one second. It shows that the external source prompt transient approximation method has an accurate solution after injecting reactivity around short time and has a capacity to solve the dynamic equation, and the sub-critical system has an inner stability while the deeper sub-criticality the less impact on the sub-critical system. (authors)
Characteristic Elemental Composition of Oil Pigments using Instrumental Neutron Activation Analysis
Energy Technology Data Exchange (ETDEWEB)
Chung, Yong Sam; Kim, Sun Ha; Sun, Gwang Min; Lim, Jong Myung; Moon, Jong Hwa; Kim, Young Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lim, Sung Jin; Song, Yu Na; Kim, Ken [National Museum of Contemporary Art, Gwacheon (Korea, Republic of)
2009-10-15
The principal aim of this study is to identify the applicability of instrumental neutron activation analysis as a non-destructive examination tool for the quantitative composition analysis associated with authentication, restoration, and conservation of art objects in the field of cultural heritage. Generally, the chemical composition of pigments are associated with the colors such as white, yellow, orange, red, green, blue and black, and it varies with raw materials of pigments. According to the colors of a different pigments, chemical compositions are as follows; for example, white pigments were used for a mixture of Pb(CO{sub 3}){sub 2}, PbSO{sub 4}, PbO, Pb(OH){sub 2}, ZnO, ZnS, TiO{sub 2}, BaSO{sub 4}, CaCO{sub 3}, Al{sub 2}O{sub 3}, As{sub 2}S{sub 3}, etc.; black pigments were series of carbon black, borne ash, MnO+Mn{sub 2}O{sub 3}, etc.; red pigments were Fe{sub 2}O{sub 3}, Pb{sub 3}O{sub 4}, HgS, PbMo{sub 4}, CdS+CdSe, etc.; brown and yellow pigments were PbCrO{sub 4}, ZnCrO{sub 4}, CdS-ZnS, K{sub 3}[Co(NO{sub 2}){sub 6}], Pb(SbO{sub 3}){sub 2}, C{sub 19}H{sub 16}O{sub 11}Mg, SrCrO{sub 4}, etc.; green pigments were Cr{sub 2}O{sub 3}, Cr{sub 2}O(OH){sub 4}, Cu(C{sub 2}H{sub 3}O{sub 2})-2Cu(OH){sub 2}), Cr{sub 2}O{sub 3}-Al{sub 2}O{sub 3}-CoO, etc.; blue pigments were Fe{sub 4}[Fe(CN){sub 6}]{sub 3}, CoO-Al{sub 2}O{sub 3}, Na{sub 8}-{sub 10}Al{sub 6}Si{sub 6}O{sub 24}S{sub 2-4}, etc. This first step is to obtain quantitative data on the concentrations of major, minor and trace elements in oil pigments and to explain pigment sources by statistical treatment as reported in many literatures. The determination of major, minor and micro elements in the subject materials are essential in many fields of basic science and technology as well as commercial and industrial fields. In particular, direct analysis of a sample offers a more effective investigation method in these fields. Instrumental neutron activation analysis (INAA) has an inherent advantage of being a
International Nuclear Information System (INIS)
Kosaka, Shinya; Saji, Etsuro
2000-01-01
A characteristics transport theory code, CHAPLET, has been developed for the purpose of making it practical to perform a whole LWR core calculation with the same level of calculational model and accuracy as that of an ordinary single assembly calculation. The characteristics routine employs the CACTUS algorithm for drawing ray tracing lines, which assists the two key features of the flux solution in the CHAPLET code. One is the direct neutron path linking (DNPL) technique which strictly connects angular fluxes at each assembly interface in the flux solution separated between assemblies. Another is to reduce the required memory storage by sharing the data related to ray tracing among assemblies with the same configuration. For faster computation, the coarse mesh rebalance (CMR) method and the Aitken method were incorporated in the code and the combined use of both methods showed the most promising acceleration performance among the trials. In addition, the parallelization of the flux solution was attempted, resulting in a significant reduction in the wall-clock time of the calculation. By all these efforts, coupled with the results of many verification studies, a whole LWR core heterogeneous transport theory calculation finally became practical. CHAPLET is thought to be a useful tool which can produce the reference solutions for analyses of an LWR (author)
International Nuclear Information System (INIS)
Field, Kevin G.; Yang, Ying; Allen, Todd R.; Busby, Jeremy T.
2015-01-01
Radiation induced segregation (RIS) is a well-studied phenomena which occurs in many structurally relevant nuclear materials including austenitic stainless steels. RIS occurs due to solute atoms preferentially coupling with mobile point defect fluxes that migrate and interact with defect sinks. Here, a 304 stainless steel was neutron irradiated up to 47.1 dpa at 320 °C. Investigations into the RIS response at specific grain boundary types were used to determine the sink characteristics of different boundary types as a function of irradiation dose. A rate theory model built on the foundation of the modified inverse Kirkendall (MIK) model is proposed and benchmarked to the experimental results. This model, termed the GiMIK model, includes alterations in the boundary conditions based on grain boundary structure and expressions for interstitial binding. This investigation, through experiment and modeling, found specific grain boundary structures exhibiting unique defect sink characteristics depending on their local structure. Such interactions were found to be consistent across all doses investigated and to have larger global implications, including precipitation of Ni–Si clusters near different grain boundary types
Directory of Open Access Journals (Sweden)
O. O. Bello
2006-09-01
Full Text Available The thermal upgrading of Nigerian petroleum residue was studied at relatively low pressure in a delayed coking reactor system. In this work, the intent was to investigate the effects of process variables such as reaction temperature (200(0C to 600(0C, reaction time (0 to 120min, additive concentration loading and additive-to-residue ratio on the amounts and quality of organic liquid product (OLP. The liquid products derived from the delayed coking process were characterized by means of instrumental analysis of gas-liquid chromatography. Results obtained from the analyses of the OLP revealed an upward trend of the conversion process and the selectivity of the aromatic compounds with additive-to-residue ratio (ARR and increase in temperature. This led to maximum yield of 37.2% achieved with ARR of 5 compared to 31% achieved with ordinary thermal conversion. The selectivity for aromatic hydrocarbons was maximum at 83.1wt% the selectivity towards aromatics and aliphatic hydrocarbons were highest for methanol-potassium hydroxide and methanol respectively. In all additive system cases, maximum OLP was produced at an optimum reaction temperature of 370(0C in the delayed coking reactor and at higher residence time. The gaseous product consisted of carbon monoxide and carbon dioxide and C1 - C6 hydrocarbons, which amounted to about 20 to 30 wt% of liquid distillate. The information obtained in this study show that the organic liquid products are amenable to characterization procedure and provided the basis for the identification of processes for upgrading Nigerian petroleum residue and such other starting materials such as bitumen or fossil fuel coal liquids.
International Nuclear Information System (INIS)
Masiello, E.
2006-01-01
The principal goal of this manuscript is devoted to the investigation of a new type of heterogeneous mesh adapted to the shape of the fuel pins (fuel-clad-moderator). The new heterogeneous mesh guarantees the spatial modelling of the pin-cell with a minimum of regions. Two methods are investigated for the spatial discretization of the transport equation: the discontinuous finite element method and the method of characteristics for structured cells. These methods together with the new representation of the pin-cell result in an appreciable reduction of calculation points. They allow an exact modelling of the fuel pin-cell without spatial homogenization. A new synthetic acceleration technique based on an angular multigrid is also presented for the speed up of the inner iterations. These methods are good candidates for transport calculations for a nuclear reactor core. A second objective of this work is the application of method of characteristics for non-structured geometries to the study of double heterogeneity problem. The letters is characterized by fuel material with a stochastic dispersion of heterogeneous grains, and until now was solved with a model based on collision probabilities. We propose a new statistical model based on renewal-Markovian theory, which makes possible to take into account the stochastic nature of the problem and to avoid the approximations of the collision probability model. The numerical solution of this model is guaranteed by the method of characteristics. (author)
International Nuclear Information System (INIS)
Zhang Yongde.
1987-03-01
In this paper, the neutron Dirac-equation is presented. After decoupling it into two equations of the simple spinors, the rigorous solution of this equation is obtained in the case of slab-like uniform magnetic fields at perpendicular incidence. At non-relativistic approximation and first order approximation of weak field (NRWFA), our results have included all results that have been obtained in references for this case up to now. The corresponding transformations of the neutron's spin vectors are given. The single particle spectrum and its approximate expression are obtained. The characteristics of quantum statistics with the approximate expression of energy spectrum are studied. (author). 15 refs
International Nuclear Information System (INIS)
Chen Qichang; Wu Hongchun; Cao Liangzhi
2008-01-01
A new 2D neutron transport code AutoMOC for arbitrary geometry has been developed. This code is based on the method of characteristics (MOCs) and the customization of AutoCAD. The MOC solves the neutron transport equation along characteristic lines. It is independent of the geometric shape of boundaries and regions. So theoretically, this method can be used to solve the neutron transport equation in highly complex geometries. However, it is important to describe the geometry and calculate intersection points of each characteristic line with every boundary and region in advance. In complex geometries, due to the complications of treating the arbitrary domain, the selection of geometric shapes and efficiency of ray tracing are generally limited. The geometry treatment through the customization of AutoCAD, a widely used computer-aided design software package, is given in this paper. Thanks to the powerful capability of AutoCAD, the description of arbitrary geometry becomes quite convenient. Moreover, with the language Visual Basic for Applications (VBAs), AutoCAD can be customized to carry out the ray tracing procedure with a high flexibility in geometry. The numerical results show that AutoMOC can solve 2D neutron transport problems in a complex geometry accurately and effectively
Energy Technology Data Exchange (ETDEWEB)
Chen Qichang; Wu Hongchun [School of Nuclear Science and Technology, Xi' an Jiaotong University, Xi' an Shaanxi 710049 (China); Cao Liangzhi [School of Nuclear Science and Technology, Xi' an Jiaotong University, Xi' an Shaanxi 710049 (China)], E-mail: caolz@mail.xjtu.edu.cn
2008-10-15
A new 2D neutron transport code AutoMOC for arbitrary geometry has been developed. This code is based on the method of characteristics (MOCs) and the customization of AutoCAD. The MOC solves the neutron transport equation along characteristic lines. It is independent of the geometric shape of boundaries and regions. So theoretically, this method can be used to solve the neutron transport equation in highly complex geometries. However, it is important to describe the geometry and calculate intersection points of each characteristic line with every boundary and region in advance. In complex geometries, due to the complications of treating the arbitrary domain, the selection of geometric shapes and efficiency of ray tracing are generally limited. The geometry treatment through the customization of AutoCAD, a widely used computer-aided design software package, is given in this paper. Thanks to the powerful capability of AutoCAD, the description of arbitrary geometry becomes quite convenient. Moreover, with the language Visual Basic for Applications (VBAs), AutoCAD can be customized to carry out the ray tracing procedure with a high flexibility in geometry. The numerical results show that AutoMOC can solve 2D neutron transport problems in a complex geometry accurately and effectively.
Kiani, Mohammad Amin; Ahmadi, Seyed Javad; Outokesh, Mohammad; Adeli, Ruhollah; Mohammadi, Aghil
2017-12-01
In this research, the characteristics of the prepared samples in epoxy matrix by means of X-ray diffraction (XRD), energy dispersive X-ray spectroscopy (EDS), as well as scanning electron microscope (SEM) are evaluated. Meanwhile, the obtained mechanical properties of the specimen are investigated. Thermogravimetric analysis (TGA) is also employed to evaluate the thermal degradation of manufactured nanocomposites. The thermal neutron absorption properties of nanocomposites containing 3 wt% of montmorillonite nanoclay (closite30B) have been studied experimentally, using an Am-Be point source. Mechanical tests reveal that the higher B4C concentrations, the more tensile strengths, but lower Young's modulus in all samples under consideration. TGA analysis also shows that thermal stability of the nanocomposite, increases in presence of B4C. Finally, neutron absorption analysis shows that increasing the B4C concentration leads to a nonlinearly build-up of neutron absorption cross section.
International Nuclear Information System (INIS)
Fu Kaifei; Peng Ruiyun; Gao Yabing; Wang Dewen; Chen Haoyu; Wu Xiaohong; Yang Yi; Hu Wenhua; Ma Junjie
2004-01-01
Objective: To compare the pathologic characteristics of gut-associated lymphoid tissues and lymphocyte apoptosis in neutron-irradiated mouse small intestines with those in γ-irradiated ones. Methods: Altogether 350 BALB/c mice were irradiated with different doses of neutrons or γ-rays, and were sacrificed on 6 h,12 h,125 d, 7 d, 14 d, 21 d and 28 d after irradiation and their total intestines were removed. Then the pathologic changes and death mode of lymphocytes in gut-associated lymphoid tissues were studied comparatively with light microscopy, electron microscopy and in situ terminal labeling method. Results: The basic pathologic changes of gut-associated lymphoid tissues after neutron irradiation included degeneration, apoptosis and necrosis of lymphocytes. The number of lymphocytes also decreased. There was no obvious regeneration after 4.0 and 5.5 Gy neutron irradiation, while after 2.5 Gy regeneration and recovery appeared, which were, there fore, dose-dependent. In the 2.5 Gy neutron group, the numbers of lymphocytes of intramucosal and submucous lymphoid tissues decreased, and karyopyknosis and a great quantity of nuclear fragments could also be observed at 6 h-3 d after irradiation. However, on the 3rd day regeneration of crypt epithelial cells appeared. On the 5th day hyperplasia of submucous lymphocytic tissues appeared, but recovery to normal level was not achieved till 14 d after irradiation. The basic pathologic changes after γ-irradiation were similar to that of neutron irradiation. Regeneration and recovery appeared in the 5.5 Gy group while no obvious regeneration in the 12.0 Gy group. The results of in situ terminal labeling indicated that at 6 h after irradiation the number of apoptotic cells in gut-associated lymphoid tissues of each group increased obviously, while in 4.0 Gy neutron group and 12.0 Gy γ-ray group it was more abundant. Conclusion: Both 2.5-5.5 Gy neutron and 5.5-12.0 Gy γ-ray irradiation can induce obvious injuries in gut
Energy Technology Data Exchange (ETDEWEB)
Yun, Sung Hwan; Baek, Min Ho; Yoo, Jae Woon; Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2012-05-15
The sodium-cooled fast reactor (SFR) has been chosen as a candidate for the Gen-IV Nuclear Energy Systems Initiative due to the advantages in utilization of uranium resources and reduction of radioactive wastes. Recently, the uranium blanket concept is omitted for a purpose of the non-proliferation, hence the reflector material plays a more important role in reactor core design. Moreover, especially in the Korean prototype SFR, the initial core should startup with low-enriched uranium ({<=} 20 w/o) for 100 {approx} 150 MWe power. This restriction causes significant difficulties to achieve sufficient excess reactivity. Thus, in this paper, core characteristic studies of various reflector materials (HT9, BeO, MgO, and ZrH{sub 1.6}) are performed to enhance the initial core excess reactivity
Mínguez-Gallego, Carlos; Vera-Remartinez, Enrique J; Albert-Coll, Monica; Roldán-Puchalt, M Concepción; Aguilar-Climent, Manuel; Rovira-Ferrando, Rosa E; Andrés-Soler, Jorge; Roig-Espert, Belén; Penadés-Vidal, María; Usó-Blasco, Jorge
2015-03-01
To describe the trend of the clinical and epidemiological characteristics of a cohort of HIV-1 infected patients in Castellón (Spain), and its impact on the delayed presentation. Data from HIV-1 infected outpatients presenting for care for the first time between 1987 and 2011 were retrospectively analyzed. There have been significant changes in the characteristics of the 1001 newly presented patients during the period studied. An increase in the mean age was observed (increasing from about 30 years before 1996, to approximately 35 after the 2000-2002 period), as well as an increase in the percentage of immigrants (de Enfermedades Infecciosas y Microbiología Clínica. All rights reserved.
International Nuclear Information System (INIS)
Vorona, P.M.; Razbudej, V.F.
2010-01-01
Calculational studies and analysis of the neutron fields of WWR-M research reactor of the Institute for Nuclear Research, National Academy of Sciences of Ukraine, as a basic nuclear facility for performing the fundamental and applied investigations and for experimentalindustrial production of radioisotope products for various spheres of application are carried out. The calculations are carried out by the method of statistic tests (Monte Carlo) applying the computer program MCNP-4C. The data on the spectra and the neutron flux density values at the 10 MW reactor power for all technological facilities designed for the works with neutrons: 19 vertical experimental channels for irradiation of specimens and 10 horizontal channels for beams extraction from the reactor are obtained. The effect of the neutron traps (water cavities) mounted in the core on the characteristics of the extracted from the reactor beams is demonstrated. Recommendations associated with optimization of the reactor core are adduced for amplification of its capabilities as a neutron source in experimental researches.
International Nuclear Information System (INIS)
Tasaki, Yutaka; Ichikawa, Yasushi; Kobo, Norio; Shinohara, Kazuhiko; Boillat, Pierre; Kramer, Denis; Scherer, Gunther G.; Lehmann, Eberhard H.
2008-01-01
The mass transfer characteristics of gas diffusion layer (GDL) are closely related to cell performance in PEFC. In this study, In situ diagnostic of water distribution in thickness direction of MEA by Neutron Imaging has been carried out for three MEAs with different GDLs on cathode side as well as I-V characteristics. It was confirmed that this method is useful for analyzing water distribution in thickness direction of MEA. The relationship between I-V characteristics and liquid water distribution has been studied. (author)
International Nuclear Information System (INIS)
Petkov, P.T.
2000-01-01
The method of characteristics (MOC) is gaining increased popularity in the reactor physics community all over the world because it gives a new degree of freedom in nuclear reactor analysis. The MARIKO code solves the neutron transport equation by the MOC in two-dimensional real geometry. The domain of solution can be a rectangle or right hexagon with periodic boundary conditions on the outer boundary. Any reasonable symmetry inside the domain can be fully accounted for. The geometry is described in three levels-macro-cells, cells, and regions. The macro-cells and cells can be any polygon. The outer boundary of a region can be any combination of straight line and circular arc segments. Any level of embedded regions is allowed. Procedures for automatic geometry description of hexagonal fuel assemblies and reflector macro-cells have been developed. The initial ray tracing procedure is performed for the full rectangular or hexagonal domain, but only azimuthal angles in the smallest symmetry interval are tracked. (Authors)
Energy Technology Data Exchange (ETDEWEB)
Abdkadyrova, I Kh [AN RU, Tashkent (Uzbekistan). Inst. Yadernoj Fiziki
1997-02-01
Lattice structure distortions in irradiated crystalline quartz were studied by refractometry methods. The refractometry constants of {alpha}-quartz for the flux of fast neutrons 10{sup 18} - 10{sup 21} neutron/cm{sup 2} were calculated. The critical kinetics of this constants at the phase transformation is observed.(author). 5 refs., 1 fig.
DEFF Research Database (Denmark)
Tömösközi, Máté; Fitzek, Frank; Roetter, Daniel Enrique Lucani
2015-01-01
Low latency and high reliability is a critical characteristics of many wireless use cases, such as real–time video surveillance. In a dense network the best way to resolve such problems is by utilizing the fastest and most reliable connection to the destination. Classical solutions, such as Reed–...
International Nuclear Information System (INIS)
Znak, N.Y.; Morgun, V.N.
1995-01-01
Efficiency of the energy transformation for CO2 fixation (E), and kinetics of the initial O-2-mediated electron transport of Spirulina platensis (Gem. ) Geitl, and Chlorella vulgaris Beijerinck cells were measured after adaptation to various growth irradiances (I) by means of the delayed fluorescence (DF) induction curves. Maxima of the membrane potential expenses during induction period were observed at I half saturating oxygen evolution; they were shifted according to growth I remaining higher in Spirulina than in Chlorella. The alterations of absorbance and fluorescence spectra at 25 degrees C after adaptation to I demonstrated changes in composition of pigments of algae, created to compensate for the imbalance in radiation absorption between the two photosystems. For Spirulina cells, the value of E was higher after growing under low I, or under blue radiation absorbed mainly by photosystem (PS) 1 (400-500 nm) with excitation by yellow (570 nm) radiation. For Chlorella cells, it was also higher after growing under low I. Under such conditions the half-rise time for DP-phase of DF induction curve decreased, which reflected an acceleration of kinetics of the initial electron transport between photosystems. An opposite situation was observed with Spirulina cells grown under high I or yellow radiation, and Chlorella cells from high I. Enhancement of effective PS2/PS1 ratio associated with decrease of reaction centre (RC) 2/RC1 stoichiometry may because of the increase of E and high membrane energization under saturating I in algae adapted to low I
Energy Technology Data Exchange (ETDEWEB)
Brookes, I. R. [Atomic Weapons Research Establishment, Aldermaston, Berks. (United Kingdom)
1965-10-15
For this method {sup 235}U is assayed by counting delayed neutrons emitted following the fission of {sup 235}U in the sample with thermal neutrons. The three groups of neutrons of interest have half-lives of 55.72, 22.72 and 6.22 s. A 100-ml sample of urine is evaporated to dryness on a water bath. The residue is transferred with about 4-5 ml of water to a 1-oz polythene bottle which is then heat sealed. The sample bottle is put into a ''rabbit'' and sent through a pneumatic tube system to the HERALD reactor core where it is irradiated in a thermal flux of 3.94 x 10{sup 12} n cm{sup 2}/s. The sample automatically returns to the laboratory after 60 s irradiation where the sample bottle is transferred to the counter. This counter is switched on 25 s after the sample has left the reactor and counts the sample for 1 min. Blank samples consist of urine from persons occupationally unexposed to uranium and the calibration standard is a known amount of {sup 235}U (as natural uranium). The limit of detection is 0.020 pCi of 93% enriched uranium (0.007 of the maximum permissible body- burden) and 0.036 {mu}g of natural uranium per 100 ml of urine. The limit of detection is governed by the magnitude of the blank count. The main component of the blank is the response of the counter to {gamma}-radiation from activation products in irradiated urine. The counting and irradiation cycle takes about 3 Vulgar-Fraction-One-Half min and 50 samples at one time may be evaporated and bottled in the working day. Interference from {sup 239}Pu is likely to be negligible for the purposes of urine analysis. (author) [French] U methode decrite consiste a doser l'uranium-235 d'un echantillon en comptant les neutrons differes emis lors de la fission de {sup 235}U par neutrons thermiques. Les trois groupes de neutrons qui presentent un interet ont des periodes de 55,72, 22, 72 et 6, 22 s. On fait evaporer 100 ml d'urine au bain-marie. Le residu est transvase avec 4 a 5 ml d'eau dans un flacon en
International Nuclear Information System (INIS)
Aleksandrov, A.S.; Gorbovskij, A.I.; Mishagin, V.V.
1994-01-01
The paper reviews designs of magnets and vacuum system of the Hydrogen Prototype of the Neutron Source. An idea of this neutron source is based on the use of neutral-beam-driven plasma in an axisymmetric magnetic mirror to generate high flux D-T neutrons. Preliminary evaluations have shown that such a source has several potential advantages when is used for fusion material and component tests. The Hydrogen Prototype is essentially full scale model of the source but operated with a hydrogen plasma. 10 refs.; 6 figs.; 1 tab
International Nuclear Information System (INIS)
Sosnin, A.N.; Polanski, A.; Petrochenkov, S.A.
2002-01-01
Neutron fields generated in extended heavy (Z ≥ 82) targets under irradiation with proton beams at energies in the range of 1 GeV are investigated. Influence of different moderators on the spectra and multiplicities of neutrons escaping the surface of the assembly consisting of a lead target (diam. 8 cm x 20 cm or diam. 8 cm x 50 cm) screened by variable thickness of polyethylene or graphite, respectively, was compared. It is shown that the effectiveness of graphite as a material used in such assemblies to moderate spallation neutrons down to thermal energies is significantly lower than that of paraffin
Sosnin, A N; Polanski, A; Petrochenkov, S A; Golovatyuk, V M; Krivopustov, M I; Bamblevski, V P; Westmeier, W; Odoj, R; Brandt, R; Robotham, H; Hashemi-Nezhad, S R; Zamani-Valassiadou, M
2002-01-01
Neutron fields generated in extended heavy (Z\\geq 82) targets under irradiation with proton beams at energies in the range of 1 GeV are investigated. Influence of different moderators on the spectra and multiplicities of neutrons escaping the surface of the assembly consisting of a lead target (\\varnothing 8 cm\\times 20 cm or \\varnothing 8cm\\times 50 cm) screened by variable thickness of polyethylene or graphite, respectively, was compared in the present work. It is shown that the effectiveness of graphite as a material used in such assemblies to moderate spallation neutrons down to thermal energies is significantly lower than that of paraffin.
Neutronics of pulsed spallation neutron sources
International Nuclear Information System (INIS)
Watanabe, Noboru
2003-01-01
Various topics and issues on the neutronics of pulsed spallation neutron sources, mainly for neutron scattering experiments, are reviewed to give a wide circle of readers a better understanding of these sources in order to achieve a high neutronic performance. Starting from what neutrons are needed, what the spallation reaction is and how to produce slow-neutrons more efficiently, the outline of the target and moderator neutronics are explained. Various efforts with some new concepts or ideas have already been devoted to obtaining the highest possible slow-neutron intensity with desired pulse characteristics. This paper also reviews the recent progress of such efforts, mainly focused on moderator neutronics, since moderators are the final devices of a neutron source, which determine the source performance. Various governing parameters for neutron-pulse characteristics such as material issues, geometrical parameters (shape and dimensions), the target-moderator coupling scheme, the ortho-para-hydrogen ratio, poisoning, etc are discussed, aiming at a high performance pulsed spallation source
International Nuclear Information System (INIS)
Kugo, Teruhiko; Mori, Takamasa; Takeda, Toshikazu
2007-01-01
Extended bias factor methods are proposed with two new concepts, the LC method and the PE method, in order to effectively use critical experiments and to enhance the applicability of the bias factor method for the improvement of the prediction accuracy of neutronic characteristics of a target core. Both methods utilize a number of critical experimental results and produce a semifictitious experimental value with them. The LC and PE methods define the semifictitious experimental values by a linear combination of experimental values and the product of exponentiated experimental values, respectively, and the corresponding semifictitious calculation values by those of calculation values. A bias factor is defined by the ratio of the semifictitious experimental value to the semifictitious calculation value in both methods. We formulate how to determine weights for the LC method and exponents for the PE method in order to minimize the variance of the design prediction value obtained by multiplying the design calculation value by the bias factor. From a theoretical comparison of these new methods with the conventional method which utilizes a single experimental result and the generalized bias factor method which was previously proposed to utilize a number of experimental results, it is concluded that the PE method is the most useful method for improving the prediction accuracy. The main advantages of the PE method are summarized as follows. The prediction accuracy is necessarily improved compared with the design calculation value even when experimental results include large experimental errors. This is a special feature that the other methods do not have. The prediction accuracy is most effectively improved by utilizing all the experimental results. From these facts, it can be said that the PE method effectively utilizes all the experimental results and has a possibility to make a full-scale-mockup experiment unnecessary with the use of existing and future benchmark
Fuel composition effects on HYPER core characteristics
International Nuclear Information System (INIS)
Han, Chi Young; Kim, Yong Nam; Kim, Jong Kyung
2001-01-01
At KAERI(Korea Atomic Energy Research Institute), a subcritical transmutation reactor is under development, named HYPER(Hybrid Power Extraction Reactor). For the HYPER system, a pyrochemical process is being considered for fuel reprocessing. Separated from the separation process, the fuel contains not only TRU but also the considerable percentages of impurity such as uranium nuclides and lanthanides. The amount of these impurities depends on strongly the refining efficiency of the reprocessing and may change the core characteristics. This paper has analyzed fuel composition effects on th HYPER core characteristics. Assuming various recovery factors of uranium and lanthanides, some dynamic parameters have been evaluated which are the neutron spectrum, the neutron reaction balance, the reactivity coefficients, the effective delayed neutron fraction, and the effective neutron lifetime
International Nuclear Information System (INIS)
Barnard, R.W.; Jensen, D.H.
1982-01-01
Uranium formations are assayed by prompt fission neutron logging techniques. The uranium in the formation is proportional to the ratio of epithermal counts to thermal or eqithermal dieaway. Various calibration factors enhance the accuracy of the measurement
International Nuclear Information System (INIS)
Cui, Shijie; Zhang, Dalin; Cheng, Jie; Tian, Wenxi; Su, G.H.
2017-01-01
As one of the candidate tritium breeding blankets for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of the helium cooled solid breeder blanket has recently been proposed. The neutronic, thermal-hydraulic and mechanical characteristics of the blanket directly affect its tritium breeding and safety performance. Therefore, neutronic/thermal-hydraulic/mechanical coupling analyses are of vital importance for a reliable blanket design. In this work, first, three-dimensional neutronics analysis and optimization of the typical outboard equatorial blanket module (No. 12) were performed for the comprehensive optimal scheme. Then, thermal and fluid dynamic analyses of the scheme under both normal and critical conditions were performed and coupled with the previous neutronic calculation results. With thermal-hydraulic boundaries, thermo-mechanical analyses of the structure materials under normal, critical and blanket over-pressurization conditions were carried out. In addition, several parametric sensitivity studies were also conducted to investigate the influences of the main parameters on the blanket temperature distributions. In this paper, the coupled analyses verify the reasonability of the optimized conceptual design preliminarily and can provide an important reference for the further analysis and optimization design of the CFETR helium cooled solid breeder blanket.
Energy Technology Data Exchange (ETDEWEB)
Cui, Shijie; Zhang, Dalin, E-mail: dlzhang@mail.xjtu.edu.cn; Cheng, Jie; Tian, Wenxi; Su, G.H.
2017-01-15
As one of the candidate tritium breeding blankets for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of the helium cooled solid breeder blanket has recently been proposed. The neutronic, thermal-hydraulic and mechanical characteristics of the blanket directly affect its tritium breeding and safety performance. Therefore, neutronic/thermal-hydraulic/mechanical coupling analyses are of vital importance for a reliable blanket design. In this work, first, three-dimensional neutronics analysis and optimization of the typical outboard equatorial blanket module (No. 12) were performed for the comprehensive optimal scheme. Then, thermal and fluid dynamic analyses of the scheme under both normal and critical conditions were performed and coupled with the previous neutronic calculation results. With thermal-hydraulic boundaries, thermo-mechanical analyses of the structure materials under normal, critical and blanket over-pressurization conditions were carried out. In addition, several parametric sensitivity studies were also conducted to investigate the influences of the main parameters on the blanket temperature distributions. In this paper, the coupled analyses verify the reasonability of the optimized conceptual design preliminarily and can provide an important reference for the further analysis and optimization design of the CFETR helium cooled solid breeder blanket.
International Nuclear Information System (INIS)
Krotov, A.D.; Son'ko, A.V.
2009-01-01
Calculation of neutron-physical properties and radiation protection of space power reactor was made by means of the MCNP code allowing simulation of neutron, γ- and electron transport by the Monte Carlo method in the systems with combined geometry. Universality of the MCNP code has been demonstrated both for the calculation of reactor-converter so for the optimization of radiation protection that allows to reserve a new level of complex simulation of SNPS [ru
Im, Ui-Su; Kim, Jiyoung; Lee, Seon Ho; Lee, Byung-Rok; Peck, Dong-Hyun; Jung, Doo-Hwan
2017-12-01
In the present study, surface texture features and chemical properties of two types of cokes, made from coal tar by either 1-stage heat treatment or 2-stage heat treatment, were researched. The relationship between surface texture characteristics and the chemical properties was identified through molecular weight distribution, insolubility of coal tar, weight loss with temperature increase, coking yield, and polarized light microscope analysis. Rapidly cleared anisotropy texture in cokes was observed in accordance with the coking temperature rise. Quinoline insolubility and toluene insolubility of coal tar increased with a corresponding increases in coking temperature. In particular, the cokes produced by the 2-stage heat treatment (2S-C) showed surface structure of needle cokes at a temperature approximately 50°C lower than the 1-stage heat treatment (1S-C). Additionally, the coking yield of 2S-C increased by approximately 14% in comparison with 1S-C.
International Nuclear Information System (INIS)
Mostafaei, F; McNeill, F E; Chettle, D R; Prestwich, W V; Inskip, M
2013-01-01
Fluorine is an element that can be either beneficial or harmful, depending on the total amount accumulated in the teeth or bones. In our laboratory, we have developed a non-invasive technique for the in vivo measurement of fluoride in bone using neutron activation analysis and performed the first pilot human study. Fluoride in humans is quantified by comparing the γ-ray signal from a person to the γ-ray signal obtained from appropriate anthropomorphic calibration phantoms. An identified problem with existing fluoride phantoms is contamination with aluminum. Aluminum creates an interfering γ-ray signal which, although it can be subtracted out, increases the uncertainty in the measurement and worsens the detection limit. This paper outlines a series of studies undertaken to develop a better calibration phantom for fluorine measurement, which does not have aluminum contamination. (paper)
International Nuclear Information System (INIS)
Hursin, Mathieu; Xiao Shanjie; Jevremovic, Tatjana
2006-01-01
This paper summarizes the theoretical and numerical aspects of the AGENT code methodology accurately applied for detailed three-dimensional (3D) multigroup steady-state modeling of neutron interactions in complex heterogeneous reactor domains. For the first time we show the fine-mesh neutron scalar flux distribution in Purdue research reactor (that was built over forty years ago). The AGENT methodology is based on the unique combination of the three theories: the method of characteristics (MOC) used to simulate the neutron transport in two-dimensional (2D) whole core heterogeneous calculation, the theory of R-functions used as a mathematical tool to describe the true geometry and fuse with the MOC equations, and one-dimensional (1D) higher-order diffusion correction of 2D transport model to account for full 3D heterogeneous whole core representation. The synergism between the radial 2D transport and the 1D axial transport (to take into account the axial neutron interactions and leakage), called the 2D/1D method (used in DeCART and CHAPLET codes), provides a 3D computational solution. The unique synergism between the AGENT geometrical algorithm capable of modeling any current or future reactor core geometry and 3D neutron transport methodology is described in details. The 3D AGENT accuracy and its efficiency are demonstrated showing the eigenvalues, point-wise flux and reaction rate distributions in representative reactor geometries. The AGENT code, comprising this synergism, represents a building block of the computational system, called the virtual reactor. Its main purpose is to perform 'virtual' experiments and demonstrations of various mainly university research reactor experiments
Energy Technology Data Exchange (ETDEWEB)
Vega C, H. R.; Hernandez D, V. M.; Letechipia de L, C.; Salas L, M. A. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas, Zac. (Mexico); Rodriguez R, J. A.; Juarez A, C. A., E-mail: fermineutron@yahoo.com [Universidad Autonoma de Nuevo Leon, Facultad de Ingenieria Civil, Pedro de Alba s/n, San Nicolas de los Garza, Nuevo Leon (Mexico)
2016-09-15
Shielding is an important element in radiation protection since allows the management of radiation sources. Currently there are different materials of natural or anthropogenic origin that are used as shielding for both photons and neutrons. The quarry is a material of natural origin and abundant in our country, which is used in construction or for the manufacture of sculptures, however its characteristics as shielding have not been reported. In this paper we report some of the properties of the quarry as shielding for monoenergetic photons and for neutrons produced by an isotopic neutron source of {sup 241}AmBe. A quarry piece was used to determine its density and its chemical composition, with the XCOM code the elemental composition was determined and the mass interaction and total attenuation coefficients of the quarry were determined with photons of 10{sup -3} to 10{sup -5} MeV; the interaction coefficients included coherent dispersion, photoelectric absorption, Compton dispersion and the production of pairs in the nuclear and electronic field. Using the MCNP5 code, a narrow geometry attenuation experiment was modeled and the photon fluence was estimated that reaches a point detector at a distance of 42 cm from a point source, isotropic and monoenergetic photon when the source and the point detector were added quarry pieces of different thicknesses. The reduction of the number of photons as a function of the thickness of the quarry was used to determine the coefficient of linear attenuation of the quarry before photons of 0.03, 0.07, 0.1, 0.3, 1, 2 and 3 MeV that were the same as those calculated with the XCOM code. With the MCNP, the K a and H(10) transmission curves were also calculated. This same model was used to determined the variation of the {sup 241}AmBe neutron spectrum as a function of quarry thickness, as well as the E{sub ROT} and H(10) transmission curves. (Author)
International Nuclear Information System (INIS)
Santandrea, Simone
2001-01-01
This research thesis addresses the resolution of the neutron transport equation inside reactor cells in non-structured grids and in general geometry by using the method of characteristics (MoC) and two acceleration methods developed during this research. The author introduces the MoC with a flat approximation of the neutron collision source within each computation area. This formulation leads to a linear approximation. The next part presents the mathematical framework for the use of the Lanczos iterative scheme. A new acceleration method is then introduced. The last part reports realistic cases with a high spatial and angular heterogeneity. Results obtained by using the Apollo2-TDT code are compared with those obtained with the Tripoli4 Monte-Carlo code [fr
International Nuclear Information System (INIS)
Zacharatou Jarlskog, Christina; Paganetti, Harald
2008-01-01
Purpose: To estimate the risk of a second malignancy after treatment of a primary brain cancer using passive scattered proton beam therapy. The focus was on the cancer risk caused by neutrons outside the treatment volume and the dependency on the patient's age. Methods and Materials: Organ-specific neutron-equivalent doses previously calculated for eight different proton therapy brain fields were considered. Organ-specific models were applied to assess the risk of developing solid cancers and leukemia. Results: The main contributors (>80%) to the neutron-induced risk are neutrons generated in the treatment head. Treatment volume can influence the risk by up to a factor of ∼2. Young patients are subject to significantly greater risks than are adult patients because of the geometric differences and age dependency of the risk models. Breast cancer should be the main concern for females. For males, the risks of lung cancer, leukemia, and thyroid cancer were significant for pediatric patients. In contrast, leukemia was the leading risk for an adult. Most lifetime risks were <1% (70-Gy treatment). The only exceptions were breast, thyroid, and lung cancer for females. For female thyroid cancer, the treatment risk can exceed the baseline risk. Conclusion: The risk of developing a second malignancy from neutrons from proton beam therapy of a brain lesion is small (i.e., presumably outweighed by the therapeutic benefit) but not negligible (i.e., potentially greater than the baseline risk). The patient's age at treatment plays a major role
International Nuclear Information System (INIS)
Antov, A.; Stoyanova, I.
2008-01-01
In conjunction with each refuelling shutdown of the reactor core, nuclear design calculations are performed to ensure that the reactor physics characteristics of the new core will be consistent with the safety limits. Prior to return to normal operation, a physics test program is required to determine if the operating characteristics of the core are consistent with the design predictions and to ensure that the core can be operated as designed. Successful completion of the physics test program is demonstrated when the test results agree with the predicted results within predetermined test criteria. Successful completion of the physics test program and successful completion of other tests, which are performed after each refuelling provides assurance that the plant can be operated as designed. The calculated neutron-physics characteristics values of Kozloduy NPP Unit 5 and Unit 6 (WWER 1000) obtained by the computer code package KASKAD are compared with the obtained results during the start up physics tests. The core fuel loading consists of 163 fuel assemblies (FAs). The calculated values are given according to actual experimental conditions of the reactor core during start up physics tests. The report includes comparisons between calculation results by code package KASKAD (BIPR7A) and experimental data values of main neutron-physics characteristics during start up physics tests in selected recent cycles of Kozloduy NPP Unit 5 and Unit 6. (authors)
International Nuclear Information System (INIS)
Stoyanova, I.; Antov, A.
2007-01-01
In conjunction with each refuelling shutdown of the reactor core, nuclear design calculations are performed to ensure that the reactor physics characteristics of the new core will be consistent with the safety limits. Prior to return to normal operation, a physics test program is required to determine if the operating characteristics of the core are consistent with the design predictions and to ensure that the core can be operated as designed. Successful completion of the physics test program is demonstrated when the test results agree with the predicted results within predetermined test criteria. Successful completion of the physics test program and successful completion of other tests which are performed after each refuelling provides assurance that the plant can be operated as designed. The calculated neutron-physics characteristics values of Kozloduy NPP Unit 5 and Unit 6 (WWER 1000) obtained by the computer code package KASKAD are compared with the obtained results during the start up physics tests. The core fuel loading consists of 163 fuel assemblies. The calculated values are given according to actual experimental conditions of the reactor core during start up physics tests. The report includes comparisons between calculation results by code package KASKAD (BIPR7A) and experimental data values of main neutron-physics characteristics during start up physics tests in selected recent cycles of Kozloduy NPP Unit 5 and Unit 6 (Authors)
Neutron and gamma characterization within the FFTF reactor cavity
International Nuclear Information System (INIS)
Bunch, W.L.; Carter, L.L.; Moore, F.S.; Werner, E.J.; Wilcox, A.D.; Wood, M.R.
1980-08-01
Neutron and gamma ray measurements were made within the reactor cavity of the Fast Flux Test Facility (FFTF) to establish the operating characteristics of the Ex-Vessel Flux Monitoring (EVFM) system as a function of reactor power level. A significant effort was made to obtain absolute flux values in order that the measurements could be compared directly with shield design calculations. Good agreement was achieved for neutrons and for both the prompt and delayed components of the gamma ray field. 8 figures, 3 tables
Leal, L C; Kitis, G; Guber, K H; Quaranta, A; Koehler, P E
2002-01-01
In the scope of our programme we propose to start in July 2000 with measurements on elements of well known cross sections, in order to check the reliability of the whole experimental installation at the CERN-TOF facility. These initial exploratory measurements will provide the key-parameters required for the further experimentation at the CERN-TOF neutron beam. The neutron fluence and energy resolution will be determined as a function of the neutron kinetic energy by reproducing standard capture and fission cross sections. The measurements of capture cross sections on elements with specific cross section features will allow to us to disentangle the different components of backgrounds and estimate their level in the experimental area. The time-energy calibration will be determined and monitored with a set of monoenergetic filters as well as by the measurements of elements with resonance-dominated cross sections. Finally, in this initial phase the behaviour of several detectors scheduled in successive measureme...
International Nuclear Information System (INIS)
Awschalom, M.; Rosenberg, I.; Ten Haken, R.K.; Pearson, D.W.; DeLuca, P.M. Jr.; Attix, F.H.
1982-01-01
The average energy necessary to produce an electron-ion pair (anti W) of a gas mixture having an atomic composition very close to that of A-150 plastic has been studied through use in different size ionization chambers made of that plastic in a p(66)Be(49) neutron therapy beam. A tentative value for anti W(A-150-gas) of 27.3 +/ -0.5 J C -1 was derived. The anti W value of the A-150 equivalent gas mixture is compared to those of methane-based tissue-equivalent gas and of air for the p(66)Be(49) neutron beam as well as to corresponding values found in similar experiments using 14.8 MeV monoenergetic neutrons
International Nuclear Information System (INIS)
1964-01-01
During recent years the interest in biological effects caused by neutrons has been increasing steadily as a result of the rapid development of neutron technology and the great number of neutron sources being used. Neutrons, because of their specific physical characteristics and biological effects, form a special type of radiation hazard but, at the same time, are a prospective tool for applied radiobiology. This Symposium, held in Brookhaven at the invitation of the United States Government from 7-11 October 1963, provided an opportunity for scientists to discuss the experimental information at present available on the biological action of neutrons and to evaluate future possibilities. It was a sequel to the Symposium on Neutron Detection, Dosimetry and Standardization, which was organized by the International Atomic Energy Agency in December 1962 at Harwell. The Symposium was attended by 128 participants from 17 countries and 6 international organizations. Fifty-four papers were presented. The following subjects were discussed in various sessions: (1) Dosimetry. Estimation of absorbed dose of neutrons in biological material. (2) Biological effects of high-energy protons. (3) Cellular and genetic effects. (4) Pathology of neutron irradiation, including acute and chronic radiation syndromes (mortality, anatomical and histological changes, biochemical and metabolic disturbances) and delayed consequences. (5) Relative biological effectiveness of neutrons evaluated by different biological tests. A Panel on Biophysical Considerations in Neutron Experimentation, with special emphasis on informal discussions, was organized during the Symposium. The views of the Panel are recorded in Volume II of the Proceedings. Many reports were presented on the important subject of the relative effectiveness of the biological action of neutrons, as well as on the general pathology of neutron irradiation and the cellular and genetic effects related to it. Three survey papers considered
A neutron amplifier: prospects for reactor-based waste transmutation
International Nuclear Information System (INIS)
Blanovsky, A.
2004-01-01
A design concept and characteristics for an epithermal breeder controlled by variable feedback and external neutron source intensity are presented. By replacing the control rods with neutron sources, we could maintain good power distribution and perform radioactive waste burning in high flux subcritical reactors (HFSR) that have primary system size, power density and cost comparable to a pressurized water reactor (PWR). Another approach for actinide transmutation is a molten salt subcritical reactor proposed by Russian scientists. To increase neutron source intensity the HFSR is divided into two zones: a booster and a blanket with solid and liquid fuels. A neutron gate (absorber and moderator) imposed between two zones permits fast neutrons from the booster to flow to the blanket. Neutrons moving in the reverse direction are moderated and absorbed in the absorber zone. In the HFSR, neptunium-plutonium fuel is circulated in the booster and blanket, and americium-curium in the absorber zone and outer reflector. Use of a liquid actinide fuel permits transport of the delayed-neutron emitters from the blanket to the booster, where they can provide additional neutrons (source-dominated mode) or all the necessary excitation without an external neutron source (self-amplifying mode). With a blanket neutron multiplication gain of 20 and a booster gain of 50, an external neutron source rate of at least 10 15 n/s (0.7 MW D-T or 2.5 MW electron beam power) is needed to control the HFSR that produces 300 MWt. Most of the power could be generated in the blanket that burns about 100 kg of actinides a year. The analysis takes into consideration a wide range of HFSR design aspects including the wave model of observed relativistic phenomena, plant seismic diagnostics, fission electric cells (FEC) with a multistage collector (anode) and layered cathode. (author)
Energy Technology Data Exchange (ETDEWEB)
Hatsukawa, Yuichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
1997-07-01
Delayed fission is a nuclear decay process that couples {beta} decay and fission. In the delayed fission process, a parent nucleus undergoes {beta} decay and thereby populates excited states in the daughter. If these states are of energies comparable to or greater than the fission barrier of the daughter, then fission may compete with other decay modes of the excited states in the daughter. In this paper, mechanism and some experiments of the delayed fission will be discussed. (author)
NGI-9 pulsed neutron generator with a fluence to 1010 n/s
International Nuclear Information System (INIS)
Allakhverdov, A.Sh.; Ogarkin, V.I.; Silicheva, G.P.; Timofeev, Yu.I.
1975-01-01
A neutron pulse generator with 14 MeV energy used for the activation analysis, is described. Its functional diagram and the technical characteristics are presented. The studies of the generator that resulted in determination of the effect of the accelerating voltage amplitude, the delay in the ion source firing with respect to the pulse of the accelerating voltage, the amount of operating ion sources and the energy imparted to them on the neutron flux magnitude are conducted. It is confirmed by the studies that the neutron generator operating in the nominal regime makes it possible to obtain a neutron flux of 5x10 9 -10 10 neutr./s. The dependence of the neutron flux variation on the frequency of pulse sequence for various ion sources is shown
Gatera, Angélique; Göök, Alf; Hambsch, Franz-Josef; Moens, André; Oberstedt, Andreas; Oberstedt, Stephan; Sibbens, Goedele; Vanleeuw, David; Vidali, Marzio
2018-03-01
Recent years have seen an increased interest in prompt fission γ-ray (PFG) measurements motivated by a high priority request of the OECD/NEA for high precision data, mainly for the nuclear fuel isotopes 235U and 239Pu. Our group has conducted a PFG measurement campaign using state-of-the-art lanthanum halide detectors for all the main actinides to a precision better than 3%. The experiments were performed in a coincidence setup between a fission trigger and γ-ray detectors. The time-of-flight technique was used to discriminate photons, traveling at the speed of light, and prompt fission neutrons. For a full rejection of all neutrons below 20 MeV, the PFG time window should not be wider than a few nanoseconds. This window includes most PFG, provided that no isomeric states were populated during the de-excitation process. When isomeric states are populated, PFGs can still be emitted up to 1 yus after the instant of fission or later. To study these γ-rays, the detector response to neutrons had to be determined and a correction had to be applied to the γ-ray spectra. The latest results for PFG characteristics from the reaction 239Pu(nth,f) will be presented, together with an analysis of PFGs emitted up to 200 ns after fission in the spontaneous fission of 252Cf as well as for thermal-neutron induced fission on 235U and 239Pu. The results are compared with calculations in the framework of the Hauser-Feshbach Monte Carlo code CGMF and FIFRELIN.
Chamberlain, Mike; Gräfe, James L; Aslam; Byun, Soo-Hyun; Chettle, David R; Egden, Lesley M; Webber, Colin E; McNeill, Fiona E
2012-03-01
Humans can be exposed to fluorine (F) through their diet, occupation, environment and oral dental care products. Fluorine, at proper dosages, is believed to have positive effects by reducing the incidence of dental caries, but fluorine toxicity can occur when people are exposed to excessive quantities of fluorine. In this paper we present the results of a small pilot in vivo study on 33 participants living in Southwestern Ontario, Canada. The mean age of participants was 45 ± 18 years with a range of 20-87 years. The observed calcium normalized hand-bone-fluorine concentrations in this small pilot study ranged from 1.1 to 8.8 mg F/g Ca. Every person measured in this study had levels of fluorine in bone above the detection limit of the system. The average fluorine concentration in bone was found to be 3.5 ± 0.4 mg F/g Ca. No difference was observed in average concentration for men and women. In addition, a significant correlation (r(2) = 0.55, p fluorine content and age. The amount of fluorine was found to increase at a rate of 0.084 ± 0.014 mg F/g Ca per year. There was no significant difference observed in this small group of subjects between the accumulation rates in men and women. To the best of our knowledge, this is the first time data from in vivo measurement of fluorine content in humans by neutron activation analysis have been presented. The data determined by this technique were found to be consistent with results from ex vivo studies from other countries. We suggest that the data demonstrate that this low risk non-invasive diagnostic technique will permit the routine assessment of bone-fluorine content with potential application in the study of clinical bone-related diseases. This small study demonstrated that people in Southern Ontario are exposed to fluoride in measureable quantities, and that fluoride can be seen to accumulate in bone with age. However, all volunteers were found to have levels below those expected with clinical fluorosis, and only
International Nuclear Information System (INIS)
Chamberlain, Mike; Gräfe, James L; Aslam; Byun, Soo-Hyun; Chettle, David R; Egden, Lesley M; Webber, Colin E; McNeill, Fiona E
2012-01-01
Humans can be exposed to fluorine (F) through their diet, occupation, environment and oral dental care products. Fluorine, at proper dosages, is believed to have positive effects by reducing the incidence of dental caries, but fluorine toxicity can occur when people are exposed to excessive quantities of fluorine. In this paper we present the results of a small pilot in vivo study on 33 participants living in Southwestern Ontario, Canada. The mean age of participants was 45 ± 18 years with a range of 20–87 years. The observed calcium normalized hand-bone-fluorine concentrations in this small pilot study ranged from 1.1 to 8.8 mg F/g Ca. Every person measured in this study had levels of fluorine in bone above the detection limit of the system. The average fluorine concentration in bone was found to be 3.5 ± 0.4 mg F/g Ca. No difference was observed in average concentration for men and women. In addition, a significant correlation (r 2 = 0.55, p < 0.001) was observed between hand-bone-fluorine content and age. The amount of fluorine was found to increase at a rate of 0.084 ± 0.014 mg F/g Ca per year. There was no significant difference observed in this small group of subjects between the accumulation rates in men and women. To the best of our knowledge, this is the first time data from in vivo measurement of fluorine content in humans by neutron activation analysis have been presented. The data determined by this technique were found to be consistent with results from ex vivo studies from other countries. We suggest that the data demonstrate that this low risk non-invasive diagnostic technique will permit the routine assessment of bone-fluorine content with potential application in the study of clinical bone-related diseases. This small study demonstrated that people in Southern Ontario are exposed to fluoride in measureable quantities, and that fluoride can be seen to accumulate in bone with age. However, all volunteers were found to have levels below those
International Nuclear Information System (INIS)
Hoffman, A. J.; Lee, J. C.
2013-01-01
A new time-dependent neutron transport method based on the method of characteristics (MOC) has been developed. Whereas most spatial kinetics methods treat time dependence through temporal discretization, this new method treats time dependence by defining the characteristics to span space and time. In this implementation regions are defined in space-time where the thickness of the region in time fulfills an analogous role to the time step in discretized methods. The time dependence of the local source is approximated using a truncated Taylor series expansion with high order derivatives approximated using backward differences, permitting the solution of the resulting space-time characteristic equation. To avoid a drastic increase in computational expense and memory requirements due to solving many discrete characteristics in the space-time planes, the temporal variation of the boundary source is similarly approximated. This allows the characteristics in the space-time plane to be represented analytically rather than discretely, resulting in an algorithm comparable in implementation and expense to one that arises from conventional time integration techniques. Furthermore, by defining the boundary flux time derivative in terms of the preceding local source time derivative and boundary flux time derivative, the need to store angularly-dependent data is avoided without approximating the angular dependence of the angular flux time derivative. The accuracy of this method is assessed through implementation in the neutron transport code DeCART. The method is employed with variable-order local source representation to model a TWIGL transient. The results demonstrate that this method is accurate and more efficient than the discretized method. (authors)
International Nuclear Information System (INIS)
Fyulep, M.; Nikodemova, D.; Grabovtsova, A.; Galan, P.; Trousil, J.
1977-01-01
Possibilities of the application of multispherical spectrometry in personnel dosimetry of neutrons (n) and gamma radiation (γ) are considered. Studies were made to elucidate a possibility of using albedo dosemeters to increase the sensitivity of personnel dosemeters. Determined were the dose due to the (n,γ) reaction in a human body, absorbed dose and dose equivalent. The effect of (n,γ) dose on the reading of personnel gamma dosemeter was considered. It is shown that the above effect on the dosemeter readings for frontal irradiation by a broad neutron beam in everyday personnel dosimetry near 252 Cf sources may be neglected. Only in the case of strongly slowed-down fission spectrum the effect of the (n,γ) reaction is considerable. The application of albedo dosemeter is expedient to take into account the corrections of personnel dosemeter readings [ru
Neutron image intensifier tubes
International Nuclear Information System (INIS)
Verat, M.; Rougeot, H.; Driard, B.
1983-01-01
The most frequently used techniques in neutron radiography employ a neutron converter consisting of either a scintillator or a thin metal sheet. The radiation created by the neutrons exposes a photographic film that is in contact with the converter: in the direct method, the film is exposed during the time that the object is irradiated with neutrons; in the transfer method, the film is exposed after the irradiation of the object with neutrons. In industrial non-destructive testing, when many identical objects have to be checked, these techniques have several disadvantages. Non-destructive testing systems without these disadvantages can be constructed around neutron-image intensifier tubes. A description and the operating characteristics of neutron-image intensifier tubes are given. (Auth.)
Directory of Open Access Journals (Sweden)
Pin Gong
2018-04-01
Full Text Available With a highly functional methyl vinyl silicone rubber (VMQ matrix and filler materials of B4C, PbO, and benzophenone (BP and through powder surface modification, silicone rubber mixing, and vulcanized molding, a flexible radiation shielding and resistant composite was prepared in the study. The dispersion property of the powder in the matrix filler was improved by powder surface modification. BP was added into the matrix to enhance the radiation resistance performance of the composites. After irradiation, the tensile strength, elongation, and tear strength of the composites decreased, while the Shore hardness of the composites and the crosslinking density of the VMQ matrix increased. Moreover, the composites with BP showed better mechanical properties and smaller crosslinking density than those without BP after irradiation. The initial degradation temperatures of the composites containing BP before and after irradiation were 323.6°C and 335.3°C, respectively. The transmission of neutrons for a 2-mm thick sample was only 0.12 for an Am–Be neutron source. The transmission of γ-rays with energies of 0.662, 1.173, and 1.332 MeV for 2-cm thick samples were 0.7, 0.782, and 0.795, respectively. Keywords: Flexible Composite, Neutron Shielding, Radiation Resistance, γ-ray Shielding
International Nuclear Information System (INIS)
Robinson, L.; Zhao, L.
2009-01-01
Previous studies including the development of methods for the determination of carbon, nitrogen, and phosphorus in cattail using cold neutron prompt gamma activation (CNPGAA) and thermal neutron prompt gamma activation analysis (TNPGAA); evaluation of the precision and accuracy of these methods through the analysis of Standard Reference Materials (SRMs); and comparison of the sensitivity of CNPGAA to TNPGAA have been done in the CNPGAA and TNPGAA facilities at the National Institute of Standards and Technology (NIST). This paper integrates the findings from all of these prior studies and presents recommendations for the application of CNPGAA and TNPGAA in environmental studies of plants based on synergistic considerations of the effects of neutron energy, matrix factors such as chlorine content, Compton scattering, hydrogen content, sample thickness, and spectral interferences from Cl on the determination of C, N, and P. This paper also provides a new approach that simulates a sensitivity curve for an element of interest (S), which is a function of hydrogen content (X) and sample thickness (Y) as follows: S = aX + bY + c (where a, b, and c are constants). This approach has provided more accurate results from the analysis of SRMs than traditional methods and an opportunity to use models to optimize experimental conditions. (author)
International Nuclear Information System (INIS)
Williams, W.G.
1988-01-01
The book on 'polarized neutrons' is intended to inform researchers in condensed matter physics and chemistry of the diversity of scientific problems that can be investigated using polarized neutron beams. The contents include chapters on:- neutron polarizers and instrumentation, polarized neutron scattering, neutron polarization analysis experiments and precessing neutron polarization. (U.K.)
Directory of Open Access Journals (Sweden)
Cousin Fabrice
2015-01-01
Full Text Available The specular neutron reflectivity is a technique enabling the measurement of neutron scattering length density profile perpendicular to the plane of a surface or an interface, and thereby the profile of chemical composition. The characteristic sizes that are probed range from around 5 Å up 5000 Å. It is a scattering technique that averages information on the entire surface and it is therefore not possible to obtain information within the plane of the interface. The specific properties of neutrons (possibility of tuning the contrast by isotopic substitution, sensitivity to magnetism, negligible absorption, low energy of the incident neutrons makes it particularly interesting in the fields of soft matter, biophysics and magnetic thin films. This course is a basic introduction to the technique and does not address the magnetic reflectivity. It is composed of three parts describing respectively its principle and its formalism, the experimental aspects of the method (spectrometers, samples and two examples related to the materials for energy.
... cases, it is due to a combination of physical and psychological concerns. Psychological causes of delayed ejaculation include: Depression, anxiety or other mental health conditions Relationship problems due to stress, poor communication ...
... Slow rate of growth; Retarded growth and development; Growth delay Images Toddler development References Cooke DW, Divall SA, Radovick S. Normal and aberrant growth in children. In: Melmed S, Polonsky KS, Larsen PR, ...
International Nuclear Information System (INIS)
Allen, L.S.
1977-01-01
A borehole logging tool includes a steady-state source of fast neutrons, two epithermal neutron detectors, and two thermal neutron detectors. A count rate meter is connected to each neutron detector. A first ratio detector provides an indication of the porosity of the formation surrounding the borehole by determining the ratio of the outputs of the two count rate meters connected to the two epithermal neutron detectors. A second ratio detector provides an indication of both porosity and macroscopic absorption cross section of the formation surrounding the borehole by determining the ratio of the outputs of the two count rate meters connected to the two thermal neutron detectors. By comparing the signals of the two ratio detectors, oil bearing zones and salt water bearing zones within the formation being logged can be distinguished and the amount of oil saturation can be determined. 6 claims, 2 figures
Three dimensional neutronic/thermal-hydraulic coupled simulation of MSR in transient state condition
International Nuclear Information System (INIS)
Zhou, Jianjun; Zhang, Daling; Qiu, Suizheng; Su, Guanghui; Tian, Wenxi; Wu, Yingwei
2015-01-01
Highlights: • Developed a three dimensional neutronic/thermal-hydraulic coupled transient analysis code for MSR. • Investigated the neutron distribution and thermal-hydraulic characters of the core under transient condition. • Analyzed three different transient conditions of inlet temperature drop, reactivity jump and pump coastdown. - Abstract: MSR (molten salt reactor) use liquid molten salt as coolant and fuel solvent, which was the only one liquid reactor of six Generation IV reactor types. As a liquid reactor the physical property of reactor was significantly influenced by fuel salt flow and the conventional analysis methods applied in solid fuel reactors are not applicable for this type of reactors. The present work developed a three dimensional neutronic/thermal-hydraulic coupled code investigated the neutronics and thermo-hydraulics characteristics of the core in transient condition based on neutron diffusion theory and numerical heat transfer. The code consists of two group neutron diffusion equations for fast and thermal neutron fluxes and six group balance equations for delayed neutron precursors. The code was separately validated by neutron benchmark and flow and heat transfer benchmark. Three different transient conditions was analyzed with inlet temperature drop, reactivity jump and pump coastdown. The results provide some valuable information in design and research this kind of reactor
Three dimensional neutronic/thermal-hydraulic coupled simulation of MSR in transient state condition
Energy Technology Data Exchange (ETDEWEB)
Zhou, Jianjun [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China); College of Mechanical and Power Engineering, China Three Gorges University, No 8, Daxue road, Yichang, Hubei 443002 (China); Zhang, Daling, E-mail: dlzhang@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China); Qiu, Suizheng; Su, Guanghui; Tian, Wenxi; Wu, Yingwei [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China)
2015-02-15
Highlights: • Developed a three dimensional neutronic/thermal-hydraulic coupled transient analysis code for MSR. • Investigated the neutron distribution and thermal-hydraulic characters of the core under transient condition. • Analyzed three different transient conditions of inlet temperature drop, reactivity jump and pump coastdown. - Abstract: MSR (molten salt reactor) use liquid molten salt as coolant and fuel solvent, which was the only one liquid reactor of six Generation IV reactor types. As a liquid reactor the physical property of reactor was significantly influenced by fuel salt flow and the conventional analysis methods applied in solid fuel reactors are not applicable for this type of reactors. The present work developed a three dimensional neutronic/thermal-hydraulic coupled code investigated the neutronics and thermo-hydraulics characteristics of the core in transient condition based on neutron diffusion theory and numerical heat transfer. The code consists of two group neutron diffusion equations for fast and thermal neutron fluxes and six group balance equations for delayed neutron precursors. The code was separately validated by neutron benchmark and flow and heat transfer benchmark. Three different transient conditions was analyzed with inlet temperature drop, reactivity jump and pump coastdown. The results provide some valuable information in design and research this kind of reactor.
International Nuclear Information System (INIS)
Anan'in, O.B.; Bespalov, D.F.; Bykovskii, Yu.A.; Kozyrev, Yu.P.; Mints, A.Z.; Riabov, E.V.; Tsybin, A.S.; Cherkasov, Yu.; Shikanov, A.E.
1986-01-01
Information is presented concerning devices for producing intense neutrons flows, and may be utilized in nuclear geophysics for carrying out pulsed neutron logging of wells, in studies of the critical characteristics of nuclear reactors, for activation analysis, radiation therapy, defectoscopy, and so on
An investigation of TRU recycling with various neutron spectrums
International Nuclear Information System (INIS)
Yong-Nam, Kim; Hong-Chul, Kim; Chi-Young, Han; Jong-Kyung, Kim; Won-Seok Park
2003-01-01
This study is intended to evaluate the dependency of TRU recycling characteristics on the neutron spectrum shift in a Pb-Bi cooled core. Considering two Pb-Bi cooled cores with the soft and the hard spectrum, respectively, various characteristics of the recycled core are carefully examined and compared with each other. Assuming very simplified fuel cycle management with the homogeneous and single batch fuel loading, the burn-up calculations are performed until the recycled core reached to the (quasi-) equilibrium state. The mechanism of TRU recycling toward the equilibrium is analysed in terms of burn-up reactivity and the isotopic compositions of TRU fuel. In the comparative analyses, the difference in the recycling behaviour between the two cores is clarified. In addition, the basic safety characteristics of the recycled core are also discussed in terms of the Doppler coefficient, the coolant loss reactivity coefficient, and the effective delayed neutron fraction. (author)
New electronically black neutron detectors
International Nuclear Information System (INIS)
Drake, D.M.; Feldman, W.C.; Hurlbut, C.
1986-03-01
Two neutron detectors are described that can function in a continuous radiation background. Both detectors identify neutrons by recording a proton recoil pulse followed by a characteristic capture pulse. This peculiar signature indicates that the neutron has lost all its energy in the scintillator. Resolutions and efficiencies have been measured for both detectors
International Nuclear Information System (INIS)
Rassow, Juergen; Stecher-Rasmussen, Finn; Voorbraak, Wim; Moss, Ray; Vroegindeweij, Corine; Hideghety, Katalin; Sauerwein, Wolfgang
2001-01-01
Background and purpose: The European Council Directive on health protection 97/43/EURATOM requires radiotherapy quality assurance programmes for performance and safety characteristics including acceptance and repeated tests. For Boron Neutron Capture therapy (BNCT) at the High Flux Reactor (HFR) in Petten/NL such a programme has been developed on the basis of IEC publications for medical electron accelerators. Results: The fundamental differences of clinical dosimetry for medical electron accelerators and BNCT are presented and the order of magnitude of dose components and their stability and that of the main other influencing parameter 10 B concentration for BNCT patient treatments. A comparison is given for requirements for accelerators and BNCT units indicating items which are not transferable, equal or additional. Preliminary results of in vivo measurements done with a set of 55 Mn, 63 Cu and 197 Au activation foils for all single fields for the four fractions at all 15 treated patients show with <±4% up to now a worse reproducibility than the used dose monitoring systems (±1.5%) caused by influence of hair position on the foil-skull distance. Conclusions: Despite the more complex clinical dosimetry (because of four relevant dose components, partly of different linear energy transfer (LET)) BNCT can be regulated following the principles of quality assurance procedures for therapy with medical electron accelerators. The reproducibility of applied neutron fluence (proportional to absorbed doses) and the main safety aspects are equal for all teletherapy methods including BNCT
Energy Technology Data Exchange (ETDEWEB)
Kinashi, Yuko [Research Reactor Institute, Kyoto University, Kumatori-cho, Sennan-gun, Osaka (Japan)], E-mail: kinashi@rri.kyoto-u.ac.jp; Suzuki, Minoru; Masunaga, Shinichiro; Ono, Koji [Research Reactor Institute, Kyoto University, Kumatori-cho, Sennan-gun, Osaka (Japan)
2009-07-15
To investigate bystander mutagenic effects induced by alpha particles during boron neutron capture therapy (BNCT), we mixed cells that were electroporated with borocaptate sodium (BSH), which led to the accumulation of {sup 10}B inside the cells, with cells that did not contain the boron compound. BSH-containing cells were irradiated with {alpha} particles produced by the {sup 10}B(n,{alpha}){sup 7}Li reaction, whereas cells without boron were only affected by the {sup 1}H(n,{gamma}){sup 2}H and {sup 14}N(n,{rho}){sup 14}C reactions. The frequency of mutations induced in the hypoxanthine-guanine phosphoribosyltransferase (HPRT) locus was examined in Chinese hamster ovary (CHO) cells irradiated with neutrons (Kyoto University Research Reactor: 5 MW). Neutron irradiation of 1:1 mixtures of cells with and without BSH resulted in a survival fraction of 0.1, and the cells that did not contain BSH made up 99.4% of the surviving cell population. Using multiplex polymerase chain reactions (PCRs), molecular structural analysis indicated that most of the mutations induced by the bystander effect were point mutations and that the frequencies of total and partial deletions induced by the bystander effect were lower than those resulting from the {alpha} particles produced by the {sup 10}B(n,{alpha}){sup 7}Li reaction or the neutron beam from the {sup 1}H(n,{gamma}){sup 2}H and {sup 14}N(n,{rho}){sup 14}C reactions. The types of point mutations induced by the BNCT bystander effect were analyzed by cloning and sequencing methods. These mutations were comprised of 65.5% base substitutions, 27.5% deletions, and 7.0% insertions. Sequence analysis of base substitutions showed that transversions and transitions occurred in 64.7% and 35.3% of cases, respectively. G:C{yields}T:A transversion induced by 8-oxo-guanine in DNA occurred in 5.9% of base substitution mutants in the BNCT bystander group. The characteristic mutations seen in this group, induced by BNCT {alpha} particles
International Nuclear Information System (INIS)
Steenstrup, S.
Briefly surveys recent developments in research work with ultracold neutrons (neutrons of very low velocity, up to 10 m/s at up to 10 -7 eV and 10 -3 K). Slow neutrons can be detected in an ionisation chamber filled with B 10 F 3 . Very slow neutrons can be used for investigations into the dipole moment of neutrons. Neutrons of large wave length have properties similar to those of light. The limit angle for total reflection is governed by the wave length and by the material. Total reflection can be used to filter ultracold neutrons out of the moderator material of a reactor. Total reflection can also be used to store ultracold neutrons but certain problems with storage have not yet been clarified. Slow neutrons can be made to lose speed in a neutron turbine, and come out as ultracold neutrons. A beam of ultracold neutrons could be used in a neutron microscope. (J.S.)
Kolotkov, Gennady A.; Penin, Sergei
2017-11-01
The paper examines an update of comparative analysis of radionuclides released into the atmosphere from Beloyarsk nuclear power plant with fast-neutron reactor for nine years in a row, from 2008 to 2016. It has been shown that the main radionuclides throw out into the atmosphere from Beloyarsk nuclear power plant are beta-active radionuclides. Based on data releases of the RPA "Typhoon", it has been conclude that radiation situation become worse insignificantly; beside on the new reactor BN-800 was put in operation in 2016. Using Spencer-Fano's equation, it was carried out the summary spectrum of emitted radionuclides. On example of Beloyarsk nuclear power plant, it was considered a question about ability of remote detection of raised radioactivity in the atmospheric radioactive plume. It has been shown that it possible to detect raised radioactivity in the emission plume from Beloyarsk nuclear power plant.
Energy Technology Data Exchange (ETDEWEB)
Blokhin, A.I.; Ivanov, A.P.; Korobeinikov, V.V.; Lunev, V.P.; Manokhin, V.N.; Khorasanov, G.L. [SSC RF A. I. Leypunsky Institute for Physics and Power Engineering, Obninsk, Kaluga Region (Russian Federation)
2000-03-01
A new type of safe fast reactor with lead coolant was proposed in Russia. The use of coolants with low moderating properties is one of the ways to get a hard neutron spectrum and an increase in the burning of Np-237, Am-243 and other miner actinides(MA) fissionable preferentially in the fast reactor. The stable lead isotope, Pb-208, is proposed as the one of such coolants. The neutron inelastic scattering cross-section of Pb-208 is 3.0-3.5 times less than the one of other lead isotopes. Calculation of the MA transmutation rates in the standard BN-type fast reactor with different coolants is performed by Monte-Carlo method using Code MMKFK. Six various models are simulated for the fast reactor blanket with different kinds of fuel and coolant. The fast reactor with natural-lead coolant practically does not differ from the reactor with sodium coolant relative to MA incineration. The use of Pb-208 as a coolant in the fast reactor results in increasing incineration of MA from 18 to 26% in comparison with a usual fast reactor. Calculation of induced radioactivity was performed using the FISPACT-3 inventory code, also. The results include total induced radioactivity and dose rate for initial material composition and selected long-lived radionuclides. The calculations show that the coolant consisting of lead isotope, Pb-206, or Pb-207, can be considered as the low-activation one because it does not practically contain long-lived toxic radionuclides. (M. Suetake)
Energy Technology Data Exchange (ETDEWEB)
Gonnelli, Eduardo; Diniz, Ricardo [Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP Travessa R-400, 05508-900, Cidade Universitaria, Sao Paulo (Brazil)
2013-05-06
The neutron lifetimes of the core, reflector, and global were experimentally obtained through macroscopic neutron noise in the IPEN/MB-01 reactor for five levels of subcriticality. The theoretical Auto Power Spectral Densities were derived by point kinetic equations taking the reflector effect into account, and one of the approaches consider an additional group of delayed neutrons.
Nutrition support is essential for the care of the child with developmental delay. After a thorough evaluation, an individualized intervention plan that accounts for the child’s nutrition status, feeding ability, and medical condition may be determined. Nutrition assessments may be performed at leas...
International Nuclear Information System (INIS)
Mazumdar, Tanay; Degweker, S.B.
2017-01-01
Highlights: • In Method of Characteristics, the neutron source within a mesh is expanded up to linear term. • This expansion reduces the number of meshes as compared to flat source assumption. • Poor representation of circular geometry with coarser meshes is corrected. • Few benchmark problems are solved to show the advantages of linear expansion of source. • The advantage of the present formalism is quite visible in problems with large flux gradient. - Abstract: A common assumption in the solution of the neutron transport equation by the Method of Characteristics (MOC) is that the source (or flux) is constant within a mesh. This assumption is adequate provided the meshes are small enough so that the spatial variation of flux within a mesh may be ignored. Whether a mesh is small enough or not depends upon the flux gradient across a mesh, which in turn depends on factors like the presence of strong absorbers, localized sources or vacuum boundaries. The flat flux assumption often requires a very large number of meshes for solving the neutron transport equation with acceptable accuracy as was observed in our earlier work on the subject. A significant reduction in the required number of meshes is attainable by using a higher order representation of the flux within a mesh. In this paper, we expand the source within a mesh up to first order (linear) terms, which permits the use of larger sized (and therefore fewer) meshes and thereby reduces the computation time without compromising the accuracy of calculation. Since the division of the geometry into meshes is through an automatic triangulation procedure using the Bowyer-Watson algorithm, representation of circular objects (cylindrical fuel rods) with coarse meshes is poorer and causes geometry related errors. A numerical recipe is presented to make a correction to the automatic triangulation process and thereby eliminate this source of error. A number of benchmark problems are analyzed to emphasize the
Neutrons in the field of metallurgy
International Nuclear Information System (INIS)
Novion, C. de
1989-01-01
Beams of thermal neutrons are now widely used for the study of material structure. Following a summary of the characteristics of the neutron-material interaction, and an outlook on the major uses of neutrons in metallurgy, we present some examples of application. The comparative advantages and drawbacks of neutrons and X-rays are discussed. 14 refs [fr
Stability and delay sensitivity of neutral fractional-delay systems.
Xu, Qi; Shi, Min; Wang, Zaihua
2016-08-01
This paper generalizes the stability test method via integral estimation for integer-order neutral time-delay systems to neutral fractional-delay systems. The key step in stability test is the calculation of the number of unstable characteristic roots that is described by a definite integral over an interval from zero to a sufficient large upper limit. Algorithms for correctly estimating the upper limits of the integral are given in two concise ways, parameter dependent or independent. A special feature of the proposed method is that it judges the stability of fractional-delay systems simply by using rough integral estimation. Meanwhile, the paper shows that for some neutral fractional-delay systems, the stability is extremely sensitive to the change of time delays. Examples are given for demonstrating the proposed method as well as the delay sensitivity.
Energy Technology Data Exchange (ETDEWEB)
Chung, Y. S.; Moon, J. H.; Kim, S. H.; Sun, G. M.; Baek, S. Y.; Kim, H. R.; Kim, Y. J
2008-03-15
A pneumatic transfer irradiation system (PTS) is one of the most important facilities used during neutron irradiation of a target material for instrumental neutron activation analysis (INAA) in a research reactor. In particular, a fast pneumatic transfer system is essential for the measurement of a short half-life nuclide and a delayed neutron counting system. The pneumatic transfer irradiation system (PTS no.2) involving a manual system and an automatic system for delayed neutron activation analysis (DNAA) were reconstructed with new designs of a functional improvement at the HANARO research reactor in 2006. In this technical report, the conception, design, operation and control of PTS no.2 was described. Also the experimental results and the characteristic parameters measured by a mock-up test, a functional operation test and an irradiation test of these systems, such as the transfer time of irradiation capsule, automatic operation control by personal computer, delayed neutron counting system, the different neutron flux, the temperature of the irradiation position with an irradiation time, the radiation dose rate when the rabbit is returned, etc. are reported to provide a user information as well as a reactor's management and safety.
International Nuclear Information System (INIS)
Fukumori, Yoshinobu; Ichikawa, Hideki; Nakatani, Yugo
2006-01-01
Preparation of Gd-nanoCPs was carried out based on the w/o emulsion-droplet coalescence technique. Chitosan with different molecular weight (Mw) (950, 50, 10 kDa) was applied a various concentrations (0.5, 1.5, 2.5%). B16F10 mouse melanoma cells were employed to evaluate the cell-association properties of Gd-nanoCPs and the antitumor effect with thermal neutron irradiation. In the formulation study, as the Mw of chitosan decreased, the mean particle diameter decreased to 155 nm with 10 kDa chitosan at the smallest. Then, decrease in the chitosan concentration in the chitosan solution contributed to a decrease in mean particle diameter and an increase in gadolinium content. The Gd content in the Gd-nanoCPs prepared with 0.5% chitosan reached 22% at the maximum. The amount of Gd associated with cell was also dependent on the Mw of chitosan. In the subsequent Gd-NCT study in vitro, the most highly Gd-containing (22%) and finest (155 nm) Gd-nanoCPs prepared at 0.5% of 10 kDa chitosan exhibited the strongest tumor cell growth suppression. It is expected that use of Gd-nanoCP prepared here may lead to improved performance in NCT. (author)
Study on self-regulation characteristics of closed two-phase thermo-siphon for cold neutron source
International Nuclear Information System (INIS)
Shen Feng; Yuan Luzheng
2006-01-01
A self-regulation model and its characteristics of closed two-phase thermosiphon loop, which including buffer tank, were proposed, under the conditions of adiabatic and transient temperature balance for connect tube between buffer tank and condenser. The comparison between these models and the model from Kyoto University is conducted. Measures to improve the self-regulation are proposed. (authors)
Investigation of the characteristics of 252Cf-detectors
International Nuclear Information System (INIS)
Karlsson, Erik
2004-12-01
In the first chapter the characteristic behaviors of two Cf detectors have been investigated by performing pilot measurements. The detector with the stronger source gives an unstable signal with a low signal/noise ratio. Therefore this detector has not been further investigated. The ionization chamber reacts on both fission products and alpha decay. An energy experiment showed that there were large difficulties to separate those decays. A plastic scintillator, which reacts on both photons and neutrons, was used for neutron detection. Energy spectrums were performed and the result showed that it is difficult to set an energy threshold to separate the neutrons and the photons. The discrimination will rather be achieved by time of flight methods which is discussed under the second chapter in this thesis; Experimental results. An other experiment was done in order to investigate whether it is possible to detect any delayed components from the spontaneous fission of Cf. The result showed that delayed components existed. Either they are delayed neutrons from exited fission products, or it is some delay related to the charge collection in the Cf detector. Correlation measurements showed that few events are coincident. Only 50% of the signals from the plastic scintillator are correlated with the Cf source
Neutrons and Nuclear Engineering
International Nuclear Information System (INIS)
Ekkebus, Allen E.
2007-01-01
Oak Ridge National Laboratory hosted two workshops in April 2007 relevant to nuclear engineering education. In the Neutron Stress, Texture, and Phase Transformation for Industry workshop (http://neutrons.ornl.gov/workshops/nst2/), several invited speakers gave examples of neutron stress mapping for nuclear engineering applications. These included John Root of National Research Council of Canada, Mike Fitzpatrick of the UK's Open University, and Yan Gao of GE Global Research on their experiences with industrial and academic uses of neutron diffraction. Xun-Li Wang and Camden Hubbard described the new instruments at ORNL that can be used for such studies. This was preceded by the Neutrons for Materials Science and Engineering educational symposium (http://neutrons.ornl.gov/workshops/edsym2007). It was directed to the broad materials science and engineering community based in universities, industry and laboratories who wish to learn what the neutron sources in the US can provide for enhancing the understanding of materials behavior, processing and joining. Of particular interest was the presentation of Donald Brown of Los Alamos about using 'Neutron diffraction measurements of strain and texture to study mechanical behavior of structural materials.' At both workshops, the ORNL neutron scattering instruments relevant to nuclear engineering studies were described. The Neutron Residual Stress Mapping Facility (NRSF2) is currently in operation at the High Flux Isotope Reactor; the VULCAN Engineering Materials Diffractometer will begin commissioning in 2008 at the Spallation Neutron Source. For characteristics of these instruments, as well as details of other workshops, meetings, capabilities, and research proposal submissions, please visit http://neutrons.ornl.gov. To submit user proposals for time on NRSF2 contact Hubbard at hubbardcratornl.gov
Microscopic Control Delay Modeling at Signalized Arterials Using Bluetooth Technology
Rajasekhar, Lakshmi
2011-01-01
Real-time control delay estimation is an important performance measure for any intersection to improve the signal timing plans dynamically in real-time and hence improve the overall system performance. Control delay estimates helps to determine the level-of-service (LOS) characteristics of various approaches at an intersection and takes into account deceleration delay, stopped delay and acceleration delay. All kinds of traffic delay calculation especially control delay calculation has always ...
Neutron radiography with the cyclotron
International Nuclear Information System (INIS)
Tazawa, Shuichi; Asada, Yorihisa; Yano, Munehiko; Nakanii, Takehiko.
1985-01-01
Neutron radiography is well recognized as a powerful tool in nondestructive testing, but not widely used yet owing to lack of high intense thermal neutron source convenient for practical use. This article presents a new neutron radiograph facility, utilizing a sub-compact cyclotron as neutron source and is equipped with vertical and horizontal irradiation ports. The article describes a series of experiments, we conducted using beams of a variable energy cyclotron at Tohoku University to investigate the characteristics of thermal neutron obtained from 9 Be(p, n) reaction and thermalized by elastic scattering process. The article also describes a computer simulation of neutron moderator to analyze conditions getting maximal thermal neutron flux. Further, some of practical neutron radiograph examinations of aero-space components and museum art objects of classic bronze mirror and an attempt realizing real time imaging technique, are introduced in the article. (author)
Neutron activation analysis at the 'Instituto de Pesquisas Energeticas e Nucleares' (SP, Brazil)
International Nuclear Information System (INIS)
Vasconcellos, M.B.A.
1984-01-01
A review of the work carried out at IPEN using neutron activation analysis is made. The main characteristics of the technique and general experimental procedures applied for different samples and elements are reported. Geological samples were analysed by using activation with thermal, epithermal and delayed neutrons (for U and Th, specifically). Metallic samples were analysed for several elements in trace amounts (Ta in Nb, Hg in steel, Sn, Sb, As, Cu, Cr and Ag in a tin-lead alloy). Biological materials, such as tomatoes, animal and human viscera, food, hair, nails were also analysed for several components (Hg, Na, K, As, Au and others). (Author) [pt
DEFF Research Database (Denmark)
Kolby, Nanna; Busch, Alexander Siegfried; Juul, Anders
2017-01-01
. The underlying reasons for the large variation in the age at pubertal onset are not fully established; however, nutritional status and socioeconomic and environmental factors are known to be influencing, and a significant amount of influencing genetic factors have also been identified. The challenges...... optimal in discriminating especially CDGP from HH. Management of the delayed puberty depends on the etiology. For boys with CDGP an observational period will often reveal imminent puberty. If puberty is not progressing spontaneously, sex steroid replacement is effective in stimulating the development...
International Nuclear Information System (INIS)
Van Well, A.A.
1999-01-01
Neutron research where reflection, refraction, and interference play an essential role is generally referred to as 'neutron optics'. The neutron wavelength, the scattering length density and the magnetic properties of the material determine the critical angle for total reflection. The theoretical background of neutron reflection, experimental methods and the interpretation of reflection data are presented. (K.A.)
A review on neutron reflectometry
Energy Technology Data Exchange (ETDEWEB)
Lee, Jeong Soo; Lee, Chang Hee; Shim, Hae Seop; Seong, Baek Seok
1999-03-01
This report contains principle and characteristic of neutron reflectometry. Therefore, in case of operating neutron reflectometer at HANARO in future, it will be a reference to the user who wishes to use the instrument effectively. Also, the current situation of neutron reflectometer operating in the world was examined. The detail of neutron reflectometer such as GANS(MURR), ADAM(ILL), POSY II(ANL), ROG(IRI) was described. The recent research situation on neutron reflectometry was also examined and it helps us to determine research field. (author)
Changes in the long-term delayed response of platinum self-powered detector with irradiation
International Nuclear Information System (INIS)
Parent, G.; Serdula, K.J.; Eng, P.
1989-01-01
Two long-term delayed response characteristics have been observed for platinum, Pt, detectors in the Gentilly-2 600 MW(e) CANDU PHWR reactor. The first effect is a dip in the signal two to three hours after a shutdown, due to the (n,beta) interactions of Mn-55 and Ni-64 which exist as impurities in the detector assembly. The second effect is an increase of the delayed fraction of the signal. The low neutron absorption cross-section of Pt-196 combined with the conversion of the Pt-194 and Pt-195 results in build-up of the Pt-196. The long half-lives associated with the beta-emission in the transmutation of Pt-196 to Hg-198 or Hg-199 give rise to the observed long-term delayed response
Neutron scattering in Australia
International Nuclear Information System (INIS)
Knott, R.B.
1994-01-01
Neutron scattering techniques have been part of the Australian scientific research community for the past three decades. The High Flux Australian Reactor (HIFAR) is a multi-use facility of modest performance that provides the only neutron source in the country suitable for neutron scattering. The limitations of HIFAR have been recognized and recently a Government initiated inquiry sought to evaluate the future needs of a neutron source. In essence, the inquiry suggested that a delay of several years would enable a number of key issues to be resolved, and therefore a more appropriate decision made. In the meantime, use of the present source is being optimized, and where necessary research is being undertaken at major overseas neutron facilities either on a formal or informal basis. Australia has, at present, a formal agreement with the Rutherford Appleton Laboratory (UK) for access to the spallation source ISIS. Various aspects of neutron scattering have been implemented on HIFAR, including investigations of the structure of biological relevant molecules. One aspect of these investigations will be presented. Preliminary results from a study of the interaction of the immunosuppressant drug, cyclosporin-A, with reconstituted membranes suggest that the hydrophobic drug interdigitated with lipid chains
Neutron scattering in Australia
Energy Technology Data Exchange (ETDEWEB)
Knott, R.B. [Australian Nuclear Science and Technology Organisation, Menai (Australia)
1994-12-31
Neutron scattering techniques have been part of the Australian scientific research community for the past three decades. The High Flux Australian Reactor (HIFAR) is a multi-use facility of modest performance that provides the only neutron source in the country suitable for neutron scattering. The limitations of HIFAR have been recognized and recently a Government initiated inquiry sought to evaluate the future needs of a neutron source. In essence, the inquiry suggested that a delay of several years would enable a number of key issues to be resolved, and therefore a more appropriate decision made. In the meantime, use of the present source is being optimized, and where necessary research is being undertaken at major overseas neutron facilities either on a formal or informal basis. Australia has, at present, a formal agreement with the Rutherford Appleton Laboratory (UK) for access to the spallation source ISIS. Various aspects of neutron scattering have been implemented on HIFAR, including investigations of the structure of biological relevant molecules. One aspect of these investigations will be presented. Preliminary results from a study of the interaction of the immunosuppressant drug, cyclosporin-A, with reconstituted membranes suggest that the hydrophobic drug interdigitated with lipid chains.
Experiment of Neutron Generation by Using Prototype D-D Neutron Generator
International Nuclear Information System (INIS)
Kim, In Jung; Kim, Suk Kwon; Park, Chang Su; Jung, Nam Suk; Jung, Hwa Dong; Park, Ji Young; Hwang, Yong Seok; Choi, H.D.
2005-01-01
Experiment of neutron generation was performed by using a prototype D-D neutron generator. The characteristics of D-D neutron generation in drive-in target was studied. The increment of neutron yield by increasing ion beam energy was investigated, too
Self powered neutron detectors
International Nuclear Information System (INIS)
Passe, J.; Petitcolas, H.; Verdant, R.
1975-01-01
The self-powered neutron detectors (SPND) enable to measure continuously high fluxes of thermal neutrons. They are particularly suitable for power reactor cores because of their robustness. Description of two kinds of SPND's characterized by the electrical current production way is given here: the first SPND's which present a V, Ag or Rh emitter are sensitive enough but they offer a few minute delay time: the second SPND's which are depending on the gamma activation have a short delay time. The emitter is made of Co or Pt. In any case, the signal is linear with reaction rates. Finally, the applications are briefly repeated here: irradiation facility monitor in research reactors, and flux map and space instability control in power reactors [fr
Neutron radiography with ultracold neutrons
International Nuclear Information System (INIS)
Bates, J.C.
1981-01-01
The neutron transmission factor of very thin films may be low if the neutron energy is comparable to the pseudo-potential of the film material. Surprisingly, perhaps, it is relatively easy to obtain neutrons with such low energies in sufficient numbers to produce neutron radiographs. (orig.)
Neutron lifetime well logging methods and apparatus
International Nuclear Information System (INIS)
Paap, H.J.; Pitts, R.W.
1974-01-01
A method for investigating the earth formations surrounding a well borehole, comprising the steps of: continuously generating high energy neutrons in the borehole and bombarding the surrounding media with such neutrons to develop a cloud of thermal neutrons therein; modulating the intensity of said high energy neutrons harmonically as a function of time in order to intensity modulate said cloud of thermal neutrons as a function of time; and measuring a time-dependant thermal neutron characteristic of said intensity modulated cloud of thermal neutrons
Neutron Skins and Neutron Stars
Piekarewicz, J.
2013-01-01
The neutron-skin thickness of heavy nuclei provides a fundamental link to the equation of state of neutron-rich matter, and hence to the properties of neutron stars. The Lead Radius Experiment ("PREX") at Jefferson Laboratory has recently provided the first model-independence evidence on the existence of a neutron-rich skin in 208Pb. In this contribution we examine how the increased accuracy in the determination of neutron skins expected from the commissioning of intense polarized electron be...
DEFF Research Database (Denmark)
HANPINITSAK, Panawit; SAITO, Kentaro; Fan, Wei
2017-01-01
Comparison of channel characteristics at many frequency bands is necessary to study the frequency-de-pendency which is important for consistent multi-frequency spatial-temporal channel model. Path loss (PL) and power spectrum characteristics of the channel measured in a typical classroom line...
International Nuclear Information System (INIS)
Lim, Jong Myoung; Lee, Jin Hong; Chung, Yong Sam
2005-01-01
An instrumental neutron activation analysis was used to measure the concentrations of about 24 elements associated with airborne particulates (PM10) that were collected in the most polluted urban region of Daejeon city, Korea from 2000 to 2002. Using the measurement data for various elements, both the extent of elemental pollution in the study area and the seasonality in their distribution characteristics were examined. Examinations of their distribution patterns indicated that most elements with crustal origin tend to exhibit seasonal peaks during spring, while most elements with anthropogenic origin tend to exhibit seasonal peaks during fall or winter. In order to explain the factors regulating their mobilization properties, the data were processed by a factor analysis. Results of the factor analysis suggested competing roles of both industrial and natural source processes, despite that the study site is located at a downwind position of the industrial complex. Based on the overall results of this study, it is concluded that the site may be strongly impacted by man-made sources but the general patterns of elemental distributions in the study area inspected over a seasonal scale are quite consistent with those typically observed from natural environment
Energy Technology Data Exchange (ETDEWEB)
Godar, Serge [Communaute europeenne de l' energie atomique - Euratom, Brussels (Belgium)
1964-08-15
This report indicates characteristics of available accelerators adapted to neutron production. Indicated information are for example: brand and type, brief description, maximum high voltage, ion source, ion current, gas introduction system, titanium self-target, beam focusing system, target diameter, reaction used for neutron production, target cooling, generator use temperature, dimensions, weight, manufacturer address, and so on [French] L'ensemble de cette etude bibliographique est traitee dans une serie de huit rapports, chacun ayant trait a un aspect ou un domaine particulier. Ce rapport (No. 7-9) a pour but de donner des renseignements generaux aux chercheurs desirant acquerir un accelerateur de particules adapte a la production de neutrons. Les recherches concernant les caracteristiques des appareils actuellement sur le marche ont ete arretees a la date du 15 octobre 1963. Il n'est malheureusement pas possible de garantir toutes les caracteristiques fournies, etant donne les modifications continuelles apportees a ces generateurs.
Computer-automated neutron activation analysis system
International Nuclear Information System (INIS)
Minor, M.M.; Garcia, S.R.
1983-01-01
An automated delayed neutron counting and instrumental neutron activation analysis system has been developed at Los Alamos National Laboratory's Omega West Reactor (OWR) to analyze samples for uranium and 31 additional elements with a maximum throughput of 400 samples per day. 5 references
Energy Technology Data Exchange (ETDEWEB)
Park, Yong Joon; Song, Byung Chul; Im, Hee-Jung [Nuclear Chemistry Research Division, Korea Atomic Energy Research Institute, Dukjin-dong 150-1, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Kim, Jong-Yun [Nuclear Chemistry Research Division, Korea Atomic Energy Research Institute, Dukjin-dong 150-1, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)], E-mail: kjy@kaeri.re.kr
2009-07-21
A new prompt gamma-ray activation analysis (PGAA) system equipped with a compact deuterium-deuterium (D-D) neutron generator has been developed for fast detection of explosives and chemical warfare agents. The PGAA system was built based on a fully high-voltage-shielded, axial D-D neutron generator with a radio frequency (RF)-driven ion source. The ionic current of the compact neutron generator was determined as a function of the acceleration voltage at various RF powers. Monoenergetic neutrons (2.45 MeV) with a neutron yield of >1x10{sup 7} n/s were obtained at a deuterium pressure of 8.0 mTorr, an acceleration voltage of 80 kV, and an RF power of 1.1 kW. The performance of the PGAA system was examined by studying the dependence of a prompt gamma-ray count rate on crucial operating parameters.
International Nuclear Information System (INIS)
Sciannandrone, Daniele
2015-01-01
The topic of our research is the application of the Method of Long Characteristics (MOC) to solve the Neutron Transport Equation in three-dimensional axial geometries. The strength of the MOC is in its precision and versatility. As a drawback, it requires a large amount of computational resources. This problem is even more severe in three dimensional geometries, for which unknowns reach the order of tens of billions for assembly-level calculations. The first part of the research has dealt with the development of optimized tracking and reconstruction techniques which take advantage of the regularities of three-dimensional axial geometries. These methods have allowed a strong reduction of the memory requirements and a reduction of the execution time of the MOC calculation. The convergence of the iterative scheme has been accelerated with a lower order transport operator (DPN) which is used for the initialization of the solution and for solving the synthetic problem during MOC iterations. The algorithms for the construction and solution of the MOC and DPN operators have been accelerated by using shared-memory parallel paradigms which are more suitable for standard desktop working stations. An important part of this research has been devoted to the implementation of scheduling techniques to improve the parallel efficiency. The convergence of the angular quadrature formula for three-dimensional cases is also studied. Some of these formulas take advantage of the reduced computational costs of the treatment of planar directions and the vertical direction to speed up the algorithm. The verification of the MOC solver has been done by comparing results with continuous-in-energy Monte Carlo calculations. For this purpose a coupling of the 3D MOC solver with the Subgroup method is proposed to take into account the effects of cross sections resonances. The full calculation of a FBR assembly requires about 2 h of execution time with differences of few pcm with respect to the
Spatial neutron kinetic module of ROSA code
International Nuclear Information System (INIS)
Cherezov, A.L.; Shchukin, N.V.
2009-01-01
A spatial neutron kinetic module was developed for computer code ROSA. The paper describes a numerical scheme used in the module for resolving neutron kinetic equations. Analytical integration for delayed neutrons emitters method and direct numerical integration method (Gear's method) were analyzed. The two methods were compared on their efficiency and accuracy. Both methods were verified with test problems. The results obtained in the verification studies were presented [ru
Energy Technology Data Exchange (ETDEWEB)
Sari, A., E-mail: adrien.sari@cea.fr [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, 91191 Gif-sur-Yvette Cedex (France); Carrel, F.; Lainé, F. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, 91191 Gif-sur-Yvette Cedex (France); Lyoussi, A. [CEA, DEN, 13108 Saint-Paul-Lez-Durance Cedex (France)
2013-10-01
In this article, we demonstrate the feasibility of neutron interrogation using the conversion target of a 17 MeV linear electron accelerator as a neutron generator. Signals from prompt neutrons, delayed neutrons, and delayed gamma-rays, emitted by both uranium and plutonium samples were analyzed. First results from photon and neutron interrogation non-simultaneous measurements combination are also reported in this paper. Feasibility of this technique is shown in the frame of the measurement of uranium enrichment. The latter was carried out by combining detection of prompt neutrons from thermal fission and delayed neutrons from photofission, and by combining delayed gamma-rays from thermal fission and delayed gamma-rays from photofission.
Radionuclide 252Cf neutron source
International Nuclear Information System (INIS)
Kolevatov, Yu.I.; Trykov, L.A.
1979-01-01
Characteristics of radionuclide neutron sourses of 252 Cf base with the activity from 10 6 to 10 9 n/s have been investigated. Energetic distributions of neutrons and gamma-radiation have been presented. The results obtained have been compared with other data available. The hardness parameter of the neutron spectrum for the energy range from 3 to 15 MeV is 1.4 +- 0.02 MeV
Dynamic behaviour and neutron noise in molten salt reactors with circulating perturbations
Energy Technology Data Exchange (ETDEWEB)
Pazsit, I.; Dykin, V. [Chalmers Univ. of Tech., Nuclear Engineering, Goteborg (Sweden)
2014-07-01
This paper concerns the calculation of the neutron noise induced in Molten Salt Reactors (MSR) by the random fluctuations in space and time of the molten fuel cross sections which travel together with the fuel and pass the core region. The effect of such fluctuations was already discussed in several publications. The novelty of the present paper is that it takes into account that in addition to the delayed neutron precursors, also the cross section perturbations themselves, whose passing through the core induces the in-core neutron noise, return to the core inlet via the external loop from the core exit. The corresponding theory is developed, and some quantitative investigations are made of the characteristics of the noise, which can be attributed to the recirculation of the perturbation to the core. It is shown that the effect of the returning of the perturbations, even though it is also associated with a temporal decay, has a much stronger effect on the neutron noise spectra than that of the recirculation of the delayed neutron precursors. (author)
Dynamic behaviour and neutron noise in molten salt reactors with circulating perturbations
International Nuclear Information System (INIS)
Pazsit, I.; Dykin, V.
2014-01-01
This paper concerns the calculation of the neutron noise induced in Molten Salt Reactors (MSR) by the random fluctuations in space and time of the molten fuel cross sections which travel together with the fuel and pass the core region. The effect of such fluctuations was already discussed in several publications. The novelty of the present paper is that it takes into account that in addition to the delayed neutron precursors, also the cross section perturbations themselves, whose passing through the core induces the in-core neutron noise, return to the core inlet via the external loop from the core exit. The corresponding theory is developed, and some quantitative investigations are made of the characteristics of the noise, which can be attributed to the recirculation of the perturbation to the core. It is shown that the effect of the returning of the perturbations, even though it is also associated with a temporal decay, has a much stronger effect on the neutron noise spectra than that of the recirculation of the delayed neutron precursors. (author)
International Nuclear Information System (INIS)
Hrdlicka, Z.
1977-01-01
Neutron radiography is a radiographic method using a neutron beam of a defined geometry. The neutron source usually consists of a research reactor, a specialized neutron radiography reactor or the 252 Cf radioisotope source. There are two types of the neutron radiography display system, viz., a system producing neutron radiography images by a photographic process or a system allowing a visual display, eg., using a television monitor. The method can be used wherever X-ray radiography is used except applications in the radiography of humans. The neutron radiography unit at UJV uses the WWR-S reactor as the neutron source and both types of the above mentioned display system. (J.P.)
Characteristics of poly- and mono-crystalline BeO and SiO{sub 2} as thermal and cold neutron filters
Energy Technology Data Exchange (ETDEWEB)
Adib, M.; Habib, N. [Reactor Physics Department, NRC, Atomic Energy Authority, Cairo (Egypt); Bashter, I.I. [Physics Department, Faculty of Science, Zagazig University (Egypt); Morcos, H.N.; El-Mesiry, M.S. [Reactor Physics Department, NRC, Atomic Energy Authority, Cairo (Egypt); Mansy, M.S., E-mail: drmohamedmansy88@hotmail.com [Reactor Physics Department, NRC, Atomic Energy Authority, Cairo (Egypt)
2015-09-01
Highlights: • Neutron filtering features of BeO and SiO{sub 2} poly- and mono-crystals. • Calculations of the cold and thermal neutron cross sections and transmission with the code “HEXA-FILTERS”. • Optimal mosaic spread, thicknesses and cutting planes for BeO and SiO{sub 2} mono-crystals. - Abstract: A simple model along with a computer code “HEXA-FILTERS” is used to carry out the calculation of the total cross-sections of BeO and SiO{sub 2} having poly or mono-crystalline form as a function of neutron wavelength at room (R.T.) and liquid nitrogen (L.N.) temperatures. An overall agreement is indicated between the calculated neutron cross-sections and experimental data. Calculation shows that 25 cm thick of polycrystalline BeO cooled at liquid nitrogen temperature was found to be a good filter for neutron wavelengths longer than 0.46 nm. While, 50 cm of SiO{sub 2}, with much less transmission, for neutrons with wavelengths longer than 0.85 nm. It was also found that 10 cm of BeO and 15 cm SiO{sub 2} thick mono-crystals cut along their (0 0 2) plane, with 0.5° FWHM on mosaic spread and cooled at L.N., are a good thermal neutron filter, with high effect-to-noise ratio.
International Nuclear Information System (INIS)
Kredov, B.M.
1979-01-01
The history of the neutron is displayed on the basis of contributions by scientists who produced outstanding results in neutron research (part 1), of summarizing discoveries and theories which led to the discovery of the neutron and the resulting development of nuclear physics (part 2), and of fundamental papers written by Rutherford, Chadwick, Iwanenko, and others (appendix). Of interest to physicists, historians, and students
International Nuclear Information System (INIS)
Charlton, J.S.
1986-01-01
The way in which neutrons interact with matter such as slowing-down, diffusion, neutron absorption and moderation are described. The use of neutron techniques in industry, in moisture gages, level and interface measurements, the detection of blockages, boron analysis in ore feedstock and industrial radiography are discussed. (author)
Characterization of Deuteron-Deuteron Neutron Generators
Waltz, Cory Scott
A facility based on a next-generation, high-flux D-D neutron generator (HFNG) was commissioned at the University of California Berkeley. The characterization of the HFNG is presented in the following study. The current generator design produces near mono-energetic 2.45 MeV neutrons at outputs of 108 n/s. Calculations provided show that future conditioning at higher currents and voltages will allow for a production rate over 1010 n/s. Characteristics that effect the operational stability include the suppression of the target-emitted back streaming electrons, target sputtering and cooling, and ion beam optics. Suppression of secondary electrons resulting from the deuterium beam striking the target was achieved via the implementation of an electrostatic shroud with a voltage offset of greater than -400 V relative to the target. Ion beam optics analysis resulted in the creation of a defocussing extraction nozzle, allowing for cooler target temperatures and a more compact design. To calculate the target temperatures, a finite difference method (FDM) solver incorporating the additional heat removal effects of subcooled boiling was developed. Validation of the energy balance results from the finite difference method calculations showed the iterative solver converged to heat removal results within about 3% of the expected value. Testing of the extraction nozzle at 1.43 mA and 100 kV determined that overheating of the target did not occur as the measured neutron flux of the generator was near predicted values. Many factors, including the target stopping power, deuterium atomic species, and target loading ratio, affect the flux distribution of the HFNG neutron generator. A detailed analysis to understand these factors effects is presented. Comparison of the calculated flux of the neutron generator using deuteron depth implantation data, neutron flux distribution data, and deuterium atomic species data matched the experimentally calculated flux determined from indium foil
Kalman filtering for rhodium self-powered neutron detectors
International Nuclear Information System (INIS)
Kantrowitz, M.L.
1988-01-01
Rhodium self-powered neutron detectors are utilized in many pressurized water reactors to determine the neutronic behavior within the core. In order to compensate for the inherent time delay associated with the response of these detectors, a dynamic compensation algorithm is currently used in Combustion Engineering plants to reconstruct the dynamic flux signal which is being sensed by the rhodium detectors. This paper describes a new dynamic compensation algorithm, based on Kalman filtering, which improves on the noise gain and response time characteristics of the algorithm currently used, and offers the possibility of utilizing the proven rhodium detector based fixed in-core detector system as an integral part of advanced core control and/or protection systems
International Nuclear Information System (INIS)
Sari, A.; Carrel, F.; Laine, F.; Lyoussi, A.
2013-01-01
Radiological characterization of nuclear waste drums is an important task for the nuclear industry. The amount of actinides, such as 235 U or 239 Pu, contained in a package can be determined using non-destructive active methods based on the fission process. One of these techniques, known as neutron interrogation, uses a neutron beam to induce fission reactions on the actinides. Optimization of the neutron flux is an important step towards improving this technique. Electron accelerators enable to achieve higher neutron flux intensities than the ones delivered by deuterium-tritium generators traditionally used on neutron interrogation industrial facilities. In this paper, we design a neutron interrogation cell based on an electron accelerator by MCNPX simulation. We carry out photoneutron interrogation measurements on uranium samples placed at the center of 220 liter nuclear waste drums containing different types of matrices. We quantify impact of the matrix on the prompt neutron signal, on the ratio between the prompt and delayed neutron signals, and on the interrogative neutron half-life time. We also show that characteristics of the conversion target of the electron accelerator enable to improve significantly measurement performances. (authors)
Assessing neutron generator output using neutron activation of silicon
International Nuclear Information System (INIS)
Kehayias, Pauli M.; Kehayias, Joseph J.
2007-01-01
D-T neutron generators are used for elemental composition analysis and medical applications. Often composition is determined by examining elemental ratios in which the knowledge of the neutron flux is unnecessary. However, the absolute value of the neutron flux is required when the generator is used for neutron activation analysis, to study radiation damage to materials, to monitor the operation of the generator, and to measure radiation exposure. We describe a method for absolute neutron output and flux measurements of low output D-T neutron generators using delayed activation of silicon. We irradiated a series of silicon oxide samples with 14.1 MeV neutrons and counted the resulting gamma rays of the 28 Al nucleus with an efficiency-calibrated detector. To minimize the photon self-absorption effects within the samples, we used a zero-thickness extrapolation technique by repeating the measurement with samples of different thicknesses. The neutron flux measured 26 cm away from the tritium target of a Thermo Electron A-325 D-T generator (Thermo Electron Corporation, Colorado Springs, CO) was 6.2 x 10 3 n/s/cm 2 ± 5%, which is consistent with the manufacturer's specifications
Assessing neutron generator output using neutron activation of silicon
Energy Technology Data Exchange (ETDEWEB)
Kehayias, Pauli M. [Body Composition Laboratory, Jean Mayer United States Department of Agriculture Human Nutrition Research Center on Aging, Tufts University, Boston, MA 02111 (United States); Kehayias, Joseph J. [Body Composition Laboratory, Jean Mayer United States Department of Agriculture Human Nutrition Research Center on Aging, Tufts University, Boston, MA 02111 (United States)]. E-mail: joseph.kehayias@tufts.edu
2007-08-15
D-T neutron generators are used for elemental composition analysis and medical applications. Often composition is determined by examining elemental ratios in which the knowledge of the neutron flux is unnecessary. However, the absolute value of the neutron flux is required when the generator is used for neutron activation analysis, to study radiation damage to materials, to monitor the operation of the generator, and to measure radiation exposure. We describe a method for absolute neutron output and flux measurements of low output D-T neutron generators using delayed activation of silicon. We irradiated a series of silicon oxide samples with 14.1 MeV neutrons and counted the resulting gamma rays of the {sup 28}Al nucleus with an efficiency-calibrated detector. To minimize the photon self-absorption effects within the samples, we used a zero-thickness extrapolation technique by repeating the measurement with samples of different thicknesses. The neutron flux measured 26 cm away from the tritium target of a Thermo Electron A-325 D-T generator (Thermo Electron Corporation, Colorado Springs, CO) was 6.2 x 10{sup 3} n/s/cm{sup 2} {+-} 5%, which is consistent with the manufacturer's specifications.
Neutron filters for producing monoenergetic neutron beams
International Nuclear Information System (INIS)
Harvey, J.A.; Hill, N.W.; Harvey, J.R.
1982-01-01
Neutron transmission measurements have been made on high-purity, highly-enriched samples of 58 Ni (99.9%), 60 Ni (99.7%), 64 Zn (97.9%) and 184 W (94.5%) to measure their neutron windows and to assess their potential usefulness for producing monoenergetic beams of intermediate energies from a reactor. Transmission measurements on the Los Alamos Sc filter (44.26 cm Sc and 1.0 cm Ti) have been made to determine the characteristics of the transmitted neutron beam and to measure the total cross section of Sc at the 2.0 keV minimum. When corrected for the Ti and impurities, a value of 0.35 +- 0.03 b was obtained for this minimum
International Nuclear Information System (INIS)
Labeckas, Gvidonas; Slavinskas, Stasys; Kanapkienė, Irena
2017-01-01
Highlights: • Diesel-HRD fuel blends involving ethanol (E) or biodiesel (B) were investigated in a turbocharged CRDI engine. • Improved cetane number of fuel blends ambiguously affected the ignition delay and maximum heat release rate. • Increased fuel-bound oxygen content enhanced combustion, heat release and in-cylinder pressure at 2500 rpm. • Fuel properties almost did not change premixed phase, but affected burn angle MBF 50 and the end of combustion. • Burn angles MBF 50 and MBF 90 were 1.0° and 5.7° CADs shorter when using oxygenated blend OE4 (3.6 wt%) at 2000 rpm. - Abstract: The study deals with the effects made by individual variation of cetane number, fuel-oxygen content, or widely differing properties of diesel-HRD fuel blends involving ethanol (E) or biodiesel (B) on the ignition delay, combustion phenomenon, maximum heat release rate, and the cyclic variation of a turbocharged CRDI diesel engine. The most important control factors one after another operated separately in this study to make a difference. Load characteristics were taken when running with a straight diesel and various (18) diesel-HRD fuel blends at maximum torque mode of 2000 rpm and speeds of 1500 and 2500 rpm to provide correct interpretation of the test results. Then, load (bmep) characteristics were plotted as a function of the relative air-fuel ratio (λ) and the analysis of combustion parameters was conducted for the ‘lambda’ values of λ = 1.30, 1.25 and 1.20, at the respective speeds of 1500, 2000 and 2500 rpm. Analysis of changes in the ignition delay, combustion characteristics, and the cyclic variation of parameters when using fuel blends of both origins was performed on comparative bases with the corresponding values measured with ‘base-line’ blends with CN = 51.2 or zero oxygen content and a straight diesel to reveal the potential developing trends. The enhanced cetane number of oxygenated fuels improved combustion and reduced cyclic variation when
International Nuclear Information System (INIS)
Semenov, Mikhail
2002-11-01
This report is continuation of studying of the experiments performed on BFS-62-3A critical assembly in Russia. The objective of work is definition of the cross section uncertainties on reactor neutronics parameters as applied to the hybrid core of the BN-600 reactor of Beloyarskaya NPP. Two-dimensional benchmark model of BFS-62-3A was created specially for these purposes and experimental values were reduced to it. Benchmark characteristics for this assembly are 1) criticality; 2) central fission rate ratios (spectral indices); and 3) fission rate distributions in stainless steel reflector. The effects of nuclear data libraries have been studied by comparing the results calculated using available modern data libraries - ENDF/B-V, ENDF/B-VI, ENDF/B-VI-PT, JENDL-3.2 and ABBN-93. All results were computed by Monte Carlo method with the continuous energy cross-sections. The checking of the cross sections of major isotopes on wide benchmark criticality collection was made. It was shown that ENDF/B-V data underestimate the criticality of fast reactor systems up to 2% Δk. As for the rest data, the difference between each other in criticality for BFS-62-3A is around 0.6% Δk. However, taking into account the results obtained for other fast reactor benchmarks (and steel-reflected also), it may conclude that the difference in criticality calculation results can achieve 1% Δk. This value is in a good agreement with cross section uncertainty evaluated for BN-600 hybrid core (±0.6% Δk). This work is related to the JNC-IPPE Collaboration on Experimental Investigation of Excess Weapons Grade Pu Disposition in BN-600 Reactor Using BFS-2 Facility. (author)
Neutron dosimetry at SLAC: Neutron sources and instrumentation
International Nuclear Information System (INIS)
Liu, J.C.; Jenkins, T.M.; McCall, R.C.; Ipe, N.E.
1991-10-01
This report summarizes in detail the dosimetric characteristics of the five radioisotopic type neutron sources ( 238 PuBe, 252 Cf, 238 PuB, 238 PuF 4 , and 238 PuLi) and the neutron instrumentation (moderated BF 3 detector, Anderson-Braun (AB) detector, AB remmeter, Victoreen 488 Neutron Survey Meter, Beam Shut-Off Ionization Chamber, 12 C plastic scintillator detector, moderated indium foil detector, and moderated and bare TLDs) that are commonly used for neutron dosimetry at the Stanford Linear Accelerator Center (SLAC). 36 refs,. 19 figs
Neutron dosimetry at SLAC: Neutron sources and instrumentation
Energy Technology Data Exchange (ETDEWEB)
Liu, J.C.; Jenkins, T.M.; McCall, R.C.; Ipe, N.E.
1991-10-01
This report summarizes in detail the dosimetric characteristics of the five radioisotopic type neutron sources ({sup 238}PuBe, {sup 252}Cf, {sup 238}PuB, {sup 238}PuF{sub 4}, and {sup 238}PuLi) and the neutron instrumentation (moderated BF{sub 3} detector, Anderson-Braun (AB) detector, AB remmeter, Victoreen 488 Neutron Survey Meter, Beam Shut-Off Ionization Chamber, {sup 12}C plastic scintillator detector, moderated indium foil detector, and moderated and bare TLDs) that are commonly used for neutron dosimetry at the Stanford Linear Accelerator Center (SLAC). 36 refs,. 19 figs.
Yu, Andrea C; Zambrano, Regina M; Cristian, Ingrid; Price, Sue; Bernhard, Birgitta; Zucker, Marc; Venkateswaran, Sunita; McGowan-Jordan, Jean; Armour, Christine M
2017-06-01
Isolated 7p22.3p22.2 deletions are rarely described with only two reports in the literature. Most other reported cases either involve a much larger region of the 7p arm or have an additional copy number variation. Here, we report five patients with overlapping microdeletions at 7p22.3p22.2. The patients presented with variable developmental delays, exhibiting relative weaknesses in expressive language skills and relative strengths in gross, and fine motor skills. The most consistent facial features seen in these patients included a broad nasal root, a prominent forehead a prominent glabella and arched eyebrows. Additional variable features amongst the patients included microcephaly, metopic ridging or craniosynostosis, cleft palate, cardiac defects, and mild hypotonia. Although the patients' deletions varied in size, there was a 0.47 Mb region of overlap which contained 7 OMIM genes: EIP3B, CHST12, LFNG, BRAT1, TTYH3, AMZ1, and GNA12. We propose that monosomy of this region represents a novel microdeletion syndrome. We recommend that individuals with 7p22.3p22.2 deletions should receive a developmental assessment and a thorough cardiac exam, with consideration of an echocardiogram, as part of their initial evaluation. © 2017 Wiley Periodicals, Inc.
Delayed photoneutrons of the of the Dalat Nuclear Research Reactor
International Nuclear Information System (INIS)
Ngo Quang Huy; Ha Van Thong; Vu Hai Long; Ngo Phu Khang; Nguyen Nhi Dien; Pham Van Lam; Huynh Dong Phuong; Luong Ba Vien; Le Vinh Vinh
1994-01-01
Time spectrum of delayed neutrons of the Dalat nuclear research reactor is measured and analyzed. It corresponds to a shut-down neutron fluxes of about 10 5 /10 8 n/cm 2 /sec after 100 hours continuous reactor operation at steady power level of 500 kW. Data processing of experimental time neutron spectrum gives 16 exponents, of which 10, resulting from photoneutrons due to (γ,n) reactions on beryllium used inside the reactor core, are obtained by using successive exponential stripping fitting method. For the Dalat reactor, the effective delayed photoneutron fraction relative to the total effective delayed neutron fraction is β B e eff =0.49%β eff for a beryllium weight relative to U 235 fuel of m B e/m U = 8.5. This result is acceptable in comparison to those obtained for other Be-U 235 media. (author). 5 refs., 2 figs., 4 tabs
The role of delay in the dynamics of nuclear reactors
International Nuclear Information System (INIS)
Svitra, D.; Bucys, K.
1999-01-01
The stability of nuclear reactors based on nonlinear models of reactor dynamics including the action of delayed neutrons is analysed. The point model of reactor dynamics with the system of seven nonlinear simple differential equations was changed to the system of two nonlinear differential equations including the action of delay. The method of the theory of bifurcations for nonlinear differential equations with delay is used. (author)
Neutron spectrum effects on TRU recycling in Pb-Bi cooled fast reactor core
International Nuclear Information System (INIS)
Kim, Yong Nam; Kim, Jong Kyung; Park, Won Seok
2003-01-01
This study is intended to evaluate the dependency of TRU recycling characteristics on the neutron spectrum shift in a Pb-Bi cooled core. Considering two Pb-Bi cooled cores with the soft and the hard spectrum, respectively, various characteristics of the recycled core are carefully examined and compared with each other. Assuming very simplified fuel cycle management with the homogeneous and single region fuel loading, the burnup calculations are performed until the recycled core reached to the (quasi-) equilibrium state. The mechanism of TRU recycling toward the equilibrium is analyzed in terms of burnup reactivity and the isotopic compositions of TRU fuel. In the comparative analyses, the difference in the recycling behavior between the two cores is clarified. In addition, the basic safety characteristics of the recycled core are also discussed in terms of the Doppler coefficient, the coolant loss reactivity coefficient, and the effective delayed neutron fraction
2004-01-01
Researcher Dr. Yi Li developed a technique to manipulate certain characteristics of plant growth such as anit-senescence. For example, the tobacco leaf was clipped from a transgenic plant (right), and a wildtype plant (left). During ground-based laboratory studies, both leaves were left in a darkened area for 4 months. When retrieved, the wildtype plant leaf was dried-out and the transgenic leaf remained fresh and green. A variation of this technology that involves manipulating plant hormones has been conducted in space-based studies on tomato plants through BioServe Space Technologies. The transport and distribution of auxin, an important plant hormone has shown to be influenced by microgravity, which could lead to improving the quality of fruits and vegetables grown on Earth.
Hyper-thermal neutron irradiation field for neutron capture therapy
International Nuclear Information System (INIS)
Sakurai, Yoshinori; Kobayashi, Tooru; Kanda, Keiji
1994-01-01
The utilization of hyper-thermal neutrons, which have an energy spectrum of a Maxwell distribution higher than the room temperature of 300 K, has been studied in order to improve the thermal neutron flux distribution in a living body for a deep-seated tumor in neutron capture therapy (NCT). Simulation calculations using MCNP-V3 were carried out in order to investigate the characteristics of the hyper-thermal neutron irradiation field. From the results of simulation calculations, the following were confirmed: (i) The irradiation field of the hyper-thermal neutrons is feasible by using some scattering materials with high temperature, such as Be, BeO, C, SiC and ZrH 1.7 . Especially, ZrH 1.7 is thought to be the best material because of good characteristics of up-scattering for thermal neutrons. (ii) The ZrH 1.7 of 1200 K yields the hyper-thermal neutrons of a Maxwell-like distribution at about 2000 K and the treatable depth is about 1.5 cm larger comparing with the irradiation of the thermal neutrons of 300 K. (iii) The contamination by the secondary gamma-rays from the scattering materials can be sufficiently eliminated to the tolerance level for NCT through the bismuth layer, without the larger change of the energy spectrum of hyper-thermal neutrons. ((orig.))
Modern techniques of structural neutron diffraction
International Nuclear Information System (INIS)
Aksenov, V.L.; )
1997-01-01
Modern techniques of neutron diffraction for structural investigations are analyzed. The time-of-flight method and the reverse time-of-flight method are considered briefly. Characteristics of two-crystal and time-of-flight neutron diffractometers are compared. It is pointed that in the future, the great importance will be possessed the development of high-resolution Fourier neutron diffractometers [ru
International Nuclear Information System (INIS)
Sanchez, R.; Jaegers, P.
1998-01-01
Rossi-Alpha measurements were performed on the SHEBA assembly to determine the prompt neutron decay constants. These prompt neutron decay constants represent an eigenvalue characteristic of this particular assembly, which can be used to infer the amount of fissile material in the assembly. In addition, subcritical measurements using Rossi-Alpha and the source-jerk techniques were also performed on the SHEBA assembly. These measurements were compared against TWODANT calculations and agreed quite well. The subcritical measurements were also used to obtain a unique signature that represented the amount of material associated with the degree of subcriticality of the SHEBA assembly. Finally, the Feynman variance-to-mean technique in conjunction with TWODANT, were used to determine the effective delayed neutron fraction for the SHEBA assembly
Pulsed neutron uranium borehole logging with prompt fission neutrons
International Nuclear Information System (INIS)
Bivens, H.M.; Smith, G.W.; Jensen, D.H.
1976-01-01
The gross count natural gamma log normally used for uranium borehole logging is seriously affected by disequilibrium. Methods for the direct measurement of uranium, such as neutron logging, which are not affected by disequilibrium have been the object of considerable effort in recent years. This paper describes a logging system for uranium which uses a small accelerator to generate pulses of 14 MeV neutrons to detect and assay uranium by the measurement of prompt fission neutrons in the epithermal energy range. After an initial feasibility study, a prototype logging probe was built for field evaluation which began in January 1976. Physical and operational characteristics of the prototype probe, the neutron tube-transformer assembly, and the neutron tube are described. In logging operations, only the epithermal prompt fission neutrons detected between 250 microseconds to 2500 microseconds following the excitation neutron pulse are counted. Comparison of corrected neutron logs with the conventional gross count natural gamma logs and the chemical assays of cores from boreholes are shown. The results obtained with this neutron probe clearly demonstrate its advantages over the gross count natural gamma log, although at this time the accuracy of the neutron log assay is not satisfactory under some conditions. The necessary correction factors for various borehole and formation parameters are being determined and, when applied, should improve the assay accuracy
Fundamental of neutron radiography and the present of neutron radiography in Japan
International Nuclear Information System (INIS)
Sekita, Junichiro
1988-01-01
Neutron radiography refers to the application of transmitted neutrons to analysis. In general, thermal neutron is used for neutron radiography. Thermal neutron is easily absorbed by light atoms, including hydrogen, boron and lithium, while it is not easily absorbed by such heavy atoms as tungsten, lead and uranium, permitting detection of impurities in heavy metals. Other neutrons than thermal neutron can also be applied. Cold neutron is produced from fast neutron using a moderator to reduce its energy down to below that of thermal neutron. Cold neutron is usefull for analysis of thick material. Epithermal neutron can induce resonance characteristic of each substance. With a relatively small reaction area, fast neutron permits observation of thick samples. Being electrically neutral, neutrons are difficult to detect by direct means. Thus a substance that releases charged particles is put in the path of neutrons for indirect measurement. X-ray film combined with converter screen for conversion of neutrons to charge particles is placed behind the sample. Photographing is carried out by a procedure similar to X-ray photography. Major institues and laboratories in Japan provided with neutron radiography facilities are listed. (Nogami, K.)
The Los Alamos Neutron Science Center Spallation Neutron Sources
International Nuclear Information System (INIS)
Nowicki, Suzanne F.; Wender, Stephen A.; Mocko, Michael
2017-01-01
The Los Alamos Neutron Science Center (LANSCE) provides the scientific community with intense sources of neutrons, which can be used to perform experiments supporting civilian and national security research. These measurements include nuclear physics experiments for the defense program, basic science, and the radiation effect programs. This paper focuses on the radiation effects program, which involves mostly accelerated testing of semiconductor parts. When cosmic rays strike the earth's atmosphere, they cause nuclear reactions with elements in the air and produce a wide range of energetic particles. Because neutrons are uncharged, they can reach aircraft altitudes and sea level. These neutrons are thought to be the most important threat to semiconductor devices and integrated circuits. The best way to determine the failure rate due to these neutrons is to measure the failure rate in a neutron source that has the same spectrum as those produced by cosmic rays. Los Alamos has a high-energy and a low-energy neutron source for semiconductor testing. Both are driven by the 800-MeV proton beam from the LANSCE accelerator. The high-energy neutron source at the Weapons Neutron Research (WNR) facility uses a bare target that is designed to produce fast neutrons with energies from 100 keV to almost 800 MeV. The measured neutron energy distribution from WNR is very similar to that of the cosmic-ray-induced neutrons in the atmosphere. However, the flux provided at the WNR facility is typically 5×107 times more intense than the flux of the cosmic-ray-induced neutrons. This intense neutron flux allows testing at greatly accelerated rates. An irradiation test of less than an hour is equivalent to many years of neutron exposure due to cosmic-ray neutrons. The low-energy neutron source is located at the Lujan Neutron Scattering Center. It is based on a moderated source that provides useful neutrons from subthermal energies to ~100 keV. The characteristics of these sources
The Los Alamos Neutron Science Center Spallation Neutron Sources
Nowicki, Suzanne F.; Wender, Stephen A.; Mocko, Michael
The Los Alamos Neutron Science Center (LANSCE) provides the scientific community with intense sources of neutrons, which can be used to perform experiments supporting civilian and national security research. These measurements include nuclear physics experiments for the defense program, basic science, and the radiation effect programs. This paper focuses on the radiation effects program, which involves mostly accelerated testing of semiconductor parts. When cosmic rays strike the earth's atmosphere, they cause nuclear reactions with elements in the air and produce a wide range of energetic particles. Because neutrons are uncharged, they can reach aircraft altitudes and sea level. These neutrons are thought to be the most important threat to semiconductor devices and integrated circuits. The best way to determine the failure rate due to these neutrons is to measure the failure rate in a neutron source that has the same spectrum as those produced by cosmic rays. Los Alamos has a high-energy and a low-energy neutron source for semiconductor testing. Both are driven by the 800-MeV proton beam from the LANSCE accelerator. The high-energy neutron source at the Weapons Neutron Research (WNR) facility uses a bare target that is designed to produce fast neutrons with energies from 100 keV to almost 800 MeV. The measured neutron energy distribution from WNR is very similar to that of the cosmic-ray-induced neutrons in the atmosphere. However, the flux provided at the WNR facility is typically 5×107 times more intense than the flux of the cosmic-ray-induced neutrons. This intense neutron flux allows testing at greatly accelerated rates. An irradiation test of less than an hour is equivalent to many years of neutron exposure due to cosmic-ray neutrons. The low-energy neutron source is located at the Lujan Neutron Scattering Center. It is based on a moderated source that provides useful neutrons from subthermal energies to ∼100 keV. The characteristics of these sources, and
Neutron activation probe for measuring the presence of uranium in ore bodies
International Nuclear Information System (INIS)
Goldstein, N.P.; Smith, R.C.
1979-01-01
A neutron activation proble comprises a pulsed neutron source in series with a plurality of delayed neutron detectors for measuring radioactivity in a well borehole together with a NaI (Tl) counter for measuring the high energy 2.62 MeV gamma line from thorium. The neutron source emits neutrons which produce fission in uranium and thorium in the ore body and the delayed neutron detectors measure the delayed neutrons produced from such fission while the NaI (Tl) counter measures the 2.62 MeV gamma line from the undisturbed thorium in the ore body. The signal from the NaI (Tl) counter is processed and subtracted from the signal from the delayed neutron detectors with the result being indicative of the amount of uranium present in the ore body
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