Inverse kinetics for subcritical systems with external neutron source

International Nuclear Information System (INIS)

Carvalho Gonçalves, Wemerson de; Martinez, Aquilino Senra; Carvalho da Silva, Fernando

2017-01-01

Highlights: • It was developed formalism for reactivity calculation. • The importance function is related to the system subcriticality. • The importance function is also related with the value of the external source. • The equations were analyzed for seven different levels of sub criticality. • The results are physically consistent with others formalism discussed in the paper. - Abstract: Nuclear reactor reactivity is one of the most important properties since it is directly related to the reactor control during the power operation. This reactivity is influenced by the neutron behavior in the reactor core. The time-dependent neutrons behavior in response to any change in material composition is important for the reactor operation safety. Transient changes may occur during the reactor startup or shutdown and due to accidental disturbances of the reactor operation. Therefore, it is very important to predict the time-dependent neutron behavior population induced by changes in neutron multiplication. Reactivity determination in subcritical systems driven by an external neutron source can be obtained through the solution of the inverse kinetics equation for subcritical nuclear reactors. The main purpose of this paper is to find the solution of the inverse kinetics equation the main purpose of this paper is to device the inverse kinetics equations for subcritical systems based in a previous paper published by the authors (Gonçalves et al., 2015) and by (Gandini and Salvatores, 2002; Dulla et al., 2006). The solutions of those equations were also obtained. Formulations presented in this paper were tested for seven different values of k eff with external neutrons source constant in time and for a powers ratio varying exponentially over time.

Reactor kinetics revisited: a coefficient based model (CBM)

International Nuclear Information System (INIS)

Ratemi, W.M.

2011-01-01

In this paper, a nuclear reactor kinetics model based on Guelph expansion coefficients calculation ( Coefficients Based Model, CBM), for n groups of delayed neutrons is developed. The accompanying characteristic equation is a polynomial form of the Inhour equation with the same coefficients of the CBM- kinetics model. Those coefficients depend on Universal abc- values which are dependent on the type of the fuel fueling a nuclear reactor. Furthermore, such coefficients are linearly dependent on the inserted reactivity. In this paper, the Universal abc- values have been presented symbolically, for the first time, as well as with their numerical values for U-235 fueled reactors for one, two, three, and six groups of delayed neutrons. Simulation studies for constant and variable reactivity insertions are made for the CBM kinetics model, and a comparison of results, with numerical solutions of classical kinetics models for one, two, three, and six groups of delayed neutrons are presented. The results show good agreements, especially for single step insertion of reactivity, with the advantage of the CBM- solution of not encountering the stiffness problem accompanying the numerical solutions of the classical kinetics model. (author)

Coupled 3D neutron kinetics and thermalhydraulic characteristics of the Canadian supercritical water reactor

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

Implementation into a CFD code of neutron kinetics and fuel pin models for nuclear reactor transient analyses

International Nuclear Information System (INIS)

Chen Zhao; Chen, Xue-Nong; Rineiski, Andrei; Zhao Pengcheng; Chen Hongli

2014-01-01

Safety analysis is an important tool for justifying the safety of nuclear reactors. The traditional method for nuclear reactor safety analysis is performed by means of system codes, which use one-dimensional lumped-parameter method to model real reactor systems. However, there are many multi-dimensional thermal-hydraulic phenomena cannot be predicated using traditional one-dimensional system codes. This problem is extremely important for pool-type nuclear systems. Computational fluid dynamics (CFD) codes are powerful numerical simulation tools to solve multi-dimensional thermal-hydraulics problems, which are widely used in industrial applications for single phase flows. In order to use general CFD codes to solve nuclear reactor transient problems, some additional models beyond general ones are required. Neutron kinetics model for power calculation and fuel pin model for fuel pin temperature calculation are two important models of these additional models. The motivation of this work is to develop an advance numerical simulation method for nuclear reactor safety analysis by implementing neutron kinetics model and fuel pin model into general CFD codes. In this paper, the Point Kinetics Model (PKM) and Fuel Pin Model (FPM) are implemented into a general CFD code FLUENT. The improved FLUENT was called as FLUENT/PK. The mathematical models and implementary method of FLUENT/PK are descripted and two demonstration application cases, e.g. the unprotected transient overpower (UTOP) accident of a Liquid Metal cooled Fast Reactor (LMFR) and the unprotected beam overpower (UBOP) accident of an Accelerator Driven System (ADS), are presented. (author)

Nuclear reactor kinetics and plant control

CERN Document Server

Oka, Yoshiaki

2013-01-01

Understanding time-dependent behaviors of nuclear reactors and the methods of their control is essential to the operation and safety of nuclear power plants. This book provides graduate students, researchers, and engineers in nuclear engineering comprehensive information on both the fundamental theory of nuclear reactor kinetics and control and the state-of-the-art practice in actual plants, as well as the idea of how to bridge the two. The first part focuses on understanding fundamental nuclear kinetics. It introduces delayed neutrons, fission chain reactions, point kinetics theory, reactivit

Disclosure of the oscillations in kinetics of the reactor pressure vessel steel damage at fast neutron intensity decreasing

Science.gov (United States)

Krasikov, E.; Nikolaenko, V.

2017-01-01

Fast neutron intensity influence on reactor materials radiation damage is a critically important question in the problem of the correct use of the accelerated irradiation tests data for substantiation of the materials workability in real irradiation conditions that is low neutron intensity. Investigations of the fast neutron intensity (flux) influence on radiation damage and experimental data scattering reveal the existence of non-monotonous sections in kinetics of the reactor pressure vessels (RPV) steel damage. Discovery of the oscillations as indicator of the self-organization processes presence give reasons for new ways searching on reactor pressure vessel (RPV) steel radiation stability increasing and attempt of the self-restoring metal elaboration. Revealing of the wavelike process in the form of non monotonous parts of the kinetics of radiation embrittlement testifies that periodic transformation of the structure take place. This fact actualizes the problem of more precise definition of the RPV materials radiation embrittlement mechanisms and gives reasons for search of the ways to manage the radiation stability (nanostructuring and so on to stimulate the radiation defects annihilation), development of the means for creating of more stableness self recovering smart materials.

Numerical Solution of Fractional Neutron Point Kinetics Model in Nuclear Reactor

Directory of Open Access Journals (Sweden)

Nowak Tomasz Karol

2014-06-01

Full Text Available This paper presents results concerning solutions of the fractional neutron point kinetics model for a nuclear reactor. Proposed model consists of a bilinear system of fractional and ordinary differential equations. Three methods to solve the model are presented and compared. The first one entails application of discrete Grünwald-Letnikov definition of the fractional derivative in the model. Second involves building an analog scheme in the FOMCON Toolbox in MATLAB environment. Third is the method proposed by Edwards. The impact of selected parameters on the model’s response was examined. The results for typical input were discussed and compared.

Physics and kinetics of TRIGA reactor

International Nuclear Information System (INIS)

Boeck, H.; Villa, M.

2007-01-01

This training module is written as an introduction to reactor physics for reactor operators. It assumes the reader has a basic, fundamental knowledge of physics, materials and mathematics. The objective is to provide enough reactor theory knowledge to safely operate a typical research reactor. At this level, it does not necessarily provide enough information to evaluate the safety aspects of experiment or non-standard operation reviews. The material provides a survey of basic reactor physics and kinetics of TRIGA type reactors. Subjects such as the multiplication factor, reactivity, temperature coefficients, poisoning, delayed neutrons and criticality are discussed in such a manner that even someone not familiar with reactor physics and kinetics can easily follow. A minimum of equations are used and several tables and graphs illustrate the text. (author)

One-dimensional reactor kinetics model for RETRAN

International Nuclear Information System (INIS)

Gose, G.C.; Peterson, C.E.; Ellis, N.L.; McClure, J.A.

1981-01-01

This paper describes a one-dimensional spatial neutron kinetics model that was developed for the RETRAN code. The RETRAN -01 code has a point kinetics model to describe the reactor core behavior during thermal-hydraulic transients. A one-dimensional neutronics model has been developed for RETRAN-02. The ability to account for flux shape changes will permit an improved representation of the thermal and hydraulic feedback effects for many operational transients. 19 refs

Reactor thermal behaviors under kinetics parameters variations in fast reactivity insertion

Energy Technology Data Exchange (ETDEWEB)

Abou-El-Maaty, Talal [Reactors Department, Atomic Energy Authority, Cairo 13759 (Egypt)], E-mail: talal22969@yahoo.com; Abdelhady, Amr [Reactors Department, Atomic Energy Authority, Cairo 13759 (Egypt)

2009-03-15

The influences of variations in some of the kinetics parameters affecting the reactivity insertion are considered in this study, it has been accomplished in order to acquire knowledge about the role that kinetic parameters play in prompt critical transients from the safety point of view. The kinetics parameters variations are limited to the effective delayed neutron fraction ({beta}{sub eff}) and the prompt neutron generation time ({lambda}). The reactor thermal behaviors under the variations in effective delayed neutron fraction and prompt neutron generation time included, the reactor power, maximum fuel temperature, maximum clad temperature, maximum coolant temperature and the mass flux variations at the hot channel. The analysis is done for a typical swimming pool, plate type research reactor with low enriched uranium. The scram system is disabled during the accidents simulations. Calculations were done using PARET code. As a result of simulations, it is concluded that, the reactor (ETRR2) thermal behavior is considerably more sensitive to the variation in the effective delayed neutron fraction than to the variation in prompt neutron generation time and the fast reactivity insertion in both cases causes a flow expansion and contraction at the hot channel exit. The amplitude of the oscillated flow is a qualitatively increases with the decrease in both {beta}{sub eff} and {lambda}.

Numerical simulations of subcritical reactor kinetics in thermal hydraulic transient phases

Energy Technology Data Exchange (ETDEWEB)

Yoo, J; Park, W S [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1999-12-31

A subcritical reactor driven by a linear proton accelerator has been considered as a nuclear waste incinerator at Korea Atomic Energy Research Institute (KAERI). Since the multiplication factor of a subcritical reactor is less than unity, to compensate exponentially decreasing fission neutrons, external neutrons form spallation reactions are essentially required for operating the reactor in its steady state. Furthermore, the profile of accelerator beam currents is very important in controlling a subcritical reactor, because the reactor power varies in accordance to the profile of external neutrons. We have developed a code system to find numerical solutions of reactor kinetics equations, which are the simplest dynamic model for controlling reactors. In a due course of our previous numerical study of point kinetics equations for critical reactors, however, we learned that the same code system can be used in studying dynamic behavior of the subcritical reactor. Our major motivation of this paper is to investigate responses of subcritical reactors for small changes in thermal hydraulic parameters. Building a thermal hydraulic model for the subcritical reactor dynamics, we performed numerical simulations for dynamic responses of the reactor based on point kinetics equations with a source term. Linearizing a set of coupled differential equations for reactor responses, we focus our research interest on dynamic responses of the reactor to variations of the thermal hydraulic parameters in transient phases. 5 refs., 8 figs. (Author)

Numerical simulations of subcritical reactor kinetics in thermal hydraulic transient phases

Energy Technology Data Exchange (ETDEWEB)

Yoo, J.; Park, W. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1998-12-31

A subcritical reactor driven by a linear proton accelerator has been considered as a nuclear waste incinerator at Korea Atomic Energy Research Institute (KAERI). Since the multiplication factor of a subcritical reactor is less than unity, to compensate exponentially decreasing fission neutrons, external neutrons form spallation reactions are essentially required for operating the reactor in its steady state. Furthermore, the profile of accelerator beam currents is very important in controlling a subcritical reactor, because the reactor power varies in accordance to the profile of external neutrons. We have developed a code system to find numerical solutions of reactor kinetics equations, which are the simplest dynamic model for controlling reactors. In a due course of our previous numerical study of point kinetics equations for critical reactors, however, we learned that the same code system can be used in studying dynamic behavior of the subcritical reactor. Our major motivation of this paper is to investigate responses of subcritical reactors for small changes in thermal hydraulic parameters. Building a thermal hydraulic model for the subcritical reactor dynamics, we performed numerical simulations for dynamic responses of the reactor based on point kinetics equations with a source term. Linearizing a set of coupled differential equations for reactor responses, we focus our research interest on dynamic responses of the reactor to variations of the thermal hydraulic parameters in transient phases. 5 refs., 8 figs. (Author)

Neutronics - thermal-hydraulics coupling: application to the helium-cooled fast reactor

International Nuclear Information System (INIS)

Vaiana, F.

2009-11-01

This thesis focuses on the study of interactions between neutron-kinetics and thermal-hydraulics. Neutron-kinetics allow to calculate the power in a nuclear reactor and the temperature evolution of materials where this power is deposited is known thanks to thermal-hydraulics. Moreover, when the temperatures evolve, the densities and cross sections change. These two disciplines are thus coupled. The first part of this work corresponds to the study and development of a method which allows to simulate transients in nuclear reactors and especially with a Monte-Carlo code for neutron-kinetics. An algorithm for the resolution of the neutron transport equation has been established and validated with a benchmark. In thermal-hydraulics, a porous media approach, based on another thesis, is considered. This gives the opportunity to solve the equations on the whole core without unconscionable computation time. Finally, a theoretical study has been performed on the statistical uncertainties which result from the use of a Monte-Carlo code and which spread from the reactivity to the power and from the power to the temperatures. The second part deals with the study of a misplaced control rod withdrawing in a GFR (helium-cooled fast reactor), a fourth generation reactor. Some models allowing to calculate neutron-kinetics and thermal-hydraulics in the core (which contains assemblies built up with fuel plates) were defined. In thermal-hydraulics, a model for the core based on the porous media approach and a fuel plate homogenization model have been set up. A similar homogenization model has been studied for neutron-kinetics. Finally, the control rod withdrawing transient where we can observe the power raising and the stabilisation by thermal feedback has been performed with the Monte-Carlo code Tripoli for neutron-kinetics and the code Trio-U for thermal-hydraulics. (author)

Thermal hydraulic and neutron kinetic simulation of the Angra 2 reactor using a RELAP5/PARCS coupled model

Energy Technology Data Exchange (ETDEWEB)

Reis, Patricia A.L.; Costa, Antonella L.; Hamers, Adolfo R.; Pereira, Claubia; Rodrigues, Thiago D.A.; Mantecon, Javier G.; Veloso, Maria A.F., E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: adolforomerohamers@hotmail.com, E-mail: claubia@nuclear.ufmg.br, E-mail: thiagodanielbh@gmail.com, E-mail: mantecon1987@gmail.com, E-mail: dora@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq), Belo Horizonte (Brazil); Miro, Rafael; Verdu, Gumersindo, E-mail: rmiro@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Universidad Politecnica de Valencia (Spain). Departamento de Ingenieria Quimica y Nuclear

2015-07-01

The computational advances observed in the last two decades have been provided direct impact on the researches related to nuclear simulations, which use several types of computer codes, including coupled between them, allowing representing with very accuracy the behavior of nuclear plants. Studies of complex scenarios in nuclear reactors have been improved by the use of thermal-hydraulic (TH) and neutron kinetics (NK) coupled codes. This technique consists in incorporating three-dimensional (3D) neutron modeling of the reactor core into codes, mainly to simulate transients that involve asymmetric core spatial power distributions and strong feedback effects between neutronics and reactor thermal-hydraulics. Therefore, this work presents preliminary results of TH RELAP5 and the NK PARCS calculations applied to model of the Angra 2 reactor. The WIMSD-5B code has been used to generate the macroscopic cross sections used in the NK code. The results obtained are satisfactory and represent important part of the development of this methodology. The next step is to couple the codes. (author)

An Explicit Finite Difference scheme for numerical solution of fractional neutron point kinetic equation

International Nuclear Information System (INIS)

Saha Ray, S.; Patra, A.

2012-01-01

Highlights: ► In this paper fractional neutron point kinetic equation has been analyzed. ► The numerical solution for fractional neutron point kinetic equation is obtained. ► Explicit Finite Difference Method has been applied. ► Supercritical reactivity, critical reactivity and subcritical reactivity analyzed. ► Comparison between fractional and classical neutron density is presented. - Abstract: In the present article, a numerical procedure to efficiently calculate the solution for fractional point kinetics equation in nuclear reactor dynamics is investigated. The Explicit Finite Difference Method is applied to solve the fractional neutron point kinetic equation with the Grunwald–Letnikov (GL) definition (). Fractional Neutron Point Kinetic Model has been analyzed for the dynamic behavior of the neutron motion in which the relaxation time associated with a variation in the neutron flux involves a fractional order acting as exponent of the relaxation time, to obtain the best operation of a nuclear reactor dynamics. Results for neutron dynamic behavior for subcritical reactivity, supercritical reactivity and critical reactivity and also for different values of fractional order have been presented and compared with the classical neutron point kinetic (NPK) equation as well as the results obtained by the learned researchers .

Neutron and thermal dynamics of a gaseous core fission reactor

International Nuclear Information System (INIS)

van Dam, H.; Kuijper, J.C.; Stekelenburg, A.J.C.; Hoogenboom, J.E.; Boersma-Klein, W.; Kistemaker, J.

1989-01-01

In this paper neutron kinetics and thermal dynamics of a Gaseous Core Fission Reactor with magnetical pumping are shown to have many unconventional aspects. Attention is focused on the properties of the fuel gas, the non-linear neutron kinetics and the energy balance in thermodynamical cycles

Performance of neutron kinetics models for ADS transient analyses

International Nuclear Information System (INIS)

Rineiski, A.; Maschek, W.; Rimpault, G.

2002-01-01

Within the framework of the SIMMER code development, neutron kinetics models for simulating transients and hypothetical accidents in advanced reactor systems, in particular in Accelerator Driven Systems (ADSs), have been developed at FZK/IKET in cooperation with CE Cadarache. SIMMER is a fluid-dynamics/thermal-hydraulics code, coupled with a structure model and a space-, time- and energy-dependent neutronics module for analyzing transients and accidents. The advanced kinetics models have also been implemented into KIN3D, a module of the VARIANT/TGV code (stand-alone neutron kinetics) for broadening application and for testing and benchmarking. In the paper, a short review of the SIMMER and KIN3D neutron kinetics models is given. Some typical transients related to ADS perturbations are analyzed. The general models of SIMMER and KIN3D are compared with more simple techniques developed in the context of this work to get a better understanding of the specifics of transients in subcritical systems and to estimate the performance of different kinetics options. These comparisons may also help in elaborating new kinetics models and extending existing computation tools for ADS transient analyses. The traditional point-kinetics model may give rather inaccurate transient reaction rate distributions in an ADS even if the material configuration does not change significantly. This inaccuracy is not related to the problem of choosing a 'right' weighting function: the point-kinetics model with any weighting function cannot take into account pronounced flux shape variations related to possible significant changes in the criticality level or to fast beam trips. To improve the accuracy of the point-kinetics option for slow transients, we have introduced a correction factor technique. The related analyses give a better understanding of 'long-timescale' kinetics phenomena in the subcritical domain and help to evaluate the performance of the quasi-static scheme in a particular case. One

Numerical solution of the point reactor kinetics equations with fuel burn-up and temperature feedback

International Nuclear Information System (INIS)

Tashakor, S.; Jahanfarnia, G.; Hashemi-Tilehnoee, M.

2010-01-01

Point reactor kinetics equations are solved numerically using one group of delayed neutrons and with fuel burn-up and temperature feedback included. To calculate the fraction of one-group delayed neutrons, a group of differential equations are solved by an implicit time method. Using point reactor kinetics equations, changes in mean neutrons density, temperature, and reactivity are calculated in different times during the reactor operation. The variation of reactivity, temperature, and maximum power with time are compared with the predictions by other methods.

Blankets for fusion reactors : materials and neutronics

International Nuclear Information System (INIS)

Carvalho, S.H. de.

1980-03-01

The studies about Fusion Reactors have lead to several problems for which there is no general agreement about the best solution. Nevertheless, several points seem to be well defined, at least for the first generation of reactors. The fuel, for example, should be a mixture of deuterium and tritium. Therefore, the reactor should be able to generate the tritium to be burned and also to transform kinetic energy of the fusion neutrons into heat in a process similar to the fission reactors. The best materials for the composition of the blanket were first selected and then the neutronics for the proposed system was developed. The neutron flux in the blanket was calculated using the discrete ordinates transport code, ANISN. All the nuclides cross sections came from the DLC-28/CTR library, that processed the ENDF/B data, using the SUPERTOG Program. (Author) [pt

Subcriticality calculation in nuclear reactors with external neutron sources

Energy Technology Data Exchange (ETDEWEB)

Silva, Adilson Costa da; Martinez, Aquilino Senra; Silva, Fernando Carvalho da [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia (COPPE). Programa de Engenharia Nuclear]. E-mails: asilva@con.ufrj.br; aquilino@lmp.ufrj.br; fernando@con.ufrj.br

2007-07-01

The main objective of this paper consists on the development of a methodology to monitor subcriticality. We used the inverse point kinetic equation with 6 precursor groups and external neutron sources for the calculation of reactivity. The input data for the inverse point kinetic equation was adjusted, in order to use the neutron counting rates obtained from the subcritical multiplication (1/M) in a nuclear reactor. In this paper, we assumed that the external neutron sources strength is constant and we define it in terms of a known initial condition. The results obtained from inverse point kinetic equation with external neutron sources were compared with the results obtained with a benchmark calculation, and showed good accuracy (author)

Subcriticality calculation in nuclear reactors with external neutron sources

International Nuclear Information System (INIS)

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

2007-01-01

The main objective of this paper consists on the development of a methodology to monitor subcriticality. We used the inverse point kinetic equation with 6 precursor groups and external neutron sources for the calculation of reactivity. The input data for the inverse point kinetic equation was adjusted, in order to use the neutron counting rates obtained from the subcritical multiplication (1/M) in a nuclear reactor. In this paper, we assumed that the external neutron sources strength is constant and we define it in terms of a known initial condition. The results obtained from inverse point kinetic equation with external neutron sources were compared with the results obtained with a benchmark calculation, and showed good accuracy (author)

A benchmark for coupled thermohydraulics system/three-dimensional neutron kinetics core models

International Nuclear Information System (INIS)

Kliem, S.

1999-01-01

During the last years 3D neutron kinetics core models have been coupled to advanced thermohydraulics system codes. These coupled codes can be used for the analysis of the whole reactor system. Although the stand-alone versions of the 3D neutron kinetics core models and of the thermohydraulics system codes generally have a good verification and validation basis, there is a need for additional validation work. This especially concerns the interaction between the reactor core and the other components of a nuclear power plant (NPP). In the framework of the international 'Atomic Energy Research' (AER) association on VVER Reactor Physics and Reactor Safety, a benchmark for these code systems was defined. (orig.)

Description of the CAREM Reactor Neutronic Calculation Codes

International Nuclear Information System (INIS)

Villarino, Eduardo; Hergenreder, Daniel

2000-01-01

In this work is described the neutronic calculation line used to design the CAREM reactor.A description of the codes used and the interfaces between the different programs are presented.Both, the normal calculation line and the alternative or verification calculation line are included.The calculation line used to obtain the kinetics parameters (effective delayed-neutron fraction and prompt-neutron lifetime) is also included

Numerical simulation of stochastic point kinetic equation in the dynamical system of nuclear reactor

International Nuclear Information System (INIS)

Saha Ray, S.

2012-01-01

Highlights: ► In this paper stochastic neutron point kinetic equations have been analyzed. ► Euler–Maruyama method and Strong Taylor 1.5 order method have been discussed. ► These methods are applied for the solution of stochastic point kinetic equations. ► Comparison between the results of these methods and others are presented in tables. ► Graphs for neutron and precursor sample paths are also presented. -- Abstract: In the present paper, the numerical approximation methods, applied to efficiently calculate the solution for stochastic point kinetic equations () in nuclear reactor dynamics, are investigated. A system of Itô stochastic differential equations has been analyzed to model the neutron density and the delayed neutron precursors in a point nuclear reactor. The resulting system of Itô stochastic differential equations are solved over each time-step size. The methods are verified by considering different initial conditions, experimental data and over constant reactivities. The computational results indicate that the methods are simple and suitable for solving stochastic point kinetic equations. In this article, a numerical investigation is made in order to observe the random oscillations in neutron and precursor population dynamics in subcritical and critical reactors.

Nuclear reactor neutron shielding

Science.gov (United States)

Speaker, Daniel P; Neeley, Gary W; Inman, James B

2017-09-12

A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactor cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.

Experimental determination of neutron lifetimes through macroscopic neutron noise in the IPEN/MB-01 reactor

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.

Measurements for kinetic parameters estimation in the RA-0 research reactor

International Nuclear Information System (INIS)

Gomez, A; Bellino, P A

2012-01-01

In the present work, measurements based on the neutron noise technique and the inverse kinetic method were performed to estimate the different kinetic parameters of the reactor in its critical state. By means of the neutron noise technique, we obtained the current calibration factor of the ionization chamber M6 belonging to the power range channels of the reactor instrumentation. The maximum current allowed compatible with the maximum power authorized by the operation license was also obtained. Using the neutron noise technique, the reduced mean reproduction time (Λ*) was estimated. This parameter plays a fundamental role in the deterministic analysis of criticality accidents. Comparison with previous values justified performing new measurements to study systematic trends in the value of Λ*. Using the inverse kinetics method, the reactivity worth of the control rods was estimated, confirming the existence of spatial effects and trends previously observed (author)

One-dimensional reactor kinetics model for RETRAN

International Nuclear Information System (INIS)

Gose, G.C.; Peterson, C.E.; Ellis, N.L.; McClure, J.A.

1981-01-01

Previous versions of RETRAN have had only a point kinetics model to describe the reactor core behavior during thermal-hydraulic transients. The principal assumption in deriving the point kinetics model is that the neutron flux may be separated into a time-dependent amplitude funtion and a time-independent shape function. Certain types of transients cannot be correctly analyzed under this assumption, since proper definitions for core average quantities such as reactivity or lifetime include the inner product of the adjoint flux with the perturbed flux. A one-dimensional neutronics model has been included in a preliminary version of RETRAN-02. The ability to account for flux shape changes will permit an improved representation of the thermal and hydraulic feedback effects. This paper describes the neutronics model and discusses some of the analyses

Space-time neutronic analysis of postulated LOCA's in CANDU reactors

International Nuclear Information System (INIS)

Luxat, J.C.; Frescura, G.M.

1978-01-01

Space-time neutronic behaviour of CANDU reactors is of importance in the analysis and design of reactor safety systems. A methodology has been developed for simulating CANDU space-time neutronics with application to the analysis of postulated LOCA'S. The approach involves the efficient use of a set of computer codes which provide a capability to perform simulations ranging from detailed, accurate 3-dimensional space-time to low-cost survey calculations using point kinetics with some ''effective'' spatial content. A new, space-time kinetics code based upon a modal expansion approach is described. This code provides an inexpensive and relatively accurate scoping tool for detailed 3-dimensional space-time simulations. (author)

Sandia reactor kinetics codes: SAK and PK1D

International Nuclear Information System (INIS)

Pickard, P.S.; Odom, J.P.

1978-01-01

The Sandia Kinetics code (SAK) is a one-dimensional coupled thermal-neutronics transient analysis code for use in simulation of reactor transients. The time-dependent cross section routines allow arbitrary time-dependent changes in material properties. The one-dimensional heat transfer routines are for cylindrical geometry and allow arbitrary mesh structure, temperature-dependent thermal properties, radiation treatment, and coolant flow and heat-transfer properties at the surface of a fuel element. The Point Kinetics 1 Dimensional Heat Transfer Code (PK1D) solves the point kinetics equations and has essentially the same heat-transfer treatment as SAK. PK1D can address extended reactor transients with minimal computer execution time

Study on neutron diffusion and time dependence heat ina fluidized bed nuclear reactor

International Nuclear Information System (INIS)

Vilhena, M.T. de.

1988-01-01

The purpose of this work is to model the neutron diffusion and heat transfer for a Fluidized Bed Nuclear Reactor and its solution by Laplace Transform Technique with numerical inversion using Fourier Series. Also Gaussian quadrature and residues techniques were applied for numerical inversion. The neutron transport, diffusion, and point Kinetic equation for this nuclear reactor concept are developed. A matricial and Taylor Series methods are proposed for the solution of the point Kinetic equation which is a time scale problem of Stiff type

New applications of neutron noise theory in power reactor physics

Energy Technology Data Exchange (ETDEWEB)

Arzhanov, Vasiliy

2000-04-01

The present thesis deals with neutron noise theory as applied to three comparatively different topics (or problems) in power reactor physics. Namely they are: theoretical investigation of the possibility to use a newly proposed current-flux (C/F) detector in Pressurized Water Reactors (PWRs) for the localisation of anomalies; both definition and studies on the point kinetic and adiabatic approximations for the relatively recently proposed Accelerator Driven Systems (ADS); development of the general theory of linear reactor kinetics and neutron noise in systems with varying size. One important practical problem is to detect and localise a vibrating control rod pin. The significance comes from the operational experience which indicates that individual pins can execute excessive mechanical vibrations that may lead to damage. Such mechanical vibrations induce neutron noise that can be detected. While the detection is relatively easy, the localisation of a vibrating control rod is much more complicated because only one measuring position is available and one needs to have at least three measured quantities. Therefore it has currently been proposed that the fluctuations of the neutron current vector, called the current noise, can be used in addition to the scalar noise in reactor diagnostic problems. The thesis investigates the possibility of the localization of a vibrating control rod pin in a PWR control assembly by using the scalar neutron noise and the 2-D radial current noise as measured at one central point in the control assembly. An explicit localisation technique is elaborated in which the searched position is determined as the absolute minimum of a minimisation function. The technique is investigated in numerical simulations. The results of the simulation tests show the potential applicability of the method. By design accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and

New applications of neutron noise theory in power reactor physics

International Nuclear Information System (INIS)

Arzhanov, Vasiliy

2000-04-01

The present thesis deals with neutron noise theory as applied to three comparatively different topics (or problems) in power reactor physics. Namely they are: theoretical investigation of the possibility to use a newly proposed current-flux (C/F) detector in Pressurized Water Reactors (PWRs) for the localisation of anomalies; both definition and studies on the point kinetic and adiabatic approximations for the relatively recently proposed Accelerator Driven Systems (ADS); development of the general theory of linear reactor kinetics and neutron noise in systems with varying size. One important practical problem is to detect and localise a vibrating control rod pin. The significance comes from the operational experience which indicates that individual pins can execute excessive mechanical vibrations that may lead to damage. Such mechanical vibrations induce neutron noise that can be detected. While the detection is relatively easy, the localisation of a vibrating control rod is much more complicated because only one measuring position is available and one needs to have at least three measured quantities. Therefore it has currently been proposed that the fluctuations of the neutron current vector, called the current noise, can be used in addition to the scalar noise in reactor diagnostic problems. The thesis investigates the possibility of the localization of a vibrating control rod pin in a PWR control assembly by using the scalar neutron noise and the 2-D radial current noise as measured at one central point in the control assembly. An explicit localisation technique is elaborated in which the searched position is determined as the absolute minimum of a minimisation function. The technique is investigated in numerical simulations. The results of the simulation tests show the potential applicability of the method. By design accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and

Characters of neutron noise in full-size molten salt reactor

International Nuclear Information System (INIS)

Wang, Jiangmeng; Cao, Xinrong

2015-01-01

Highlights: • The larger system size makes full-size MSR deviate from point kinetic behavior. • The increasing velocity has non-monotonic effect on the effective delayed neutron fraction. • The amplitude of Green’s function at low frequencies is inversely proportional to the effective delayed neutron fraction. • The range of plateau region is smaller due to the more prominent point kinetic effect. - Abstract: In the present paper, the frequency-dependent and space-dependent behavior of the neutron noise in a full-size Molten Salt Reactor (MSR) is investigated. The theoretical models considering the fuel circulation are established based on one-group neutron diffusion theory. Green’s function of the neutron noise induced by a propagating perturbation is introduced with linear noise theory, due to the small perturbation. The equations are numerically calculated by developing a code, in which the eigenfunction expansion method is adopted. The static results show that the effective delayed neutron fraction changes non-monotonically with the increasing fuel velocity. In the dynamic case, the results are compared to those obtained in 1D MSR and a traditional reactor, in order to figure out the effects of both the fuel circulation and the system size. It is found that there is no difference in 1D and 3D MSR systems from the view of fuel circulation, i.e., the fuel circulation enhances the spatial neutronic coupling and leads to the stronger point kinetic effect. The more prominent space-dependent effect founded in 3D traditional reactors is also observed in the MSR, due to the looser neutronic coupling and the unique singularity of Green’s function in the location of the perturbation. Another interesting finding is that Green’s function for low frequencies changes non-monotonically with increasing velocity. The largest magnitude of Green’s function is observed at the velocity where the effective delayed neutron fraction reaches its minimum. Finally, the

The reflector effect on the neutron lifetimes in the IPEN/MB-01 reactor

International Nuclear Information System (INIS)

Gonnelli, Eduardo

2013-01-01

The aim of this study is to present the reflector effect on the neutron lifetimes in the IPEN/MB-01 reactor. The proposed method requires an approach which takes into account both the reflector and the core, so that the point kinetics equations, which constitute the theoretical basis of all mathematical development, contemplate both regions of the reactor. From these equations, as known as two regions kinetics point equations, theoretical expressions are obtained for the Auto Power Spectral Densities (APSD), which are used for least squares fit of the experimental data of APSD obtained in several subcritical states. The prompt neutron generation time, the neutron lifetimes in the reflector and the neutron return fraction from the reflector to the core are derived from the fitting. (author)

Study of the stochastic point reactor kinetic equation

International Nuclear Information System (INIS)

Gotoh, Yorio

1980-01-01

Diagrammatic technique is used to solve the stochastic point reactor kinetic equation. The method gives exact results which are derived from Fokker-Plank theory. A Green's function dressed with the clouds of noise is defined, which is a transfer function of point reactor with fluctuating reactivity. An integral equation for the correlation function of neutron power is derived using the following assumptions: 1) Green's funntion should be dressed with noise, 2) The ladder type diagrams only contributes to the correlation function. For a white noise and the one delayed neutron group approximation, the norm of the integral equation and the variance to mean-squared ratio are analytically obtained. (author)

Nuclear reactor kinetics and control

International Nuclear Information System (INIS)

Lewins, J.

1978-01-01

A consistent, integrated account of modern developments in the study of nuclear reactor kinetics and the problem of their efficient and safe control. It aims to prepare the student for advanced study and research or practical work in the field. Special features include treatments of noise theory, reliability theory and safety related studies. It covers all aspects of the operation and control of nuclear reactors, power and research and is complete in providing physical data methods of calculation and solution including questions of equipment reliability. The work uses illustrations of the main types of reactors in use in the UK, USA and Europe. Each chapter contains problems and worked examples suitable for course work and study. The subject is covered in chapters, entitled: introductory review; neutron and precursor equations; elementary solutions at low power; linear reactor process dynamics with feedback; power reactor control systems; fluctuations and reactor noise; safety and reliability; nonlinear systems (safety and control); analogue computing. (author)

Kinetics of Pressurized Water Reactors with Hot or Cold Moderators

Energy Technology Data Exchange (ETDEWEB)

Norinder, O

1960-11-15

The set of neutron kinetic equations developed in this report permits the use of long integration steps during stepwise integration. Thermal relations which describe the transfer of heat from fuel to coolant are derived. The influence upon the kinetic behavior of the reactor of a number of parameters is studied. A comparison of the kinetic properties of the hot and cold moderators is given.

Measurements of kinetic parameters by noise techniques on the MINERVE reactor

International Nuclear Information System (INIS)

Carre, J.C.; Da Costa Oliveira, J.

1975-01-01

Noise measurements were determined on ERMINE a fast thermal coupled reactor built in MINERVE. A reactor without feedback, and a reactor with an automatic control rod were both considered. The first case concerned the measurements of auto and cross power spectral density obtained with one or two neutron detectors, and the determination of: neutron lifetime; efficiency for one ion chamber; power level of the reactor; maximal speed and acceleration of the control rod for the design of an automatic reactor control actuator. The second case was concerned with measurements of the auto power spectral density in reactivity for the control rod, and the estimation of: the transfer function of the automatic pilot; the neutron lifetime; and the standard error affecting the results obtained by the oscillation method. The results proved that the pile noise theory with a point kinetic model is sufficient for application on zero power reactors. (U.K.)

Modeling Taylor series approximations for prompt neutron kinetics with lab view simulations

International Nuclear Information System (INIS)

Adzri, E. P.

2012-09-01

The reactor point kinetics equations have been subjected to intense research in an effort to find simple yet accurate numerical solutions methods. The equations are very stiff numerically, meaning that there is a wide variation in the decay constants, so that using a particular time step in the numerical solution may provide sufficient accuracy for the group, but not for another. Several solutions techniques have been presented on the point kinetics equations with varying degrees of complexity. These include Power Series Solutions, CORE, PCA, Genapol and Taylor series methods. In this research, algorithms were developed based on the first and second order Taylor series expansion and simulated in LabVIEW to solve the Reactor Point Kinetics equations using block diagram nodes implemented within stacked sequences. The algorithms developed were fast,accurate and simple to code. Several reactivity insertions were used to simulate the change in neutron population with time. The LabVIEW- Taylor series solutions were compared with other solution techniques such as Power Series Solutions, CORE, PCA, Genapol and McMahon and Pierson's Taylor series approximation. The results of LabVIEW-Taylor series technique used by McMahon and Pearson The LabVIEW-implemented techniques were found to agree very well with these other methods. At 1x10 -8 s the neutron population was 1.000220 neutrons / cm 3 , at 1 x 10 -2 s it was 2.007681 neutrons / cm 3 and at 1x10 -1 s it was 2.075317 neutrons / cm 3 ; same results reported by Genapol for a fast reactor, it produced good and accurate results and compared very favorably with other methods found in the literature. Using much smaller time steps to the order or 10 -8 s commensurate with fast reactor parameters also produced very satisfactory results, indicating that the LabVIEW-based Taylor series technique is suitable for simulating the kinetics of fast reactors as well as thermal reactors. Algorithms developed that included second order terms

Thermogravimetric analysis of reactor-neutrons-irradiated LEXAN polycarbonate film

International Nuclear Information System (INIS)

Kalsi, P.C.

2000-01-01

The effects of reactor-neutrons irradiation on the thermogravimetric (TG) analysis of LEXAN polycarbonate film in air were studied. Irradiation enhances the degradation rate and the effect increases further with increasing neutron fluence. The kinetics of the different steps of degradation were also evaluated from the TG curves. The activation energy values calculated for all the degradation stages decrease on irradiation. (author)

Kinetic studies on a repetitively pulsed fast reactor

International Nuclear Information System (INIS)

Das, S.

1982-01-01

Neutronic analysis of an earlier proposed periodically pulsed fast reactor at Kalpakkam (KPFR) has been carried out numerically under equilibrium and transient conditions using the one-point model of reactor kinetics and the experimentally measured total worth of reactivity modulator, the parabolic coefficient of reactivity of the movable reflector and the mean prompt neutron lifetime. Results of steady-state calculations - treated on the basis of delayed neutron precursor and energy balances during a period of operation - have been compared with the analytical formulae of Larrimore for a parabolic reactivity input. Empirical relations for half-width of the fast neutron pulse, the peak pulse power and the power at first crossing of prompt criticality have been obtained and shown to be accurate enough for predicting steady-state power pulse characteristics of a periodically pulsed fast reactor. The concept of a subprompt-critical reactor has been used to calculate the fictitious delayed neutron fraction, β of the KPFR through a numerical experiment. Relative pulse height stability and pulse shape sensitivity to changes of maximum reactivity is discussed. With the aid of new safety concepts, the Power Amplification Factor (PAF) and the Pulse Growth Factor (Rsub(p)), the dynamics KPFR under accidental conditions has been studied for step and ramp reactivity perturbations. All the analysis has been done without taking account of reactivity feedback. (orig.)

Sensitivity analysis of the kinetic behaviour of a Gas Cooled Fast Reactor to variations of the delayed neutron parameters

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)

Nanostructure evolution of neutron-irradiated reactor pressure vessel steels: Revised Object kinetic Monte Carlo model

Energy Technology Data Exchange (ETDEWEB)

Chiapetto, M., E-mail: mchiapet@sckcen.be [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, B-2400 Mol (Belgium); Unité Matériaux Et Transformations (UMET), UMR 8207, Université de Lille 1, ENSCL, F-59600 Villeneuve d’Ascq Cedex (France); Messina, L. [DEN-Service de Recherches de Métallurgie Physique, CEA, Université Paris-Saclay, F-91191 Gif-sur-Yvette (France); KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-114 21 Stockholm (Sweden); Becquart, C.S. [Unité Matériaux Et Transformations (UMET), UMR 8207, Université de Lille 1, ENSCL, F-59600 Villeneuve d’Ascq Cedex (France); Olsson, P. [KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-114 21 Stockholm (Sweden); Malerba, L. [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, B-2400 Mol (Belgium)

2017-02-15

This work presents a revised set of parameters to be used in an Object kinetic Monte Carlo model to simulate the microstructure evolution under neutron irradiation of reactor pressure vessel steels at the operational temperature of light water reactors (∼300 °C). Within a “grey-alloy” approach, a more physical description than in a previous work is used to translate the effect of Mn and Ni solute atoms on the defect cluster diffusivity reduction. The slowing down of self-interstitial clusters, due to the interaction between solutes and crowdions in Fe is now parameterized using binding energies from the latest DFT calculations and the solute concentration in the matrix from atom-probe experiments. The mobility of vacancy clusters in the presence of Mn and Ni solute atoms was also modified on the basis of recent DFT results, thereby removing some previous approximations. The same set of parameters was seen to predict the correct microstructure evolution for two different types of alloys, under very different irradiation conditions: an Fe-C-MnNi model alloy, neutron irradiated at a relatively high flux, and a high-Mn, high-Ni RPV steel from the Swedish Ringhals reactor surveillance program. In both cases, the predicted self-interstitial loop density matches the experimental solute cluster density, further corroborating the surmise that the MnNi-rich nanofeatures form by solute enrichment of immobilized small interstitial loops, which are invisible to the electron microscope.

Equipment for neutron measurements at VR-1 Sparrow training reactor

International Nuclear Information System (INIS)

Kolros, Antonin; Huml, Ondrej; Kos, Josef

2008-01-01

of neutron detection system in environment of VR-1 reactor was developed, verified and is used nowadays. Other application of the devices for the reactivity determination by means of the inverse kinetics method has been developed using the equation of the inverse kinetics method transformed into numeric form

Kinetic calculations for miniature neutron source reactor using analytical and numerical techniques

International Nuclear Information System (INIS)

Ampomah-Amoako, E.

2008-06-01

The analytical methods, step change in reactivity and ramp change in reactivity as well as numerical methods, fixed point iteration and Runge Kutta-gill were used to simulate the initial build up of neutrons in a miniature neutron source reactor with and without temperature feedback effect. The methods were modified to include photo neutron concentration. PARET 7.3 was used to simulate the transients behaviour of Ghana Research Reactor-1. The PARET code was capable of simulating the transients for 2.1 mk and 4 mk insertions of reactivity with peak powers of 49.87 kW and 92.34 kW, respectively. PARET code however failed to simulate 6.71 mk of reactivity which was predicted by Akaho et al through TEMPFED. (au)

The solution of the neutron point kinetics equation with stochastic extension: an analysis of two moments

Energy Technology Data Exchange (ETDEWEB)

Silva, Milena Wollmann da; Vilhena, Marco Tullio M.B.; Bodmann, Bardo Ernst J.; Vasques, Richard, E-mail: milena.wollmann@ufrgs.br, E-mail: vilhena@mat.ufrgs.br, E-mail: bardobodmann@ufrgs.br, E-mail: richard.vasques@fulbrightmail.org [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil). Programa de Pos-Graduacao em Engenharia Mecanica

2015-07-01

The neutron point kinetics equation, which models the time-dependent behavior of nuclear reactors, is often used to understand the dynamics of nuclear reactor operations. It consists of a system of coupled differential equations that models the interaction between (i) the neutron population; and (II) the concentration of the delayed neutron precursors, which are radioactive isotopes formed in the fission process that decay through neutron emission. These equations are deterministic in nature, and therefore can provide only average values of the modeled populations. However, the actual dynamical process is stochastic: the neutron density and the delayed neutron precursor concentrations vary randomly with time. To address this stochastic behavior, Hayes and Allen have generalized the standard deterministic point kinetics equation. They derived a system of stochastic differential equations that can accurately model the random behavior of the neutron density and the precursor concentrations in a point reactor. Due to the stiffness of these equations, this system was numerically implemented using a stochastic piecewise constant approximation method (Stochastic PCA). Here, we present a study of the influence of stochastic fluctuations on the results of the neutron point kinetics equation. We reproduce the stochastic formulation introduced by Hayes and Allen and compute Monte Carlo numerical results for examples with constant and time-dependent reactivity, comparing these results with stochastic and deterministic methods found in the literature. Moreover, we introduce a modified version of the stochastic method to obtain a non-stiff solution, analogue to a previously derived deterministic approach. (author)

The solution of the neutron point kinetics equation with stochastic extension: an analysis of two moments

International Nuclear Information System (INIS)

Silva, Milena Wollmann da; Vilhena, Marco Tullio M.B.; Bodmann, Bardo Ernst J.; Vasques, Richard

2015-01-01

The neutron point kinetics equation, which models the time-dependent behavior of nuclear reactors, is often used to understand the dynamics of nuclear reactor operations. It consists of a system of coupled differential equations that models the interaction between (i) the neutron population; and (II) the concentration of the delayed neutron precursors, which are radioactive isotopes formed in the fission process that decay through neutron emission. These equations are deterministic in nature, and therefore can provide only average values of the modeled populations. However, the actual dynamical process is stochastic: the neutron density and the delayed neutron precursor concentrations vary randomly with time. To address this stochastic behavior, Hayes and Allen have generalized the standard deterministic point kinetics equation. They derived a system of stochastic differential equations that can accurately model the random behavior of the neutron density and the precursor concentrations in a point reactor. Due to the stiffness of these equations, this system was numerically implemented using a stochastic piecewise constant approximation method (Stochastic PCA). Here, we present a study of the influence of stochastic fluctuations on the results of the neutron point kinetics equation. We reproduce the stochastic formulation introduced by Hayes and Allen and compute Monte Carlo numerical results for examples with constant and time-dependent reactivity, comparing these results with stochastic and deterministic methods found in the literature. Moreover, we introduce a modified version of the stochastic method to obtain a non-stiff solution, analogue to a previously derived deterministic approach. (author)

A three-dimensional nodal neutron kinetics capability for relaps

International Nuclear Information System (INIS)

Judd, J.L.; Weaver, W.L.

1996-01-01

The incorporation of a three-dimensional neutron kinetics capability into the DOE version of the RELAP5/MOD3.2 reactor safety code is discussed. A brief discussion of the kinetics method is given along with a discussion of the cross section parameterization models available in RELAP5/MOD3.2. The RELAP5/MOD3.2 code is then used to perform calculations of the NEACRP rod ejection and rod withdrawal benchmarks, and results are presented

Reactor Neutron Sources

International Nuclear Information System (INIS)

Aksenov, V.L.

1994-01-01

The present status and the prospects for development of reactor neutron sources for neutron scattering research in the world are considered. The fields of application of neutron scattering relative to synchrotron radiation, the creation stages of reactors (steady state and pulsed) and their position in comparison with spallation neutron sources at present and in the foreseen future are discussed. (author). 15 refs.; 8 figs.; 3 tabs

Neutron kinetics of fluid-fuel systems by the quasi-static method

International Nuclear Information System (INIS)

Dulla, S.; Ravetto, P.; Rostagno, M.M.

2004-01-01

The quasi-static method for the neutron kinetics of nuclear reactors is generalized for application to neutron multiplying systems fueled by a fluid multiplying material, typically a mixture of fissile molten salts. The method is derived by the application of factorization formulae for both the neutron density and the delayed precursor concentrations and the projection of the balance equations upon a weighting function. A physically meaningful weight can be assumed as the solution of the adjoint model, which is constructed for the situation considered, including delayed neutrons. The quasi-static scheme is then applied to calculations of some transients for a typical configuration of a molten-salt reactor, in a multigroup diffusion model with a one-dimensional slug-flow velocity field. The physical features associated to the motion of the fissile material are highlighted

Kinetics, dynamics and neutron noise in Molten Salt Reactors

International Nuclear Information System (INIS)

Pazsit, Imre

2013-01-01

Reactor kinetic and dynamic properties of Molten Salt Reactors (MSR) are investigated in a simple model, which allows closed compact analytical solutions to be obtained. The goal is to gain insight, rather than to produce high-quality quantitative data. Through an interpretation of the different terms in the basic equations, and by means of analytical solutions, various approximations are introduced and their validity discussed. The dynamical behaviour of MSRs and their response to small stationary perturbations is described and discussed in comparison with traditional systems. (author)

Study on the numerical analysis of nuclear reactor kinetics equations

International Nuclear Information System (INIS)

Yang, J.C.

1980-01-01

A two-step alternating direction explict method is proposed for the solution of the space-and time-dependent diffusion theory reactor kinetics equations in two space dimensions as a special case of the general class of alternating direction implicit method and the truncation error of this method is estimated. To test the validity of this method it is applied to the Pressurized Water Reactor and CANDU-PHW reactor which have been operating and underconstructing in Korea. The time dependent neutron flux of the PWR reactor during control rod insertion and time dependent neutronic power of CANDU-PHW reactor in the case of postulated loss of coolant accident are obtained from the numerical calculation results. The results of the PWR reactor problem are shown the close agreement between implicit-difference method used in the TWIGL program and this method, and the results of the CANDU-PHW reactor are compared with the results of improved quasistic method and modal method. (Author)

Neutron behavior, reactor control, and reactor heat transfer. Volume four

International Nuclear Information System (INIS)

Anon.

1986-01-01

Volume four covers neutron behavior (neutron absorption, how big are nuclei, neutron slowing down, neutron losses, the self-sustaining reactor), reactor control (what is controlled in a reactor, controlling neutron population, is it easy to control a reactor, range of reactor control, what happens when the fuel burns up, controlling a PWR, controlling a BWR, inherent safety of reactors), and reactor heat transfer (heat generation in a nuclear reactor, how is heat removed from a reactor core, heat transfer rate, heat transfer properties of the reactor coolant)

Neutron source for a reactor

International Nuclear Information System (INIS)

Kobayashi, Hiromasa.

1975-01-01

Object: To easily increase a start-up power of a reactor without irradiation in other reactors. Structure: A neutron source comprises Cf 252 , a natural antimony rod, a layer of beryllium, and a vessel of neutron source. On upper and lower portion of Cf 252 are arranged natural antimony rods, which are surrounded by the Be layer, the entirety being charged into the vessel. The Cf 252 may emit neutron, has a half life more than a period of operating cycle of the reactor and is less deteriorated even irradiated by radioactive rays while being left within the reactor. The natural antimony rod is radioactivated by neutron from Cf 252 and neutron as reactor power increases to emit γ rays. The Be absorbs γ rays to emit the neutron. The antimony rod is irradiated within the reactor. Further, since the Cf 252 is small in neutron absorption cross section, it is hard to be deteriorated even while being inserted within the reactor. (Kamimura, M.)

On a closed form approach to the fractional neutron point kinetics equation with temperature feedback

International Nuclear Information System (INIS)

Schramm, Marcelo; Bodmann, Bardo E.J.; Vilhena, Marco T.M.B.; Petersen, Claudio Z.; Alvim, Antonio C.M.

2013-01-01

Following the quest to find analytical solutions, we extend the methodology applied successfully to timely fractional neutron point kinetics (FNPK) equations by adding the effects of temperature. The FNPK equations with temperature feedback correspond to a nonlinear system and “stiff” type for the neutron density and the concentration of delayed neutron precursors. These variables determine the behavior of a nuclear reactor power with time and are influenced by the position of control rods, for example. The solutions of kinetics equations provide time information about the dynamics in a nuclear reactor in operation and are useful, for example, to understand the power fluctuations with time that occur during startup or shutdown of the reactor, due to adjustments of the control rods. The inclusion of temperature feedback in the model introduces an estimate of the transient behavior of the power and other variables, which are strongly coupled. Normally, a single value of reactivity is used across the energy spectrum. Especially in case of power change, the neutron energy spectrum changes as well as physical parameters such as the average cross sections. However, even knowing the importance of temperature effects on the control of the reactor power, the character of the set of nonlinear equations governing this system makes it difficult to obtain a purely analytical solution. Studies have been published in this sense, using numerical approaches. Here the idea is to consider temperature effects to make the model more realistic and thus solve it in a semi-analytical way. Therefore, the main objective of this paper is to obtain an analytical representation of fractional neutron point kinetics equations with temperature feedback, without having to resort to approximations inherent in numerical methods. To this end, we will use the decomposition method, which has been successfully used by the authors to solve neutron point kinetics problems. The results obtained will

On a closed form approach to the fractional neutron point kinetics equation with temperature feedback

Energy Technology Data Exchange (ETDEWEB)

Schramm, Marcelo; Bodmann, Bardo E.J.; Vilhena, Marco T.M.B., E-mail: marceloschramm@hotmail.com, E-mail: bardo.bodmann@ufrgs.br, E-mail: mtmbvilhena@gmail.com [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil). Departamento de Engenharia Mecanica; Petersen, Claudio Z., E-mail: claudiopetersen@yahoo.com.br [Universidade Federal de Pelotas (UFPel), RS (Brazil). Departamento de Matematica; Alvim, Antonio C.M., E-mail: alvim@nuclear.ufrj.br [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Instituto Alberto Luiz Coimbra de Pos-Graduacao e Pesquisa em Engenharia

2013-07-01

Following the quest to find analytical solutions, we extend the methodology applied successfully to timely fractional neutron point kinetics (FNPK) equations by adding the effects of temperature. The FNPK equations with temperature feedback correspond to a nonlinear system and “stiff” type for the neutron density and the concentration of delayed neutron precursors. These variables determine the behavior of a nuclear reactor power with time and are influenced by the position of control rods, for example. The solutions of kinetics equations provide time information about the dynamics in a nuclear reactor in operation and are useful, for example, to understand the power fluctuations with time that occur during startup or shutdown of the reactor, due to adjustments of the control rods. The inclusion of temperature feedback in the model introduces an estimate of the transient behavior of the power and other variables, which are strongly coupled. Normally, a single value of reactivity is used across the energy spectrum. Especially in case of power change, the neutron energy spectrum changes as well as physical parameters such as the average cross sections. However, even knowing the importance of temperature effects on the control of the reactor power, the character of the set of nonlinear equations governing this system makes it difficult to obtain a purely analytical solution. Studies have been published in this sense, using numerical approaches. Here the idea is to consider temperature effects to make the model more realistic and thus solve it in a semi-analytical way. Therefore, the main objective of this paper is to obtain an analytical representation of fractional neutron point kinetics equations with temperature feedback, without having to resort to approximations inherent in numerical methods. To this end, we will use the decomposition method, which has been successfully used by the authors to solve neutron point kinetics problems. The results obtained will

Beam transient analyses of Accelerator Driven Subcritical Reactors based on neutron transport method

Energy Technology Data Exchange (ETDEWEB)

He, Mingtao; Wu, Hongchun [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049, Shaanxi (China); Zheng, Youqi, E-mail: yqzheng@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049, Shaanxi (China); Wang, Kunpeng [Nuclear and Radiation Safety Center, PO Box 8088, Beijing 100082 (China); Li, Xunzhao; Zhou, Shengcheng [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049, Shaanxi (China)

2015-12-15

Highlights: • A transport-based kinetics code for Accelerator Driven Subcritical Reactors is developed. • The performance of different kinetics methods adapted to the ADSR is investigated. • The impacts of neutronic parameters deteriorating with fuel depletion are investigated. - Abstract: The Accelerator Driven Subcritical Reactor (ADSR) is almost external source dominated since there is no additional reactivity control mechanism in most designs. This paper focuses on beam-induced transients with an in-house developed dynamic analysis code. The performance of different kinetics methods adapted to the ADSR is investigated, including the point kinetics approximation and space–time kinetics methods. Then, the transient responds of beam trip and beam overpower are calculated and analyzed for an ADSR design dedicated for minor actinides transmutation. The impacts of some safety-related neutronics parameters deteriorating with fuel depletion are also investigated. The results show that the power distribution varying with burnup leads to large differences in temperature responds during transients, while the impacts of kinetic parameters and feedback coefficients are not very obvious. Classification: Core physic.

Reactor neutron dosimetry

International Nuclear Information System (INIS)

Najzer, M.; Pauko, M.; Glumac, B.; Acquah, I.N.; Moskon, F.

1977-01-01

An analysis of requirements and possibilities for experimental neutron spectrum determination during the reactor pressure vessel surveil lance programme is given. Fast neutron spectrum and neutron dose rate were measured in the Fast neutron irradiation facility of our TRIGA reactor. It was shown that the facility can be used for calibration of neutron dosimeters and for irradiation of samples sensitive to neutron radiation. The investigation of the unfolding algorithm ITER was continued. Based on this investigations are two specialized unfolding program packages ITERAD and ITERGS written this year. They are able to unfold data from activation detectors and NaI(T1) gamma spectrometer respectively

Prompt Neutron Lifetime for the NBSR Reactor

Energy Technology Data Exchange (ETDEWEB)

Hanson, A.L.; Diamond, D.

2012-06-24

In preparation for the proposed conversion of the National Institute of Standards and Technology (NIST) research reactor (NBSR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel, certain point kinetics parameters must be calculated. We report here values of the prompt neutron lifetime that have been calculated using three independent methods. All three sets of calculations demonstrate that the prompt neutron lifetime is shorter for the LEU fuel when compared to the HEU fuel and longer for the equilibrium end-of-cycle (EOC) condition when compared to the equilibrium startup (SU) condition for both the HEU and LEU fuels.

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

Application of the exact distribution pjk in the determination of kinetic parameters in a reactor

International Nuclear Information System (INIS)

Alcala Ruiz, F.

1982-01-01

In this report one distribution of neutron counts obtained by a detector placed in a reactor is studied in order to be used in the determination of reactor kinetic parameters such as β/Λ and reactivities. The parameters accuracy from this new method is compared with the Feynman and Mogilner method, based too in Reactor Neutron Noise Analysis. These three methods have been applied to JEN-2 reactor and the better accuracy and faster collection of experimental data give some interest to the new method which only requires a good footing code. (Author) 68 refs

Development of 3-dimensional neutronics kinetics analysis code for CANDU-PHWR

International Nuclear Information System (INIS)

Kim, M. W.; Kim, C. H.; Hong, I. S.

2005-02-01

The followings are the major contents and scope of the research : development of kinetics power calculation module, formulation of space-dependent neutron transient analysis - implementation of 3-D and 2-G unified nodal method, verification of the kinetics module by benchmark problem - 3-D PHWR kinetics benchmark problem suggested by AECL, reactor trip simulation by shutdown system 1 in Wolsong unit 2. Development of a dynamic linked library code, SCAN D LL, for the coupled calculation with RELAP-CANDU : modeling of shutdown system 1, development of automatic shutdown module - automatic trip module based on rate log power control logic, automatic insertion of shutdown system 1. Development of a link code for coupled calculation - development of SCAN D LL(windows version), verification of coupled code by - 40% reactor inlet header break LOCA power pulse, 100% reactor outlet header break LOCA power pulse, 50% pump suction break LOCA power pulse

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)

Semiconductor research with reactor neutrons

International Nuclear Information System (INIS)

Kimura, Itsuro

1992-01-01

Reactor neutrons play an important role for characterization of semiconductor materials as same as other advanced materials. On the other hand reactor neutrons bring about not only malignant irradiation effects called radiation damage, but also useful effects such as neutron transmutation doping and defect formation for opto-electronics. Research works on semiconductor materials with the reactor neutrons of the Kyoto University Reactor (KUR) are briefly reviewed. In this review, a stress is laid on the present author's works. (author)

Evaluation of energy collapsing effect on reactor kinetics parameters by diffusion theory

International Nuclear Information System (INIS)

Unesaki, Hironobu

1989-01-01

Reactor kinetics parameters play an important role as scaling factors between observed and calculated reactivities in the analysis of reactor physics experiments. In this report, energy collapsing errors in two kinetic parameters, the effective delayed neutron fraction and the neutron life time, are investigated by means of the diffusion theory. Coarse group calculations are made for various energy group structures. Cores of various moderator-to-fuel volume ratios are selected to investigate the influence of neutron spectrum changes on the energy collapsing error. The energy collapsing errors in the effective delayed neutron fraction and neutron life time are much larger than those in k eff . This might be because the former two parameters are functions of both the foward and adjoint flux, whereas the latter is a function of the forward flux alone. The use of coarse constants will cause errors in both fluxes, and the resulting errors in the former will be much more emphasized. As the effective delayed neutron fraction is sensitive to the treatment of an energy region in the vicinity of the fission spectrum peak, the coarse group error in it might differ between cores with different enrichment and composition. Inaccurate weighting of group constants leads to neutron spectra which do not conserve the fine group spectra, and those errors will be emphasized in calculated integral parameters. (N.K.)

R-102, 1 Group Space-Independent Inverse Reactor Kinetics

International Nuclear Information System (INIS)

Kaganove, J.J.

1966-01-01

1 - Description of problem or function: Given the space-independent, one energy group reactor kinetics equations and the initial conditions, this program determines the time variation of reactivity required to produce the given input of flux-time data. 2 - Method of solution: Time derivatives of neutron density are obtained by application of (a) five-point quartic, (b) three-point parabolic, (c) five-point least-mean-square cubic, (d) five-point least-mean-square parabolic, or (e) five-point least-mean-square linear formulae to the neutron density or to the natural logarithm of the neutron density. Between each data point the neutron density is assumed to be (a) exponential*(third-order polynomial), (b) exponential, or (c) linear. Changes in reactivity between data points are obtained algebraically from the kinetics equations, neutron density derivatives, and the algebraic representation of neutron density. First and second time derivatives of the reactivity are obtained by use of any of the formulae applicable to the neutron density. 3 - Restrictions on the complexity of the problem: Maxima of - 50 delay groups; 1000 data points; 99 data blocks (A data block is a sequence of input points characterized by a fixed time-interval between points, a smoothing option, and a number of repetitions of the smoothing option)

CORTAP: a coupled neutron kinetics-heat transfer digital computer program for the dynamic simulation of the high temperature gas cooled reactor core

International Nuclear Information System (INIS)

Cleveland, J.C.

1977-01-01

CORTAP (Core Transient Analysis Program) was developed to predict the dynamic behavior of the High Temperature Gas Cooled Reactor (HTGR) core under normal operational transients and postulated accident conditions. CORTAP is used both as a stand-alone component simulation and as part of the HTGR nuclear steam supply (NSS) system simulation code ORTAP. The core thermal neutronic response is determined by solving the heat transfer equations for the fuel, moderator and coolant in an average powered region of the reactor core. The space independent neutron kinetics equations are coupled to the heat transfer equations through a rapidly converging iterative technique. The code has the capability to determine conservative fuel, moderator, and coolant temperatures in the ''hot'' fuel region. For transients involving a reactor trip, the core heat generation rate is determined from an expression for decay heat following a scram. Nonlinear effects introduced by temperature dependent fuel, moderator, and coolant properties are included in the model. CORTAP predictions will be compared with dynamic test results obtained from the Fort St. Vrain reactor owned by Public Service of Colorado, and, based on these comparisons, appropriate improvements will be made in CORTAP

Analytical solution of point kinetics equations for linear reactivity variation during the start-up of a nuclear reactor

International Nuclear Information System (INIS)

Palma, Daniel A.P.; Martinez, Aquilino S.; Goncalves, Alessandro C.

2009-01-01

The analytical solution of point kinetics equations with a group of delayed neutrons is useful in predicting the variation of neutron density during the start-up of a nuclear reactor. In the practical case of an increase of nuclear reactor power resulting from the linear insertion of reactivity, the exact analytical solution cannot be obtained. Approximate solutions have been obtained in previous articles, based on considerations that need to be verifiable in practice. In the present article, an alternative analytic solution is presented for point kinetics equations in which the only approximation consists of disregarding the term of the second derivative for neutron density in relation to time. The results proved satisfactory when applied to practical situations in the start-up of a nuclear reactor through the control rods withdraw.

Analytical solution of point kinetics equations for linear reactivity variation during the start-up of a nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Palma, Daniel A.P. [CEFET QUIMICA de Nilopolis/RJ, 21941-914 Rio de Janeiro (Brazil)], E-mail: agoncalves@con.ufrj.br; Martinez, Aquilino S.; Goncalves, Alessandro C. [COPPE/UFRJ - Programa de Engenharia Nuclear, Rio de Janeiro (Brazil)

2009-09-15

The analytical solution of point kinetics equations with a group of delayed neutrons is useful in predicting the variation of neutron density during the start-up of a nuclear reactor. In the practical case of an increase of nuclear reactor power resulting from the linear insertion of reactivity, the exact analytical solution cannot be obtained. Approximate solutions have been obtained in previous articles, based on considerations that need to be verifiable in practice. In the present article, an alternative analytic solution is presented for point kinetics equations in which the only approximation consists of disregarding the term of the second derivative for neutron density in relation to time. The results proved satisfactory when applied to practical situations in the start-up of a nuclear reactor through the control rods withdraw.

Need for nuclear data for thermal neutron reactors

International Nuclear Information System (INIS)

Bouchard, J.; Golinelli, C.; Tellier, H.

1983-01-01

The need for nuclear data for thermal neutron reactors is conditioned by the persisting lack of agreement between the calculation and measurement of certain parameters, by the benefit that can be drawn from reduction of the marginal areas and by envisaged modifications. Three particular fields are delineated. Reduction of the deviation in temperature coefficients by modification of the shape of the effective capture cross sections of uranium-238 and -235 in the thermal range. The increase in precision of kinetic measurements by a better knowledge of data connected to slowed-down neutrons. Improvement in predicting the neutron activity of the fuels used in measuring the effective capture cross sections of plutonium-242 and americium-243. (Auth.)

Effect of Utilization of Silicide Fuel with the Density 4.8 gU/cc on the Kinetic Parameters of RSG-GAS Reactor

International Nuclear Information System (INIS)

Setiyanto; Sembiring, Tagor M.; Pinem, Surian

2007-01-01

Presently, the RSG-GAS reactor using silicide fuel element of 2.96 gU/cc. For increasing reactor operation time, its planning to change to higher density fuel. The kinetic calculation of silicide core with density 4.8 gU/cc has been carried out, since it has an influence on the reactor operation safety. The calculated kinetic parameters are the effective delayed neutron fraction, the delayed neutron decay constant, prompt neutron lifetime and feedback reactivity coefficient very important for reactor operation safety. the calculation is performed in 2-dimensional neutron diffusion-perturbation method using modified Batan-2DIFF code. The calculation showed that the effective delayed neutron fraction is 7. 03256x10 -03 , total delay neutron time constant is 7.85820x10 -02 s -1 and the prompt neutron lifetime is 55.4900 μs. The result of prompt neutron lifetime smaller 10 % compare with silicide fuel of 4.8 gU/cc. The calculated results showed that all of the feedback reactivity coefficient silicide core 4.8 gU/cc is negative. Totally, the feedback reactivity coefficient of silicide fuel of 4.8 gU/cc is 10% less than that of silicide fuel of 2.96 gU/cc. The results shown that kinetic parameters result decrease compared with the silicide core with density 2.96 gU/cc, but no significant influence in the RSG-GAS reactor operation. (author)

Analytic solutions of the multigroup space-time reactor kinetics equations

International Nuclear Information System (INIS)

Lee, C.E.; Rottler, S.

1986-01-01

The development of analytical and numerical solutions to the reactor kinetics equations is reviewed. Analytic solutions of the multigroup space-time reactor kinetics equations are developed for bare and reflected slabs and spherical reactors for zero flux, zero current and extrapolated endpoint boundary conditions. The material properties of the reactors are assumed constant in space and time, but spatially-dependent source terms and initial conditions are investigated. The system of partial differential equations is reduced to a set of linear ordinary differential equations by the Laplace transform method. These equations are solved by matrix Green's functions yielding a general matrix solution for the neutron flux and precursor concentration in the Laplace transform space. The detailed pole structure of the Laplace transform matrix solutions is investigated. The temporally- and spatially-dependent solutions are determined from the inverse Laplace transform using the Cauchy residue theorem, the theorem of Frobenius, a knowledge of the detailed pole structure and matrix operators. (author)

Neutronics and thermohydraulics of the reactor C.E.N.E. Pt. 1

International Nuclear Information System (INIS)

Caro, R.; Ahnert, C.; Esteban Naudin, A.; Martinez Fanegas, R.; Minguez, E.; Rovira, A.

1976-01-01

The analysis of neutronics (both statics and kinetics), of the 10 Mwt swimming pool reactor C.E.N.E. is included. A short description of the theoretical model used, along with the theoretical versus experimental cheking, carried out, whenever possible, with the reactors JEN-1 and JEN-2 of Junta de Energia Nuclear, is given in each of these chapters. (author) [es

First records of thermal neutrons with the spectrometer for time of flight (TOF) in the RP-10 Nuclear Reactor

International Nuclear Information System (INIS)

Munive, M.; Baltuano, O; Soto, C; Ravello, Y

2002-01-01

To obtain the first spectrum of an emergent beam of neutrons of a nuclear reactor is the main parameter of the characterization in the use of this reactor; one of ways to get this spectrum is for the technique of time of flight, TOF, which registers the time that a neutron need to cover a certain distance, associating this time then to the kinetic energy of the neutron. The kinetic study of the beam of neutrons is carried out on neutron pulses that are generated by a revolving choke called Chopper; and the analysis in the time of the detected pulses is carried out for a system MCS. Using this technique it is achieved the record of the spectra in energy, or in wavelength , of the irradiation facilities No 2 and 4, and of the exit N o 5 of the thermal column of the Nuclear Reactor RP-10 of the Nuclear Center Oscar de la Guerra RACSO, Peru (au)

Effect of lattice-level adjoint-weighting on the kinetics parameters of CANDU reactors

International Nuclear Information System (INIS)

Nichita, Eleodor

2009-01-01

Space-time kinetics calculations for CANDU reactors are routinely performed using the Improved Quasistatic (IQS) method. The IQS method calculates kinetics parameters such as the effective delayed-neutron fraction and generation time using adjoint weighting. In the current implementation of IQS, the direct flux, as well as the adjoint, is calculated using a two-group cell-homogenized reactor model which is inadequate for capturing the effect of the softer energy spectrum of the delayed neutrons. Additionally, there may also be fine spatial effects that are lost because the intra-cell adjoint shape is ignored. The purpose of this work is to compare the kinetics parameters calculated using the two-group cell-homogenized model with those calculated using lattice-level fine-group heterogeneous adjoint weighting and to assess whether the differences are large enough to justify further work on incorporating lattice-level adjoint weighting into the IQS method. A second goal is to evaluate whether the use of a fine-group cell-homogenized lattice-level adjoint, such as is the current practice for Light Water Reactors (LWRs), is sufficient to capture the lattice effects in question. It is found that, for CANDU lattices, the generation time is almost unaffected by the type of adjoint used to calculate it, but that the effective delayed-neutron fraction is affected by the type of adjoint used. The effective delayed-neutron fraction calculated using the two-group cell-homogenized adjoint is 5.2% higher than the 'best' effective delayed-neutron fraction value obtained using the detailed lattice-level fine-group heterogeneous adjoint. The effective delayed-neutron fraction calculated using the fine-group cell-homogenized adjoint is only 1.7% higher than the 'best' effective delayed-neutron fraction value but is still not equal to it. This situation is different from that encountered in LWRs where weighting by a fine-group cell-homogenized adjoint is sufficient to calculate the

Experimental methods of investigation of kinetics and dynamics of nuclear reactors

International Nuclear Information System (INIS)

Costa Oliveira, Jaime M.

1969-03-01

The author presents experimental methods used to study kinetic and dynamic properties of nuclear reactors. Kinetic methods aim at determining characteristic parameters of the behaviour in time of neutrons. Dynamic methods aim at establishing the relationships between the reactor behaviour and its internal and external causes (notably the measurement of transfer functions). The author proposes a classification with respect to the excitation type: periodic excitation (reactivity sinusoidal modulation, source sinusoidal modulation, periodic pulse excitation), non periodic excitation (reactivity monitoring, reactivity linear variation, reactivity variation according to any given law, removal of starting source), random excitation (random reactivity or source excitation), natural fluctuations (alpha-Rossi method, methods of reduced variance, probabilistic methods, correlation methods, spectral analysis method). He also addresses space and energy effects. Applications are reported for low power and power reactors

An efficient technique for the point reactor kinetics equations with Newtonian temperature feedback effects

International Nuclear Information System (INIS)

Nahla, Abdallah A.

2011-01-01

Highlights: → An efficient technique for the nonlinear reactor kinetics equations is presented. → This method is based on Backward Euler or Crank Nicholson and fundamental matrix. → Stability of efficient technique is defined and discussed. → This method is applied to point kinetics equations of six-groups of delayed neutrons. → Step, ramp, sinusoidal and temperature feedback reactivities are discussed. - Abstract: The point reactor kinetics equations of multi-group of delayed neutrons in the presence Newtonian temperature feedback effects are a system of stiff nonlinear ordinary differential equations which have not any exact analytical solution. The efficient technique for this nonlinear system is based on changing this nonlinear system to a linear system by the predicted value of reactivity and solving this linear system using the fundamental matrix of the homogenous linear differential equations. The nonlinear point reactor kinetics equations are rewritten in the matrix form. The solution of this matrix form is introduced. This solution contains the exponential function of a variable coefficient matrix. This coefficient matrix contains the unknown variable, reactivity. The predicted values of reactivity in the explicit form are determined replacing the exponential function of the coefficient matrix by two kinds, Backward Euler and Crank Nicholson, of the rational approximations. The nonlinear point kinetics equations changed to a linear system of the homogenous differential equations. The fundamental matrix of this linear system is calculated using the eigenvalues and the corresponding eigenvectors of the coefficient matrix. Stability of the efficient technique is defined and discussed. The efficient technique is applied to the point kinetics equations of six-groups of delayed neutrons with step, ramp, sinusoidal and the temperature feedback reactivities. The results of these efficient techniques are compared with the traditional methods.

RELAP5 kinetics model development for the Advanced Test Reactor

International Nuclear Information System (INIS)

Judd, J.L.; Terry, W.K.

1990-01-01

A point-kinetics model of the Advanced Test Reactor has been developed for the RELAP5 code. Reactivity feedback parameters were calculated by a three-dimensional analysis with the PDQ neutron diffusion code. Analyses of several hypothetical reactivity insertion events by the new model and two earlier models are discussed. 3 refs., 10 figs., 6 tabs

BR2 reactor neutron beams

International Nuclear Information System (INIS)

Neve de Mevergnies, M.

1977-01-01

The use of reactor neutron beams is becoming increasingly more widespread for the study of some properties of condensed matter. It is mainly due to the unique properties of the ''thermal'' neutrons as regards wavelength, energy, magnetic moment and overall favorable ratio of scattering to absorption cross-sections. Besides these fundamental reasons, the impetus for using neutrons is also due to the existence of powerful research reactors (such as BR2) built mainly for nuclear engineering programs, but where a number of intense neutron beams are available at marginal cost. A brief introduction to the production of suitable neutron beams from a reactor is given. (author)

Neutronics and thermohydraulics of the reactor C.E.N.E.-Part I

International Nuclear Information System (INIS)

Caro, R.; Ahnert, C.; Naudin, A. E.; Martinez Fanegas, R.; Minguez, E.; Rovira, A.

1976-01-01

In this report the analysis of neutronics (both statics and kinetics), of the 10 MWt swimming pool reactor C.E.N.E, is included. In each of these chapters is given a short description of the theoretical model used, along with the theoretical versus experimental checking, carried out, whenever possible, with the reactors JEN-I and JEN-II of Junta de Energia Nuclear. (Author) 11 refs

Neutron inverse kinetics via Gaussian Processes

International Nuclear Information System (INIS)

Picca, Paolo; Furfaro, Roberto

2012-01-01

Highlights: ► A novel technique for the interpretation of experiments in ADS is presented. ► The technique is based on Bayesian regression, implemented via Gaussian Processes. ► GPs overcome the limits of classical methods, based on PK approximation. ► Results compares GPs and ANN performance, underlining similarities and differences. - Abstract: The paper introduces the application of Gaussian Processes (GPs) to determine the subcriticality level in accelerator-driven systems (ADSs) through the interpretation of pulsed experiment data. ADSs have peculiar kinetic properties due to their special core design. For this reason, classical – inversion techniques based on point kinetic (PK) generally fail to generate an accurate estimate of reactor subcriticality. Similarly to Artificial Neural Networks (ANNs), Gaussian Processes can be successfully trained to learn the underlying inverse neutron kinetic model and, as such, they are not limited to the model choice. Importantly, GPs are strongly rooted into the Bayes’ theorem which makes them a powerful tool for statistical inference. Here, GPs have been designed and trained on a set of kinetics models (e.g. point kinetics and multi-point kinetics) for homogeneous and heterogeneous settings. The results presented in the paper show that GPs are very efficient and accurate in predicting the reactivity for ADS-like systems. The variance computed via GPs may provide an indication on how to generate additional data as function of the desired accuracy.

Research and Development Program in Reactor Diagnostics and Monitoring with Neutron Noise Methods. Stages 14 and 15

International Nuclear Information System (INIS)

Pazsit, Imre; Wihlstrand, Gustav; Tambouratzis, Tatiana; Jonsson, Anders; Dahl, Berit

2009-12-01

This report constitutes Stages 14 and 15 of a long-term research and development program concerning the development of diagnostics and monitoring methods for nuclear reactors. Stage 14 was a full one-year project, whereas Stage 15 consisted of a half-year project. The program executed in Stages 14 and 15 consists of the following three parts: - Study of criticality, neutron kinetics and neutron noise in molten salt reactors (MSR) (Stages 14 and 15); - An overview and introduction to fuzzy logics (Stage 14), and an application to two-phase flow identification (Stage 15) - Preparations for and execution of an IAEA-ICTP workshop on Neutron fluctuations, reactor noise and their applications in nuclear reactors (Stage 14)

Research and Development Program in Reactor Diagnostics and Monitoring with Neutron Noise Methods. Stages 14 and 15

Energy Technology Data Exchange (ETDEWEB)

Pazsit, Imre; Wihlstrand, Gustav; Tambouratzis, Tatiana; Jonsson, Anders; Dahl, Berit (Chalmers Univ. of Technology, Dept. of Nuclear Engineering, SE-412 96 Goeteborg (Sweden))

2009-12-15

This report constitutes Stages 14 and 15 of a long-term research and development program concerning the development of diagnostics and monitoring methods for nuclear reactors. Stage 14 was a full one-year project, whereas Stage 15 consisted of a half-year project. The program executed in Stages 14 and 15 consists of the following three parts: - Study of criticality, neutron kinetics and neutron noise in molten salt reactors (MSR) (Stages 14 and 15); - An overview and introduction to fuzzy logics (Stage 14), and an application to two-phase flow identification (Stage 15) - Preparations for and execution of an IAEA-ICTP workshop on Neutron fluctuations, reactor noise and their applications in nuclear reactors (Stage 14)

Inverse kinetics method with source term for subcriticality measurements during criticality approach in the IPEN/MB-01 research reactor

International Nuclear Information System (INIS)

Loureiro, Cesar Augusto Domingues; Santos, Adimir dos

2009-01-01

In reactor physics tests which are performed at the startup after refueling the commercial PWRs, it is important to monitor subcriticality continuously during criticality approach. Reactivity measurements by the inverse kinetics method are widely used during the operation of a nuclear reactor and it is possible to perform an online reactivity measurement based on the point reactor kinetics equations. This technique is successful applied at sufficiently high power level or to a core without an external neutron source where the neutron source term in point reactor kinetics equations may be neglected. For operation at low power levels, the contribution of the neutron source must be taken into account and this implies the knowledge of a quantity proportional to the source strength, and then it should be determined. Some experiments have been performed in the IPEN/MB-01 Research Reactor for the determination of the Source Term, using the Least Square Inverse Kinetics Method (LSIKM). A digital reactivity meter which neglects the source term is used to calculate the reactivity and then the source term can be determined by the LSIKM. After determining the source term, its value can be added to the algorithm and the reactivity can be determined again, considering the source term. The new digital reactivity meter can be used now to monitor reactivity during the criticality approach and the measured value for the reactivity is more precise than the meter which neglects the source term. (author)

A new formulation for the importance function in the kinetics of subcritical reactors

International Nuclear Information System (INIS)

Silva, Cristiano da; Senra Martinez, Aquilino; Carvalho da Silva, Fernando

2012-01-01

Highlights: ► In this paper we propose a new formulation for the importance function in the kinetics of subcritical systems. ► We analyze the relevance of an external neutron source for the subcritical interval 0.95 eff eff is the multiplication factor according to the physical properties of the nuclear reactor. For the purposes of validation of the proposed method we will use, as a reference method, the expansion in modes of the time-dependent neutron flux for the solution of the onedimensional diffusion equation. It will be presented results that demonstrate the precision of the proposed method when compared to the conventional point kinetic equations. The results show that the new point kinetic equations are rather precise in the subcriticality range considered.

Multigroup perturbation model for kinetic analysis of nuclear reactors

International Nuclear Information System (INIS)

Souza, G.M.

1989-01-01

The scope of this work is the development of a multigroup perturbation theory for the purpose of Kinetic and dynamic analysis of nuclear reactors. The equations that describe the reactor behavior were presented in all generality and written in the shorthand notation of matrices and vectors. In the derivation of those equations indetermined operators and discretizing factors were introduced and then determined by comparision with conventional equations. Fick's Law was developed in higher orders for neutron and importance current density. The solution of the direct and adjoint fields were represented by combination of the eigenfunctions of the B and B* operators and the eigenvalue modulus equality was established mathematically. In the derivation of the reactivity expression the B operator perturbation was split in two non coupled to the flux form and level. The prompt neutrons effective mean life was derived from reactor equations and importance conservation. The establishment of the Nordheim's equation, although modified, was based on Gandini. Finally, a mathematical interpretation of the flux-trap region was avented. (author)

Neutronics methods for transient and safety analysis of fast reactors

Energy Technology Data Exchange (ETDEWEB)

Marchetti, Marco

2017-07-01

Modeling the evolution of possible or postulated accidents in nuclear reactors is fundamental in designing safe systems. For the next generation of reactors, in particular fast reactors, fuel movement during an accident can, in principle, drive an energetic event. Such is the issue of recriticality. The thermal energy produced during these events will, possibly, be converted into mechanical energy by some mechanisms. For example, the nuclear heat deposited in the fuel could cause fuel vaporization and its subsequent expansion. This movement would accelerate the surrounding sodium: part of the initial energy in the fuel is thus converted into sodium kinetic energy. This mechanical energy will finally be absorbed, in some way or another, by the reactor vessel. Providing an accurate estimate for the maximum mechanical work that any accidental sequence can do onto the reactor vessel is an essential step in designing a reactor containment that would withstand any load generated by any accident. That would assure accident containment, without consequences for the general public. Fast reactor accident modeling is a complicated task. The outcome of an accident is determined by different physical phenomena, all acting at almost the same time. Safety analysts must track all these different phenomena. Multi-physics codes have been developed for this task. They must contain accurate models for fluid-dynamics, neutronics, and structures. This work has to do with neutronics modeling of such accidents. Past and recent analyses have been limited to the approximate description of the neutronic field, for example by using a rough description of the energy and/or of the angular dependence of the neutron flux. In this work, different neutronic solvers are selected and coupled into a general multi-physics code for fast reactor accident analysis. Performances of each of them is then assessed. Some emphasis has been put also in assessing the speed of these solvers for determining the

Verification of a three-dimensional neutronics model based on multi-point kinetics equations for transient problems

Energy Technology Data Exchange (ETDEWEB)

Park, Kyung Seok; Kim, Hyun Dae; Yeom, Choong Sub [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-07-01

A computer code for solving the three-dimensional reactor neutronic transient problems utilizing multi-point reactor kinetics equations recently developed has been developed. For evaluating its applicability, the code has been tested with typical 3-D LWR and CANDU reactor transient problems. The performance of the method and code has been compared with the results by fine and coarse meshes computer codes employing the direct methods.

NEUTRONIC REACTOR

Science.gov (United States)

Wigner, E.P.; Weinberg, A.W.; Young, G.J.

1958-04-15

A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.

Neutron-nucleus interactions and fission. Chapter 1

International Nuclear Information System (INIS)

1998-01-01

The central problem in nuclear-reactor kinetics is to predict the evolution in time of the neutron population in a multiplying medium. Point kinetics allows study of the global behaviour of the neutron population from the average properties of the medium. Before tackling, in the following chapters, the equations governing the time variation of the reactor power (proportional to the total neutron population), the properties of a neutron-multiplying medium shall be discussed briefly. After recalling a number of definitions, a qualitative description shall be given of the principal nuclear reactions at play in a self-sustaining chain reaction, with emphasis on the source of fission neutrons. Since delayed neutrons play a crucial role in reactor kinetics, their production in a reactor shall be described in greater detail. (author)

Advanced methodology to simulate boiling water reactor transient using coupled thermal-hydraulic/neutron-kinetic codes

Energy Technology Data Exchange (ETDEWEB)

Hartmann, Christoph Oliver

2016-06-13

Coupled Thermal-hydraulic/Neutron-kinetic (TH/NK) simulations of Boiling Water Reactor transients require well validated and accurate simulation tools. The generation of cross-section (XS) libraries, depending on the individual thermal-hydraulic state parameters, is of paramount importance for coupled simulations. Problem-dependent XS-sets for 3D core simulations are being generated mainly by well validated, fast running commercial and user-friendly lattice codes such as CASMO and HELIOS. In this dissertation a computational route, based on the lattice code SCALE6/TRITON, the cross-section interface GenPMAXS, the best-estimate thermal-hydraulic system code TRACE and the core simulator PARCS, for best-estimate simulations of Boiling Water (BWR) transients has been developed and validated. The computational route has been supplemented by a subsequent uncertainty and sensitivity study based on Monte Carlo sampling and propagation of the uncertainties of input parameters to the output (SUSA code). The analysis of a single BWR fuel assembly depletion problem with PARCS using SCALE/TRITON cross-sections has been shown a good agreement with the results obtained with CASMO cross-section sets. However, to compensate the deficiencies of the interface program GenPMAXS, PYTHON scripts had to be developed to incorporate missing data, as the yields of Iodine, Xenon and Promethium, into the cross-section-data sets (PMAXS-format) generated by GenPMAXS from the SCALE/TRITON output. The results of the depletion analysis of a full BWR core with PARCS have indicated the importance of considering history effects, adequate modeling of the reflector region and the control rods, as the PARCS simulations for depleted fuel and all control rods inserted (ARI) differs significantly at the fuel assembly top and bottom. Systematic investigations with the coupled codes TRACE/PARCS have been performed to analyse the core behaviour at different thermal conditions using nuclear data (XS

Non-equilibrium statistical thermodynamics of neutron gas in reactor

International Nuclear Information System (INIS)

Hayasaka, Hideo

1977-01-01

The thermodynamic structures of non-equilibrium steady states of highly rarefied neutron gas in various media are considered for the irreversible processes owing to creative and destructive reactions of neutrons with nuclei of these media and supply from the external sources. Under the so-called clean and cold condition in reactor, the medium is regarded virtually as offering the different chemical potential fields for each subsystem of a steady neutron gas system. The fluctuations around a steady state are considered in a Markovian-Gaussian process. The generalized Einstein relations are derived for stationary neutron gas systems. The forces and flows of neutron gases in a medium are defined upon the general stationary solution of the Fokker-Planck equation. There exist the symmetry of the kinetic coefficients, and the minimum entropy production upon neutron-nuclear reactions. The distribution functions in various media are determined by each corresponding extremum condition under the vanishing of changes of the respective total entropies in the Gibbs equation. (auth.)

Experimental estimations of the kinetics parameters of the IBR-2M reactor by stochastic noises

International Nuclear Information System (INIS)

Pepelyshev, Yu.N.; Tajybov, L.A.; Garibov, A.A.; Mekhtieva, R.N.

2012-01-01

Experimental investigations of stochastic fluctuations of pulse energy of the IBR-2M reactor have been carried out which allowed us to obtain some of the parameters of the reactor kinetics. At different levels of average power a sequence of values of pulse energy was recorded with the calculation of the distribution parameters. An ionization chamber with boron installed near the active zone was used as a neutron detector. The research results allowed us to estimate the average lifetime of prompt neutrons τ = (6.53±0.2)·10 -8 s, absolute power of the reactor and intensity of the source of spontaneous neutrons S sp ≤(6.72±0.12)·10 6 s -1 . It was shown that the experimental results are close to the calculated ones

Development of a point-kinetic verification scheme for nuclear reactor applications

Energy Technology Data Exchange (ETDEWEB)

Demazière, C., E-mail: demaz@chalmers.se; Dykin, V.; Jareteg, K.

2017-06-15

In this paper, a new method that can be used for checking the proper implementation of time- or frequency-dependent neutron transport models and for verifying their ability to recover some basic reactor physics properties is proposed. This method makes use of the application of a stationary perturbation to the system at a given frequency and extraction of the point-kinetic component of the system response. Even for strongly heterogeneous systems for which an analytical solution does not exist, the point-kinetic component follows, as a function of frequency, a simple analytical form. The comparison between the extracted point-kinetic component and its expected analytical form provides an opportunity to verify and validate neutron transport solvers. The proposed method is tested on two diffusion-based codes, one working in the time domain and the other working in the frequency domain. As long as the applied perturbation has a non-zero reactivity effect, it is demonstrated that the method can be successfully applied to verify and validate time- or frequency-dependent neutron transport solvers. Although the method is demonstrated in the present paper in a diffusion theory framework, higher order neutron transport methods could be verified based on the same principles.

One dimensional neutron kinetics in the TRAC-BF1 code

International Nuclear Information System (INIS)

Weaver, W.L. III; Wagner, K.C.

1987-01-01

The TRAC-BWR code development program at the Idaho National Engineering Laboratory is developing a version of the TRAC code for the U.S. Nuclear Regulatory Commission (USNRC) to provide a best-estimate analysis capability for the simulation of postulated accidents in boiling water reactor (BWR) power systems and related experimental facilities. Recent development efforts in the TRAC-BWR program have focused on improving the computational efficiency through the incorporation of a hybrid Courant- limit-violating numerical solution scheme in the one-dimensional component models and on improving code accuracy through the development of a one-dimensional neutron kinetics model. Many other improvements have been incorporated into TRAC-BWR to improve code portability, accuracy, efficiency, and maintainability. This paper will describe the one- dimensional neutron kinetics model, the generation of the required input data for this model, and present results of the first calculations using the model

Coupled hydro-neutronic calculations for fast burst reactor accidents

International Nuclear Information System (INIS)

Paternoster, R.; Kimpland, R.; Jaegers, P.; McGhee, J.

1994-01-01

Methods are described for determining the fully coupled neutronic/hydrodynamic response of fast burst reactors (FBR) under disruptive accident conditions. Two code systems, PAD (1 -D Lagrangian) and NIKE-PAGOSA (3-D Eulerian) were used to accomplish this. This is in contrast to the typical methodology that computes these responses by either single point kinetics or in a decoupled manner. This methodology is enabled by the use of modem supercomputers (CM-200). Two examples of this capability are presented: an unreflected metal fast burst assembly, and a reflected fast burst assembly typical of the Skua or SPR-III class of fast burst reactor

Opportunities for TRIGA reactors in neutron radiography

International Nuclear Information System (INIS)

Barton, John P.

1978-01-01

In this country the two most recent installations of TRIGA reactors have both been for neutron radiography, one at HEDL and the other at ANL. Meanwhile, a major portion of the commercial neutron radiography is performed on a TRIGA fueled reactor at Aerotest. Each of these installations has different primary objectives and some comparative observations can be drawn. Another interesting comparison is between the TRIGA reactors for neutron radiography and other small reactors that are being installed for this purpose such as the MIRENE slow pulse reactors in France, a U-233 fueled reactor for neutron radiography in India and the L88 solution reactor in Denmark. At Monsanto Laboratory, in Ohio, a subcritical reactor based on MTR-type fuel has recently been purchased for neutron radiography. Such systems, when driven by a Van de Graaff neutron source, will be compared with the standard TRIGA reactor. Future demands on TRIGA or competitive systems for neutron radiography are likely to include the pulsing capability of the reactor, and also the extraction of cold neutron beams and resonance energy beams. Experiments recently performed on the Oregon State TRIGA Reactor provide information in each of these categories. A point of particular current concern is a comparison made between the resonance energy beam intensity extracted from the edge of the TRIGA core and from a slot which penetrated to the center of the TREAT reactor. These results indicate that by using such slots on a TRIGA, resonance energy intensities could be extracted that are much higher than previously predicted. (author)

Measurements of neutron flux in the RA reactor

International Nuclear Information System (INIS)

Raisic, N.

1961-12-01

This report includes the following separate parts: Thermal neutron flux in the experimental channels od RA reactor; Epithermal neutron flux in the experimental channels od RA reactor; Fast neutron flux in the experimental channels od RA reactor; Thermal neutron flux in the thermal column and biological experimental channel; Neutronic measurements in the RA reactor cell; Temperature reactivity coefficient of the RA reactor; design of the device for measuring the activity of wire [sr

Neutronic reactor

International Nuclear Information System (INIS)

Wende, C.W.J.

1976-01-01

A safety rod for a nuclear reactor has an inner end portion having a gamma absorption coefficient and neutron capture cross section approximately equal to those of the adjacent shield, a central portion containing materials of high neutron capture cross section and an outer end portion having a gamma absorption coefficient at least equal to that of the adjacent shield

Studies on the molten salt reactor. Code development and neutronics analysis of MSRE-type design

International Nuclear Information System (INIS)

Zhuang Kun; Cao Liangzhi; Zheng Youqi; Wu Hongchun

2015-01-01

The molten salt reactor is characterized by its use of the fluid-fuel, which serves both as a fuel and as a coolant simultaneously. The position of delayed neutron precursors continuously changes both in the core and in the external loop due to the fuel circulation, and the fission products are extracted by an online fuel reprocessing unit, which all lead to the modeling methods for the conventional reactors using solid fuel not applicable. This study establishes suitable calculation models for the neutronics analysis of the molten salt reactor and develops a new code named MOREL based on the three-dimensional diffusion steady and transient calculations. Some numerical tests are chosen to verify the code and the numerical results indicate that MOREL can be used for the analysis of the molten salt reactor. After verification, it is applied to analyze the characteristics of a typical molten salt reactor, including the steady characteristics, the influence of fuel circulation on the kinetic behaviors. Besides, the influence of online fuel reprocessing simulation is also examined. The results show that inherent safety is the character of the molten salt reactor from the aspect of reactivity feedback and the fuel circulation has great influence on the kinetic characteristics of molten salt reactor. (author)

Neutronic of heterogenous gas cooled reactors

International Nuclear Information System (INIS)

Maturana, Roberto Hernan

2008-01-01

At present, one of the main technical features of the advanced gas cooled reactor under development is its fuel element concept, which implies a neutronic homogeneous design, thus requiring higher enrichment compared with present commercial nuclear power plants.In this work a neutronic heterogeneous gas cooled reactor design is analyzed by studying the neutronic design of the Advanced Gas cooled Reactor (AGR), a low enrichment, gas cooled and graphite moderated nuclear power plant.A search of merit figures (some neutronic parameter, characteristic dimension, or a mixture of both) which are important and have been optimized during the reactor design stage is been done, to aim to comprise how a gas heterogeneous reactor is been design, given that semi-infinity arrangement criteria of rods in LWRs and clusters in HWRs can t be applied for a solid moderator and a gas refrigerator.The WIMS code for neutronic cell calculations is been utilized to model the AGR fuel cell and to calculate neutronic parameters such as the multiplication factor and the pick factor, as function of the fuel burnup.Also calculation is been done for various nucleus characteristic dimensions values (fuel pin radius, fuel channel pitch) and neutronic parameters (such as fuel enrichment), around the design established parameters values.A fuel cycle cost analysis is carried out according to the reactor in study, and the enrichment effect over it is been studied.Finally, a thermal stability analysis is been done, in subcritical condition and at power level, to study this reactor characteristic reactivity coefficients.Present results shows (considering the approximation used) a first set of neutronic design figures of merit consistent with the AGR design. [es

COSTANZA-AX, 1-D Neutronics and Thermodynamics of Liquid Cooled Reactor in Axial Geometry. COSTANZA-CYL, 1-D Neutronics and Thermodynamics of Liquid Cooled Reactor in Cylindrical Geometry

International Nuclear Information System (INIS)

Agazzi, A.; Forti, G.; Vincenti, E.

1984-01-01

1 - Nature of physical problem solved: Purpose of the programmes is to study reactor dynamics, considering the variation of the spatial flux distribution. The two programmes COSTANZA-CYL and COSTANZA-AX, solve the kinetics diffusion equations in two groups and one dimension (plane geometry for COSTANZA-AX, radial geometry for COSTANZA-CYL). The neutronic calculation is coupled with the calculation of the heat transmission from the fuel to the cladding and to the coolant, and with the thermo-hydraulics of channels with forced circulation of liquid coolant. The geometry of fuel element and channel may be cylindrical or slab. Up to ten groups of delayed neutrons are allowed. Temperature feedback of fuel (Doppler) and coolant are considered independently and affect the nuclear constants. Control rod movement or diffused poison concentrations are simulated by externally imposed variations of the thermal absorption cross section in the different regions of the reactors. Inlet temperatures and mass flow in the coolant channels may be varied according to any externally given time table. 2 - Method of solution: The kinetic diffusion equations in two groups are solved by finite-difference method. 3 - Restrictions on the complexity of the problem: 10 concentric regions; 10 coolant channels; 10 groups of delayed neutrons

A Derivation of Source-based Kinetics Equation with Time Dependent Fission Kernel for Reactor Transient Analyses

International Nuclear Information System (INIS)

Kim, Song Hyun; Woo, Myeong Hyun; Shin, Chang Ho; Pyeon, Cheol Ho

2015-01-01

In this study, a new balance equation to overcome the problems generated by the previous methods is proposed using source-based balance equation. And then, a simple problem is analyzed with the proposed method. In this study, a source-based balance equation with the time dependent fission kernel was derived to simplify the kinetics equation. To analyze the partial variations of reactor characteristics, two representative methods were introduced in previous studies; (1) quasi-statics method and (2) multipoint technique. The main idea of quasistatics method is to use a low-order approximation for large integration times. To realize the quasi-statics method, first, time dependent flux is separated into the shape and amplitude functions, and shape function is calculated. It is noted that the method has a good accuracy; however, it can be expensive as a calculation cost aspect because the shape function should be fully recalculated to obtain accurate results. To improve the calculation efficiency, multipoint method was proposed. The multipoint method is based on the classic kinetics equation with using Green's function to analyze the flight probability from region r' to r. Those previous methods have been used to analyze the reactor kinetics analysis; however, the previous methods can have some limitations. First, three group variables (r g , E g , t g ) should be considered to solve the time dependent balance equation. This leads a big limitation to apply large system problem with good accuracy. Second, the energy group neutrons should be used to analyze reactor kinetics problems. In time dependent problem, neutron energy distribution can be changed at different time. It can affect the change of the group cross section; therefore, it can lead the accuracy problem. Third, the neutrons in a space-time region continually affect the other space-time regions; however, it is not properly considered in the previous method. Using birth history of the neutron sources

G4-STORK: A Geant4-based Monte Carlo reactor kinetics simulation code

International Nuclear Information System (INIS)

Russell, Liam; Buijs, Adriaan; Jonkmans, Guy

2014-01-01

Highlights: • G4-STORK is a new, time-dependent, Monte Carlo code for reactor physics applications. • G4-STORK was built by adapting and expanding on the Geant4 Monte Carlo toolkit. • G4-STORK was designed to simulate short-term fluctuations in reactor cores. • G4-STORK is well suited for simulating sub- and supercritical assemblies. • G4-STORK was verified through comparisons with DRAGON and MCNP. - Abstract: In this paper we introduce G4-STORK (Geant4 STOchastic Reactor Kinetics), a new, time-dependent, Monte Carlo particle tracking code for reactor physics applications. G4-STORK was built by adapting and expanding on the Geant4 Monte Carlo toolkit. The toolkit provides the fundamental physics models and particle tracking algorithms that track each particle in space and time. It is a framework for further development (e.g. for projects such as G4-STORK). G4-STORK derives reactor physics parameters (e.g. k eff ) from the continuous evolution of a population of neutrons in space and time in the given simulation geometry. In this paper we detail the major additions to the Geant4 toolkit that were necessary to create G4-STORK. These include a renormalization process that maintains a manageable number of neutrons in the simulation even in very sub- or supercritical systems, scoring processes (e.g. recording fission locations, total neutrons produced and lost, etc.) that allow G4-STORK to calculate the reactor physics parameters, and dynamic simulation geometries that can change over the course of simulation to illicit reactor kinetics responses (e.g. fuel temperature reactivity feedback). The additions are verified through simple simulations and code-to-code comparisons with established reactor physics codes such as DRAGON and MCNP. Additionally, G4-STORK was developed to run a single simulation in parallel over many processors using MPI (Message Passing Interface) pipes

Calculation of the neutron importance and weighted neutron generation time using MCNIC method in accelerator driven subcritical reactors

Energy Technology Data Exchange (ETDEWEB)

Hassanzadeh, M. [Nuclear Science and Technology Research Institute, AEOI, Tehran, Islamic Republic of Iran (Iran, Islamic Republic of); Feghhi, S.A.H., E-mail: a_feghhi@sbu.ac.ir [Department of Radiation Application, Shahid Beheshti University, G.C., Tehran, Islamic Republic of Iran (Iran, Islamic Republic of); Khalafi, H. [Nuclear Science and Technology Research Institute, AEOI, Tehran, Islamic Republic of Iran (Iran, Islamic Republic of)

2013-09-15

Highlights: • All reactor kinetic parameters are importance weighted quantities. • MCNIC method has been developed for calculating neutron importance in ADSRs. • Mean generation time has been calculated in spallation driven systems. -- Abstract: The difference between non-weighted neutron generation time (Λ) and the weighted one (Λ{sup †}) can be quite significant depending on the type of the system. In the present work, we will focus on developing MCNIC method for calculation of the neutron importance (Φ{sup †}) and importance weighted neutron generation time (Λ{sup †}) in accelerator driven systems (ADS). Two hypothetic bare and graphite reflected spallation source driven system have been considered as illustrative examples for this means. The results of this method have been compared with those obtained by MCNPX code. According to the results, the relative difference between Λ and Λ{sup †} is within 36% and 24,840% in bare and reflected illustrative examples respectively. The difference is quite significant in reflected systems and increases with reflector thickness. In Conclusion, this method may be used for better estimation of kinetic parameters rather than the MCNPX code because of using neutron importance function.

Calculation of the neutron importance and weighted neutron generation time using MCNIC method in accelerator driven subcritical reactors

International Nuclear Information System (INIS)

Hassanzadeh, M.; Feghhi, S.A.H.; Khalafi, H.

2013-01-01

Highlights: • All reactor kinetic parameters are importance weighted quantities. • MCNIC method has been developed for calculating neutron importance in ADSRs. • Mean generation time has been calculated in spallation driven systems. -- Abstract: The difference between non-weighted neutron generation time (Λ) and the weighted one (Λ † ) can be quite significant depending on the type of the system. In the present work, we will focus on developing MCNIC method for calculation of the neutron importance (Φ † ) and importance weighted neutron generation time (Λ † ) in accelerator driven systems (ADS). Two hypothetic bare and graphite reflected spallation source driven system have been considered as illustrative examples for this means. The results of this method have been compared with those obtained by MCNPX code. According to the results, the relative difference between Λ and Λ † is within 36% and 24,840% in bare and reflected illustrative examples respectively. The difference is quite significant in reflected systems and increases with reflector thickness. In Conclusion, this method may be used for better estimation of kinetic parameters rather than the MCNPX code because of using neutron importance function

Programmed elimination of neutronic poisons in nuclear reactors

International Nuclear Information System (INIS)

Perriere, G. de la

1967-11-01

This work deals with the use of salts of elements having a large neutron capture cross-section, so-called 'soluble poisons' which are dissolved in the moderating water to control the reactivity of heavy-water reactors, and more particularly to compensate the xenon effect in the reactor EL 4. The report describes the controlled elimination of these poisons by fixation on ion-exchange resins. The poisons considered are lithium-6, cadmium and gadolinium in the sulphate form, and boron as boric acid. The thermodynamic and kinetic constants of the ion-exchange reactions were first determined and a study was then made of the fixation of these compounds in beds of small-calibre resins placed in columns. Lithium-6 is the poison which is most easily applicable to compensate the xenon effect in the reactor EL 4. It can be eliminated rapidly and completely from heavy water, and its use does not lead to supplementary problems of protection against the gamma radiation of the reactor circuits. (author) [fr

Study on thermal neutron spectra in reactor moderators by time-of-flight method

International Nuclear Information System (INIS)

Akino, Fujiyoshi

1982-12-01

Prediction of thermal neutron spectra in a reactor core plays very important role in the neutronic design of the reactor for obtaining the accurate thermal group constants. It is well known that the neutron scattering properties of the moderator materials markedly influence the thermal neutron spectra. Therefore, 0 0 angular dependent thermal neutron spectra were measured by the time-of-flight method in the following moderator bulks 1) Graphite bulk poisoned with boron at the temperatures from 20 to 800 0 C, 2) Light water bulk poisoned with Cadmium and/or Indium, 3) Light water-natural uranium heterogeneous bulk. The measured results were compared with calculation utilizing Young-Koppel and Haywood scattering model for graphite and light water respectively. On the other hand, a variety of 20% enriched uranium loaded and graphite moderated cores consisting of the different lattice cell in a wide range of the carbon to uranium atomic ratio have been built at Semi-Homogeneous Critical Experimental Assembly (SHE) to perform the critical experiments related to Very High Temperature Reactor (VHTR). The experimental data were for the critical masses in 235 U, reactivity worths of experimental burnable poison rods, thorium rods, natural-uranium rods and experimental control rods and kinetic parameters. It is made clear from comparison between measurement and calculation that the accurate thermal group constants can be obtained by use of the Young-Koppel and Haywood neutron scattering models if heterogeneity of reactor core lattices is taken into account precisely. (author)

Description of the CAREM Reactor Neutronic Calculation Codes; Descripcion de la Linea de Calculo Neutronica del CAREM

Energy Technology Data Exchange (ETDEWEB)

Villarino, Eduardo; Hergenreder, Daniel [Investigacion Aplicada, INVAP, San Carlos de Bariloche (Argentina)

2000-07-01

In this work is described the neutronic calculation line used to design the CAREM reactor.A description of the codes used and the interfaces between the different programs are presented.Both, the normal calculation line and the alternative or verification calculation line are included.The calculation line used to obtain the kinetics parameters (effective delayed-neutron fraction and prompt-neutron lifetime) is also included.

Analytic method study of point-reactor kinetic equation when cold start-up

International Nuclear Information System (INIS)

Zhang Fan; Chen Wenzhen; Gui Xuewen

2008-01-01

The reactor cold start-up is a process of inserting reactivity by lifting control rod discontinuously. Inserting too much reactivity will cause short-period and may cause an overpressure accident in the primary loop. It is therefore very important to understand the rule of neutron density variation and to find out the relationships among the speed of lifting control rod, and the duration and speed of neutron density response. It is also helpful for the operators to grasp the rule in order to avoid a start-up accident. This paper starts with one-group delayed neutron point-reactor kinetics equations and provides their analytic solution when reactivity is introduced by lifting control rods discontinuously. The analytic expression is validated by comparison with practical data. It is shown that the analytic solution agrees well with numerical solution. Using this analytical solution, the relationships among neutron density response with the speed of lifting control rod and its duration are also studied. By comparing the results with those under the condition of step inserted reactivity, useful conclusions are drawn

Roles of plasma neutron source reactor in development of fusion reactor engineering: Comparison with fission reactor engineering

International Nuclear Information System (INIS)

Hirayama, Shoichi; Kawabe, Takaya

1995-01-01

The history of development of fusion power reactor has come to a turning point, where the main research target is now shifting from the plasma heating and confinement physics toward the burning plasma physics and reactor engineering. Although the development of fusion reactor system is the first time for human beings, engineers have experience of development of fission power reactor. The common feature between them is that both are plants used for the generation of nuclear reactions for the production of energy, nucleon, and radiation on an industrial scale. By studying the history of the development of the fission reactor, one can find the existence of experimental neutron reactors including irradiation facilities for fission reactor materials. These research neutron reactors played very important roles in the development of fission power reactors. When one considers the strategy of development of fusion power reactors from the points of fusion reactor engineering, one finds that the fusion neutron source corresponds to the neutron reactor in fission reactor development. In this paper, the authors discuss the roles of the plasma-based neutron source reactors in the development of fusion reactor engineering, by comparing it with the neutron reactors in the history of fission power development, and make proposals for the strategy of the fusion reactor development. 21 refs., 6 figs

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

Neutron beam facilities at the Australian Replacement Research Reactor

International Nuclear Information System (INIS)

Kennedy, Shane; Robinson, Robert; Hunter, Brett

2001-01-01

Australia is building a research reactor to replace the HIFAR reactor at Lucas Heights by the end of 2005. Like HIFAR, the Replacement Research Reactor will be multipurpose with capabilities for both neutron beam research and radioisotope production. It will be a pool-type reactor with thermal neutron flux (unperturbed) of 4 x 10 14 n/cm 2 /sec and a liquid D 2 cold neutron source. Cold and thermal neutron beams for neutron beam research will be provided at the reactor face and in a large neutron guide hall. Supermirror neutron guides will transport cold and thermal neutrons to the guide hall. The reactor and the associated infrastructure, with the exception of the neutron beam instruments, is to be built by INVAP S.E. under contract. The neutron beam instruments will be developed by ANSTO, in consultation with the Australian user community. This status report includes a review the planned scientific capabilities, a description of the facility and a summary of progress to date. (author)

Monitor for reactor neutron detector

International Nuclear Information System (INIS)

Shirakami, Hisayuki; Shibata, Masatoshi

1992-01-01

The device of the present invention judges as to whether a neutron detector is normal or not while considering the change of indication value depending on the power change of a reactor core. That is, the device of the present invention comprises a standard value setting device for setting the standard value for calibrating the neutron detector and an abnormality judging device for comparing the standard value with a measured value of the neutron detector and judging the abnormality when the difference is greater than a predetermined value. The measured value upon initialization of each of the neutron detectors is determined as a quasi-standard value. An average value of the difference between the measured value and the quasi-standard value of a plurality of effective neutron detectors at a same level for the height of the reactor core is multiplied to a power rate based on the reactor core power at a position where the neutron detector is disposed upon calibration. The value obtained by adding the multiplied value and the quasi-standard value is determined as a standard value. The abnormality judging device compares the standard value with the measured value of the neutron detector and, if the difference is greater than a predetermined value, the neutron detector is determined as abnormal. As a result, judgement can be conducted more accurately than conventional cases. (I.S.)

Calculation of Kinetic Parameters of TRIGA Reactor

International Nuclear Information System (INIS)

Snoj, Luka; Kavcic, Andrej; Zerovnik, Gasper; Ravnik, Matjaz

2008-01-01

Modern Monte Carlo transport codes in combination of fast computer clusters enable very accurate calculations of the most important reactor kinetic parameters. Such are the effective delayed neutron fraction, β eff , and mean neutron generation time, Λ. We calculated the β eff and Λ for various realistic and hypothetical annular TRIGA Mark II cores with different types and amount of fuel. It can be observed that the effective delayed neutron fraction strongly depends on the number of fuel elements in the core or on the core size. E.g., for 12 wt. % uranium standard fuel with 20 % enrichment, β eff varies from 0.0080 for a small core (43 fuel rods) to 0.0075 for a full core (90 fuel rods). It is interesting to note that calculated value of β eff strongly depends also on the delayed neutron nuclear data set used in calculations. The prompt neutron life-time mainly depends on the amount (due to either content or enrichment) of 235 U in the fuel as it is approximately inversely proportional to the average absorption cross-section of the fuel. E.g., it varies from 28 μs for 30 wt. % uranium content fuelled core to 48 μs for 8.5 wt. % uranium content LEU fuelled core. The results are especially important for pulse mode operation and analysis of the pulses. (authors)

Report on neutron reflectometry for the Australian Replacement Reactor

International Nuclear Information System (INIS)

James, M.

2001-01-01

There is a clear need for at least one neutron reflectometer at the Australian Replacement Research Reactor when it commences operation in 2005. The participants at the reflectometry workshop have identified that the neutron reflectometer to be built at the Australian Replacement Research Reactor must be capable of the study of: 1. Specular scattering from air/solid, solid/liquid and in particular 'free liquid' samples; and 2. Off-specular' scattering from the above sample types. 3. Kinetics phenomena on a minute or slower time scale; 4. A range of samples of differing thicknesses, ranging from ultra-thin films to thousand angstrom thick films. In order to achieve this the reflectometer should have the capacity to vary its resolution. Interest was also expressed at the ability to conduct glancing-angle and wide-angle scattering studies for the investigation of short length scale, in-plane structures. There was little interest expressed by the workshop participants for polarised neutron reflectometry. This report contains a scientific case for a neutron reflectometer to be built at the Australian Replacement Research Reactor on a cold neutron guide, which is based on the areas of scientific research expressed by the workshop participants. In addition, trends in neutron reflectometry research conducted at major overseas neutron facilities are noted. The new neutron Reflectometer should: 1. Be based on the Time-of-Flight method; 2. Have a vertical scattering plane (i.e. operate for horizontal samples); 3. Be located on the end of a cold neutron guide, or be built off the guide axis using a bender, 4. Have a position sensitive area detector, 5. Be similar in spirit to the new D17 reflectometer at the ILL. Basic aspects of a reflectometer design are discussed which meet the above-stated scientific criteria and include a preliminary list of instrument specifications, capabilities and ancillary equipment requested by the workshop participants. A preliminary instrument

Development of neutronic models for the thermal hydraulics coupling of the MSFR and the calculation of effective kinetic parameters

International Nuclear Information System (INIS)

Laureau, Axel

2015-01-01

In this PhD thesis, we describe the development of innovative neutronic models for their coupling with thermal hydraulics such that they combine precision and reasonable computational times. One of the main cases where this method is applied is the Molten Salt Fast Reactor (MSFR) whose combines a fast neutron spectrum with a thorium cycle. In this fourth generation reactor, the motion of the delayed neutron precursors and the associated phenomena have to be taken into account due to the liquid fuel circulation. The starting point for these developments was the preliminary design of this type of system where a dedicated multi-physical representation was needed to study the reactor performance in steady and transient conditions. As a first step, a stationary coupling was developed. A neutronic model based on a stochastic approach was associated to a CFD (Computational Fluid Dynamics) code to solve the Navier Stokes equations for turbulent flows and the transport of the delayed neutron precursors. The impact of this precursor motion is taken into account by reconstructing the prompt shower that they generate. This approach, called by shower, views the critical reactor as a prompt subcritical reactor that amplifies a source of delayed neutrons. A second step consisted in developing a neutronic model based on a time dependent version of the fission matrices (Transient Fission Matrix or TFM) so as to enable reactor transient studies. With the TFM model, an initial computation of the matrices with a stochastic code (MCNP, SERPENT) allows the characterization of the global spatial and time dependent neutronic response of the reactor with a precision close to that of a Monte Carlo calculation. The information thus obtained is then used to calculate transients, while retaining the advantage of reduced computational time. The TFM model, which can be used for various system concepts, also allows the evaluation of effective kinetic parameters such as the effective fraction of

Reactor physics and reactor computations

International Nuclear Information System (INIS)

Ronen, Y.; Elias, E.

1994-01-01

Mathematical methods and computer calculations for nuclear and thermonuclear reactor kinetics, reactor physics, neutron transport theory, core lattice parameters, waste treatment by transmutation, breeding, nuclear and thermonuclear fuels are the main interests of the conference

Calculations on neutron irradiation damage in reactor materials

International Nuclear Information System (INIS)

Sone, Kazuho; Shiraishi, Kensuke

1976-01-01

Neutron irradiation damage calculations were made for Mo, Nb, V, Fe, Ni and Cr. Firstly, damage functions were calculated as a function of neutron energy with neutron cross sections of elastic and inelastic scatterings, and (n,2n) and (n,γ) reactions filed in ENDF/B-III. Secondly, displacement damage expressed in displacements per atom (DPA) was estimated for neutron environments such as fission spectrum, thermal neutron reactor (JMTR), fast breeder reactor (MONJU) and two fusion reactors (The Conceptual Design of Fusion Reactor in JAERI and ORNL-Benchmark). then, damage cross section in units of dpa. barn was defined as a factor to convert a given neutron fluence to the DPA value, and was calculated for the materials in the above neutron environments. Finally, production rates of helium and hydrogen atoms were calculated with (n,α) and (n,p) cross sections in ENDF/B-III for the materials irradiated in the above reactors. (auth.)

Neutron noise in nuclear reactors

International Nuclear Information System (INIS)

Blaquiere, A.; Pachowska, R.

1961-06-01

The power of a nuclear reactor, in the operating conditions, presents fluctuations due to various causes. This random behaviour can be included in the study of 'noises'. Among other sources of noise, we analyse hereafter the fluctuations due: a) to the discontinuous emissions of neutrons from an independent source; b) to the multiplication of neutrons inside the reactor. The method which we present makes use of the analogies between the rules governing a nuclear reactor in operation and a number of radio-electrical systems, in particular the feed-back loops. The reactor can be characterized by its 'passing band' and is described as a system submitted to a sequence of random pulses. In non linear operating condition, the effect of neutron noise is defined by means of a non-linear functional, this theory is thus related to previous works the references of which are given at the end of the present report. This leads us in particular in the case of nuclear reactors to some results given by A. Blaquiere in the case of radio-electrical loops. (author) [fr

A Derivation of Source-based Kinetics Equation with Time Dependent Fission Kernel for Reactor Transient Analyses

Energy Technology Data Exchange (ETDEWEB)

Kim, Song Hyun; Woo, Myeong Hyun; Shin, Chang Ho [Hanyang University, Seoul (Korea, Republic of); Pyeon, Cheol Ho [Kyoto University, Osaka (Japan)

2015-10-15

In this study, a new balance equation to overcome the problems generated by the previous methods is proposed using source-based balance equation. And then, a simple problem is analyzed with the proposed method. In this study, a source-based balance equation with the time dependent fission kernel was derived to simplify the kinetics equation. To analyze the partial variations of reactor characteristics, two representative methods were introduced in previous studies; (1) quasi-statics method and (2) multipoint technique. The main idea of quasistatics method is to use a low-order approximation for large integration times. To realize the quasi-statics method, first, time dependent flux is separated into the shape and amplitude functions, and shape function is calculated. It is noted that the method has a good accuracy; however, it can be expensive as a calculation cost aspect because the shape function should be fully recalculated to obtain accurate results. To improve the calculation efficiency, multipoint method was proposed. The multipoint method is based on the classic kinetics equation with using Green's function to analyze the flight probability from region r' to r. Those previous methods have been used to analyze the reactor kinetics analysis; however, the previous methods can have some limitations. First, three group variables (r{sub g}, E{sub g}, t{sub g}) should be considered to solve the time dependent balance equation. This leads a big limitation to apply large system problem with good accuracy. Second, the energy group neutrons should be used to analyze reactor kinetics problems. In time dependent problem, neutron energy distribution can be changed at different time. It can affect the change of the group cross section; therefore, it can lead the accuracy problem. Third, the neutrons in a space-time region continually affect the other space-time regions; however, it is not properly considered in the previous method. Using birth history of the

Neutronics and thermohydraulics of the reactor C.E.N.E.-Part I; Analisis neutronico y termohidraulico del reactor C.E.N.E. Parte I

Energy Technology Data Exchange (ETDEWEB)

Caro, R; Ahnert, C; Naudin, A E; Martinez Fanegas, R; Minguez, E; Rovira, A

1976-07-01

In this report the analysis of neutronics (both statics and kinetics), of the 10 MWt swimming pool reactor C.E.N.E, is included. In each of these chapters is given a short description of the theoretical model used, along with the theoretical versus experimental checking, carried out, whenever possible, with the reactors JEN-I and JEN-II of Junta de Energia Nuclear. (Author) 11 refs.

Nuclear reactor, fuel assembly and neutron measuring system

International Nuclear Information System (INIS)

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

1998-01-01

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

Solution of Point Reactor Neutron Kinetics Equations with Temperature Feedback by Singularly Perturbed Method

Directory of Open Access Journals (Sweden)

Wenzhen Chen

2013-01-01

Full Text Available The singularly perturbed method (SPM is proposed to obtain the analytical solution for the delayed supercritical process of nuclear reactor with temperature feedback and small step reactivity inserted. The relation between the reactivity and time is derived. Also, the neutron density (or power and the average density of delayed neutron precursors as the function of reactivity are presented. The variations of neutron density (or power and temperature with time are calculated and plotted and compared with those by accurate solution and other analytical methods. It is shown that the results by the SPM are valid and accurate in the large range and the SPM is simpler than those in the previous literature.

Reactor-moderated intermediate-energy neutron beams for neutron-capture therapy

International Nuclear Information System (INIS)

Less, T.J.

1987-01-01

One approach to producing an intermediate energy beam is moderating fission neutrons escaping from a reactor core. The objective of this research is to evaluate materials that might produce an intermediate beam for NCT via moderation of fission neutrons. A second objective is to use the more promising moderator material in a preliminary design of an NCT facility at a research reactor. The evaluations showed that several materials or combinations of materials could produce a moderator source for an intermediate beam for NCT. The best neutron spectrum for use in NCT is produced by Al 2 O 3 , but mixtures of Al metal and D 2 O are also attractive. Using the best moderator materials, results were applied to the design of an NCT moderator at the Georgia Institute of Technology Research Reactor's bio-medical facility. The amount of photon shielding and thermal neutron absorber were optimized with respect to the desired photon dose rate and intermediate neutron flux at the patient position

Fundamentals of 3-D Neutron Kinetics and Current Status

International Nuclear Information System (INIS)

Aragones, J.M.

2008-01-01

This lecture includes the following topics: 1) A summary of the cell and lattice calculations used to generate the neutron reaction data for neutron kinetics, including the spectral and burnup calculations of LWR cells and fuel assembly lattices, and the main nodal kinetics parameters: mean neutron generation time and delayed neutron fraction; 2) the features of the advanced nodal methods for 3-D LWR core physics, including the treatment of partially inserted control rods, fuel assembly grids, fuel burnup and xenon and samarium transients, and excore detector responses, that are essential for core surveillance, axial offset control and operating transient analysis; 3) the advanced nodal methods for 3-D LWR core neutron kinetics (best estimate safety analysis, real-time simulation); and 4) example applications to 3-D neutron kinetics problems in transient analysis of PWR cores, including model, benchmark and operational transients without, or with simple, thermal-hydraulics feedback.

A calculational study on neutron kinetics and thermodynamics of a gaseous core fission reactor

International Nuclear Information System (INIS)

Kuijper, J.C.

1992-06-01

A numerical and analytical study of the static and dynamic properties of a GCFR with oscillating fuel gas (uranium and carbon fluorides) is presented. Neutron kinetics parts of combined GCFR models are introduced. Thermodynamic properties of the GCFR and of the fuel gas are treated. (HP)

Application of the exact distribution pj{sub k} in the determination of kinetic parameters in a reactor; Aplicacion de la distribucion exacta p{sub k} a la determinacion de parametros cineticos de un reactor

Energy Technology Data Exchange (ETDEWEB)

Alca Ruiz, F

1982-07-01

In this report one distribution of neutron counts obtained by a detector placed in a reactor is studied in order to be used in the determination of reactor kinetic parameters such as {beta}/{lambda} and reactivities. The parameters accuracy from this new method is compared with the Feynman and Mogilner method, based too in Reactor Neutron Noise Analysis. These three methods have been applied to JEN-2 reactor and the better accuracy and faster collection of experimental data give some interest to the new method which only requires a good footing code. (Author) 68 refs.

General remarks on fast neutron reactor physics

International Nuclear Information System (INIS)

Barre, J.Y.

1980-01-01

The main aspects of fast reactor physics, presented in these lecture notes, are restricted to LMFBR's. The emphasis is placed on the core neutronic balance and the burn-up problems. After a brief description of the power reactor main components and of the fast reactor chronology, the fundamental parameters of the one-group neutronic balance are briefly reviewed. Then the neutronic burn-up problems related to the Pu production and to the doubling time are considered

Effect of kinetic parameters on simultaneous ramp reactivity insertion plus beam tube flooding accident in a typical low enriched U{sub 3}Si{sub 2}-Al fuel-based material testing reactor-type research reactor

Energy Technology Data Exchange (ETDEWEB)

Nasir, Rubina; Mirza, Nasir M. [Dept. of, Physics, Air University, Islamabad (Pakistan); Mirza, Sikander M. [Dept. of, Physics and Applied Mathematics, Pakistan Institute of Engineering and Applied Sciences, Post Office Nilore, Islamabad (Pakistan)

2017-06-15

This work looks at the effect of changes in kinetic parameters on simultaneous reactivity insertions and beam tube flooding in a typical material testing reactor-type research reactor with low enriched high density (U{sub 3}Si{sub 2}-Al) fuel. Using a modified PARET code, various ramp reactivity insertions (from $0.1/0.5 s to $1.3/0.5 s) plus beam tube flooding ($0.5/0.25 s) accidents under uncontrolled conditions were analyzed to find their effects on peak power, net reactivity, and temperature. Then, the effects of changes in kinetic parameters including the Doppler coefficient, prompt neutron lifetime, and delayed neutron fractions on simultaneous reactivity insertion and beam tube flooding accidents were analyzed. Results show that the power peak values are significantly sensitive to the Doppler coefficient of the system in coupled accidents. The material testing reactor-type system under such a coupled accident is not very sensitive to changes in the prompt neutron life time; the core under such a coupled transient is not very sensitive to changes in the effective delayed neutron fraction.

Reactor physics and thermodynamics of a gaseous core fission reactor

International Nuclear Information System (INIS)

Kuijper, J.C.; Van Dam, H.; Stekelenburg, A.J.C.; Hoogenboom, J.E.; Boersma-Klein, W.; Kistemaker, J.

1990-01-01

Neutron kinetics and thermodynamics of a Gaseous Core Fission Reactor with magnetical pumping are shown to have many unconventional aspects. Attention is focussed on the properties of the fuel gas, the stationary temperature distribution, the non-linear neutron kinetics and the energy balance in thermodynamical cycles

Implementation of the neutron noise technique for subcritical reactors using a new data acquisition system

International Nuclear Information System (INIS)

Bellino, Pablo A.; Gomez, Angel

2009-01-01

A new data acquisition system was designed and programmed for nuclear kinetics parameter estimations in subcritical reactors. The system allows using any of the neutron noise techniques, since it could store the whole information available in the neutron detection system. The α Rossi, α Feynman and spectral analysis methods were performed in order to estimate the prompt neutron decay constant (and hence the reactivity). The measurements were done in the nuclear research reactor RA-1, where introducing the control rods, different reactivity levels where reached (until -7 dollars). With the three methods used, agreement was found between the estimations and the reference reactivities in each level, even when the detector efficiency was low. All the measurements were performed with a high gamma flux, although the results were found to be satisfactory. (author)

Neutron flux distribution forecasting device of reactor

International Nuclear Information System (INIS)

Uematsu, Hitoshi

1991-01-01

A neutron flux distribution is forecast by using current data obtained from a reactor. That is, the device of the present invention comprises (1) a neutron flux monitor disposed in various positions in the reactor, (2) a forecasting means for calculating and forecasting a one-dimensional neutron flux distribution relative to imaginable events by using data obtained from the neutron flux monitor and physical models, and (3) a display means for displaying the results forecast in the forecasting means to a reactor operation console. Since the forecast values for the one-dimensional neutron flux distribution relative to the imaginable events are calculated in the device of the present invention by using data obtained from the neutron flux monitor and the physical models, the data as a base of the calculation are new and the period for calculating the forecast values can be shortened. Accordingly, although there is a worry of providing some errors in the forecast values, they can be utilized sufficiently as reference data. As a result, the reactor can be operated more appropriately. (I.N.)

Theory of stochastic space-dependent neutron kinetics with a Gaussian parametric excitation

International Nuclear Information System (INIS)

Saito, K.

1980-01-01

Neutron kinetics and statics in a multiplying medium with a statistically fluctuating reactivity are unified and systematically studied by applying the Novikov-Furutsu formula. The parametric or multiplicative noise is spatially distributed and of Gaussian nature with an arbitrary spectral profile. It is found that the noise introduces a new definite production term into the conventional balance equation for the mean neutron number. The term is characterized by the magnitude and the correlation function of the random excitation. Its relaxation phenomena bring forth a non-Markoffian or a memory effect, which is conceptualised by introducing 'pseudo-precursors' or 'pseudo-delayed neutrons'. By using the concept, some typical reactor physical problems are solved; they are (1) reactivity and flux perturbation originating from the random dispersal of core materials and (2) analysis of neutron decay mode and it relaxation constant, and derivation of the corresponding new inhour equation. (author)

Calculation of neutron spectra in the reactor cell of the RA experimental reactor in Vinca

International Nuclear Information System (INIS)

Bosevski, T.; Altiparmakov, D.; Marinkovic, N.

1974-01-01

In the frame of neutron properties of RA experimental reactor the study of energy neutron spectra in the reactor cell are planned. Complex reactor cell geometry, nine cylindrical regions causes high space-energy variations of neutron flux with a significant gradient both in energy and space variables. Treatment of such a complex problem needs adequate methodology which ensures reliable results and control of accuracy. This paper describes in detail the method for calculating group constants based on lattice cell calculation for the need of calculation of reactor core parameters. In 26 group approximation for the energy region from 0 - 10.5 MeV, values of neutron spectra are obtained in 18 space points chosen to describe, with high accuracy, integral reactor cell parameters of primary importance for the reactor core calculation. Obtained space-energy distribution of neutron flux in the reactor cell is up to now unique in the study of neutron properties of Ra reactor [sr

Reactor cold neutron source facility, the first in Japan

International Nuclear Information System (INIS)

Utsuro, Masahiko; Maeda, Yutaka; Kawai, Takeshi; Tashiro, Tameyoshi; Sakakibara, Shoji; Katada, Minoru.

1986-01-01

In the Research Reactor Institute, Kyoto University, the first cold neutron source facility for the reactor in Japan was installed, and various tests are carried out outside the reactor. Nippon Sanso K.K. had manufactured it. After the prescribed tests outside the reactor, this facility will be installed soon in the reactor, and its outline is described on this occasion. Cold neutrons are those having very small energy by being cooled to about-250 deg C. Since the wavelength of the material waves of cold neutrons is long, and their energy is small, they are very advantageous as an experimental means for clarifying the structure of living body molecules and polymers, the atom configuration in alloys, and atomic and molecular movements by neutron scattering and neutron diffraction. The basic principle of the cold neutron source facility is to irradiate thermal neutrons on a cold moderator kept around 20 K, and to moderate and cool the neutrons by nuclear scattering to convert to cold neutrons. The preparatory research on cold neutrons and hydrogen liquefaction, the basic design to put the cold neutron source facility in the graphite moderator facility, the safety countermeasures, the manufacture and quality control, the operation outside the reactor and the performance are reported. The cold neutron source facility comprises a cold moderator tank and other main parts, a deuterium gas tank, a helium refrigerator and instrumentation. (Kako, I.)

Utilizing horizontal reactors channels for neutron therapy

International Nuclear Information System (INIS)

Stankovsky, E.Yu.; Kurachenko, Yu.A.

2000-01-01

Two experimental heterogeneous reactors have been considered. The reactors may be applied in neutron capture therapy and in a conventional manner. The channel out of the core serves as the neutron source. At each of these facilities, both fast and epithermal neutron fluxes for BNCT research, human clinical trials, and characterized common computational techniques have been evaluated. (authors)

Relation between nonlinear or 'not-linear' characteristics in nuclear kinetics and noise analysis of neutron flux

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)

Measures of the zero power nuclear reactor's kinetic parameters with application of noise analysis

International Nuclear Information System (INIS)

Martins, F.R.

1992-01-01

The purpose of this work was to establish an experimental technique based on noise analysis for measuring the ratio of kinetic parameters β/ Λ and the power of the Zero Power Nuclear Reactor IPEN-MB 01. A through study of the microscopic and macroscopic noise analysis techniques has been carried out. The Langevin technique and the point kinetic model were chosen to describe the stochastic phenomena that occur in the zero power reactor. Measurements have been made using two compensated ionization chambers localized in the water reflector at symmetric positions in order to minimize spatial effects on the neutron flux fluctuation. Power calibrations based on the low frequency plateau of the cross-power spectral density has also been carried out. (author)

Preliminary Development of the MARS/FREK Spatial Kinetics Coupled System Code for Square Fueled Fast Reactor Applications

International Nuclear Information System (INIS)

Bae, Moo Hoon; Joo, Han Gyu

2009-01-01

Incorporation of a three-dimensional (3-D) reactor kinetics model into a system thermal-hydraulic (T/H) code enhances the capability to perform realistic analyses of the core neutronic behavior and the plant system dynamics which are coupled each other. For this advantage, several coupled system T/H and spatial kinetics codes, such as RELAP/PARCS, RELAP5/ PANBOX, and MARS/MASTER have been developed. These codes, however, so far limited to LWR applications. The objective of this work is to develop such a coupled code for fast reactor applications. Particularly, applications to lead-bismuth eutectic (LBE) cooled fast reactor are of interest which employ open square lattices. A fast reactor kinetics code applicable to square fueled cores called FREK is coupled the LBE version of the MARS code. The MARS/MASTER coupled code is used as the reference for the integration. The coupled code MARS/FREK is examined for a conceptual reactor called P-DEMO which is being developed by NUTRECK. In order to check the validity of the coupled code, however, the OECD MSLB benchmark exercise III calculation is solved first

Neutronic reactor

International Nuclear Information System (INIS)

Wende, C.W.J.

1976-01-01

The method of operating a water-cooled neutronic reactor having a graphite moderator is described which comprises flowing a gaseous mixture of carbon dioxide and helium, in which the helium comprises 40--60 volume percent of the mixture, in contact with the graphite moderator. 2 claims, 4 figures

Seminar on Heat-transfer fluids for fast neutron reactors

International Nuclear Information System (INIS)

Brechet, Yves; Dautray, Robert; Friedel, Jacques; Brezin, Edouard; Martin, Georges; Pineau, Andre; Carre, Francois; Gauche, Francois; Rodriguez, Guillaume; Latge, Christian; Cabet, Celine; Garnier, Jean-Claude; Bamberger, Yves; Sauvage, Jean-Francois; Buisine, Denis; Agostini, Pietro; Ulyanov, Vladimir; Auger, Thierry; Heuer, Daniel; Ghetta, Veronique; Bubelis, Evaldas; Charlaix, Elisabeth; Barrat, Jean-Louis; Boquet, Lyderic; Glickman, Evgueny; Escaravage, Claude

2014-03-01

This book reports the content of a two-day meeting held by the Academy of Sciences on the use of heat-transfer fluids in fast neutron reactors. After a first part which proposes an overview of scientific and technical problems related to these heat-transfer fluids (heat transfer process, nuclear properties, chemistry, materials, risks), a contribution proposes a return on experience on the use of heat-transfer fluids in the different design options of reactors of fourth generation: from mercury to NaK in the first fast neutron reactor projects, specific assets and constraints of sodium used as heat-transfer fluid, concepts of fast neutron reactors cooled by something else than sodium, perspectives for projects and research in fast neutron reactors. The next contribution discusses the specifications of future fast-neutron reactors: expectations for fourth-generation reactors, expectations in terms of performance and of safety, specific challenges. The last contribution addresses actions to be undertaken in the field of research and development: actions regarding all reactor types or specific types as sodium-cooled reactors, lead cooled reactors, molten salt reactors, and gas-cooled fast reactors

Investigating The Neutron Flux Distribution Of The Miniature Neutron Source Reactor MNSR Type

International Nuclear Information System (INIS)

Nguyen Hoang Hai; Do Quang Binh

2011-01-01

Neutron flux distribution is the important characteristic of nuclear reactor. In this article, four energy group neutron flux distributions of the miniature neutron source reactor MNSR type versus radial and axial directions are investigated in case the control rod is fully withdrawn. In addition, the effect of control rod positions on the thermal neutron flux distribution is also studied. The group constants for all reactor components are generated by the WIMSD code, and the neutron flux distributions are calculated by the CITATION code. The results show that the control rod positions only affect in the planning area for distribution in the region around the control rod. (author)

Neutron beam facilities at the replacement research reactor

International Nuclear Information System (INIS)

Kennedy, S.

1999-01-01

Full text: On September 3rd 1997 the Australian Federal Government announced their decision to replace the HIFAR research reactor by 2005. The proposed reactor will be a multipurpose reactor with improved capabilities for neutron beam research and for the production of radioisotopes for pharmaceutical, scientific and industrial use. The neutron beam facilities are intended to cater for Australian scientific needs well into the 21st century. In the first stage of planning the neutron Beam Facilities at the replacement reactor, a Consultative Group was formed (BFCG) to determine the scientific capabilities of the new facility. Members of the group were drawn from academia, industry and government research laboratories. The BFCG submitted their report in April 1998, outlining the scientific priorities to be addressed. Cold and hot neutron sources are to be included, and cold and thermal neutron guides will be used to position most of the instruments in a neutron guide hall outside the reactor confinement building. In 2005 it is planned to have eight instruments installed with a further three to be developed by 2010, and seven spare instrument positions for development of new instruments over the life of the reactor. A beam facilities technical group (BFTG) was then formed to prepare the engineering specifications for the tendering process. The group consisted of some members of the BFCG, several scientists and engineers from ANSTO, and scientists from leading neutron scattering centres in Europe, USA and Japan. The BFTG looked in detail at the key components of the facility such as the thermal, cold and hot neutron sources, neutron collimators, neutron beam guides and overall requirements for the neutron guide hall. The report of the BFTG, completed in August 1998, was incorporated into the draft specifications for the reactor project, which were distributed to potential reactor vendors. An assessment of the first stage of reactor vendor submissions was completed in

CINESP - computational program of spatial kinetics for nuclear reactors in the one-two dimension multigroup diffusion theory

International Nuclear Information System (INIS)

Santos, R.S. dos

1993-01-01

This paper presents a computational program to solve numerically the reactor kinetics equations in the multigroup diffusion theory. One or two-dimensional problems in cylindrical or Cartesian geometries, with any number of energy and delayed-neutron precursors groups are dealt with. The main input and output of the program are briefly discussed. Various results demonstrate the accuracy and versatility of the program, when compared with other kinetics programs. (author)

EL-2 reactor: Thermal neutron flux distribution

International Nuclear Information System (INIS)

Rousseau, A.; Genthon, J.P.

1958-01-01

The flux distribution of thermal neutrons in EL-2 reactor is studied. The reactor core and lattices are described as well as the experimental reactor facilities, in particular, the experimental channels and special facilities. The measurement shows that the thermal neutron flux increases in the central channel when enriched uranium is used in place of natural uranium. However the thermal neutron flux is not perturbed in the other reactor channels by the fuel modification. The macroscopic flux distribution is measured according the radial positioning of fuel rods. The longitudinal neutron flux distribution in a fuel rod is also measured and shows no difference between enriched and natural uranium fuel rods. In addition, measurements of the flux distribution have been effectuated for rods containing other material as steel or aluminium. The neutron flux distribution is also studied in all the experimental channels as well as in the thermal column. The determination of the distribution of the thermal neutron flux in all experimental facilities, the thermal column and the fuel channels has been made with a heavy water level of 1825 mm and is given for an operating power of 1000 kW. (M.P.)

On the research activities in reactor and neutron physics using the first egyptian research reactor

International Nuclear Information System (INIS)

Hassan, A.M.

2000-01-01

A review on the most important research activities in reactor and neutron physics using the first Egyptian Research Reactor (ET-RR-1) is given. An out look on: neutron cross-sections, neutron flux, neutron capture gamma-ray spectroscopy, neutron activation analysis, neutron diffraction and radiation shielding experiments, is presented

Neutronics analysis of Dalat Research Reactor

International Nuclear Information System (INIS)

Pham Van Lam; Luong Ba Vien; Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Kien Cuong; Nguyen Manh Hung; Pham Hong Son; Tran Quoc Duong

2006-01-01

Many neutronics codes have been used to calculate for Dalat Research Reactor (DRR) from 1983 (the first critical of DRR in December, 1983). The purposes of all calculations are to know exactly many important parameters related to Reactor Physics and Neutron Physics in reactor core. The results from calculation play important role in core and fuel management for DRR. Especially basing on the results we can predict about fuel cycle, fuel burn up distribution and plan for using optimize remain fresh fuel assemblies of DRR. By using system neutronics code including transport codes, diffusion codes and Mote Carlo code, many characteristics of fuel assemblies and other parameters of whole core were received such as main features of VVR-M2 fuel assembly type, multiplication factor, neutron flux distribution, power distribution, burn up distribution, excess reactivity, control rods worth, neutron spectrum, temperature reactivity coefficient ect. In the paper, brief description all computer codes to being used in DRR and the calculation results from the codes above are presented. (author)

Theory of neutron slowing down in nuclear reactors

CERN Document Server

Ferziger, Joel H; Dunworth, J V

2013-01-01

The Theory of Neutron Slowing Down in Nuclear Reactors focuses on one facet of nuclear reactor design: the slowing down (or moderation) of neutrons from the high energies with which they are born in fission to the energies at which they are ultimately absorbed. In conjunction with the study of neutron moderation, calculations of reactor criticality are presented. A mathematical description of the slowing-down process is given, with particular emphasis on the problems encountered in the design of thermal reactors. This volume is comprised of four chapters and begins by considering the problems

Numerical nodal simulation of the axial power distribution within nuclear reactors using a kinetics diffusion model. I

International Nuclear Information System (INIS)

Barros, R.C. de.

1992-05-01

Presented here is a new numerical nodal method for the simulation of the axial power distribution within nuclear reactors using the one-dimensional one speed kinetics diffusion model with one group of delayed neutron precursors. Our method is based on a spectral analysis of the nodal kinetics equations. These equations are obtained by integrating the original kinetics equations separately over a time step and over a spatial node, and then considering flat approximations for the forward difference terms. These flat approximations are the only approximations that are considered in the method. As a result, the spectral nodal method for space - time reactor kinetics generates numerical solutions for space independent problems or for time independent problems that are completely free from truncation errors. We show numerical results to illustrate the method's accuracy for coarse mesh calculations. (author)

Achievement and development of neutron beam utilization in research reactors

International Nuclear Information System (INIS)

Isshiki, Masahiko

1996-01-01

Especially regarding the neutron beam experiment in Japan, the basic research has been developed by utilizing the JRR-2 of Japan Atomic Energy Research Institute and the KUR of Kyoto University over long years. Now, the JRR-3M of JAERI was revived as a high performance, general purpose reactor, and bears important roles as the neutron beam experiment center in Japan. Thanks to one of the most powerful reactor neutron sources in the world and the cold neutron source, the environment of research was greatly improved, and the excellent results of researches began to be reported. The discovery of neutrons by Chadwick and the history of the related researches are described. As neutron sources, radioisotopes, accelerators and nuclear reactors are properly used corresponding to purposes. As the utilization of research reactors for neutron sources, the utilization for irradiation and neutron beam experiment are carried out. The outline of the research reactor JRR-3M is explained. The state of utilization in neutron scattering experiment, neutron radiography, prompt γ-ray analysis and the medical irradiation of neutrons is reported. (K.I.)

Neutron importance and the generalized Green function for the conventionally critical reactor with normalized neutron distribution

International Nuclear Information System (INIS)

Khromov, V.V.

1978-01-01

The notion of neutron importance when applied to nuclear reactor statics problems described by time-independent homogeneous equations of neutron transport with provision for normalization of neutron distribution is considered. An equation has been obtained for the function of neutron importance in a conditionally critical reactor with respect to an arbitrary nons linear functional determined for the normalized neutron distribution. Relation between this function and the generalized Green function of the selfconjugated operator of the reactor equation is determined and the formula of small perturbations for the functionals of a conditionally critical reactor is deduced

Neutron converter at reactor RB; Konvertor neutrona na reaktoru RB

Energy Technology Data Exchange (ETDEWEB)

Strugar, P; Sotic, P; Ninkovic, M; Pesic, M [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

1977-07-01

A neutron converter at Reactor RB in the 'Boris Kidric' Institute of Nuclear Sciences - Vinca has been constructed. Preliminary measurements have been shown that the converted neutron spectrum is very similar to the fission neutron spectrum. For the same integral reactor power, the measured neutron radiation dose has been for about ten times larger with the neutron converter. The neutron converter offers wide possibilities, as in investigations in the reactor physics, where the fission neutron spectra have been required, as well as in the field of neutron dosimetry and biological irradiations (author)

Experimental neutronic science and instrumentation: from hybrid reactors to fourth generation reactors

International Nuclear Information System (INIS)

Jammes, Ch.

2010-07-01

After an overview of his academic career and scientific and research activities, the author proposes a rather detailed synthesis and overview of his scientific activities in the fields of cross sections and Doppler effect (development and validation of a code), on the MUSE-4 hybrid reactor (experiments, static and dynamic measurements), on the TRADE hybrid reactor (experimental means, sub-critical reactivity measurement), on the RACE hybrid reactor (experimental results, modelling and interpretation), and on neutron detection (design and modelling of fission chamber, on-line measurement of the fast flow). The next part gives an overview of some research programs (neutron monitoring in sodium-cool fast reactors, research and development on fission chambers, improvement of effective delayed neutron measurements)

Neuro-diffuse algorithm for neutronic power identification of TRIGA Mark III reactor

International Nuclear Information System (INIS)

Rojas R, E.; Benitez R, J. S.; Segovia de los Rios, J. A.; Rivero G, T.

2009-10-01

In this work are presented the results of design and implementation of an algorithm based on diffuse logic systems and neural networks like method of neutronic power identification of TRIGA Mark III reactor. This algorithm uses the punctual kinetics equation as data generator of training, a cost function and a learning stage based on the descending gradient algorithm allow to optimize the parameters of membership functions of a diffuse system. Also, a series of criteria like part of the initial conditions of training algorithm are established. These criteria according to the carried out simulations show a quick convergence of neutronic power estimated from the first iterations. (Author)

Neutron flux measuring system for nuclear reactor

International Nuclear Information System (INIS)

Aoki, Kazuo.

1977-01-01

Purpose: To avoid the generation of an undesired scram signal due to abrupt changes in the neutron level given to the detectors disposed near the boundary between the moderator and the atmosphere. Constitution: In a nuclear reactor adapted to conduct power control by the change of the level in the moderator such as heavy water, the outputs from the neutron detectors disposed vertically are averaged and the nuclear reactor is scramed corresponding to the averaged value. In this system, moderator level detectors are additionally provided to the nuclear reactor and their outputs, moderator level signal, are sent to a power averaging device where the output signals of the neutron detectors are judged if they are delivered from neutrons in the moderator or not depending on the magnitude of the level signal and the outputs of the detectors out of the moderator are substantially excluded. The reactor interlock signal from the device is utilized as a scram signal. (Seki, T.)

Neutronic study of the two french heavy water reactors

International Nuclear Information System (INIS)

Horowitz, J.

1955-01-01

The two french reactors - the reactor of Chatillon, named Zoe, and the reactor of Saclay - P2 - were the object of detailed neutronic studies which the main ideas are exposed in this report. These studies were mostly done by the Department of the Reactor Studies (D.E.P.). We have thus studied the distribution of neutronic fluxes; the factors influencing reactivity; the link between reactivity and divergence with the formula of Nordheim; the mean time life of neutrons; neutron spectra s of P2; the xenon effect; or the effect of the different adjustments of the plates and controls bar. (M.B.) [fr

AUS - the Australian modular scheme for reactor neutronics computations

International Nuclear Information System (INIS)

Robinson, G.S.

1975-12-01

A general description is given of the AUS modular scheme for reactor neutronics calculations. The scheme currently includes modules which provide the capacity for lattice calculations, 1D transport calculations, 1 and 2D diffusion calculations (with feedback-free kinetics), and burnup calculations. Details are provided of all system aspects of AUS, but individual modules are only outlined. A complete specification is given of that part of user input which controls the calculation sequence. The report also provides sufficient details of the supervisor program and of the interface data sets to enable additional modules to be incorporated in the scheme. (author)

The reflector effect on the neutron lifetimes in the IPEN/MB-01 reactor; O efeito do refletor sobre o tempo de vida neutronico no reator IPEN/MB-01

Energy Technology Data Exchange (ETDEWEB)

Gonnelli, Eduardo

2013-07-01

The aim of this study is to present the reflector effect on the neutron lifetimes in the IPEN/MB-01 reactor. The proposed method requires an approach which takes into account both the reflector and the core, so that the point kinetics equations, which constitute the theoretical basis of all mathematical development, contemplate both regions of the reactor. From these equations, as known as two regions kinetics point equations, theoretical expressions are obtained for the Auto Power Spectral Densities (APSD), which are used for least squares fit of the experimental data of APSD obtained in several subcritical states. The prompt neutron generation time, the neutron lifetimes in the reflector and the neutron return fraction from the reflector to the core are derived from the fitting. (author)

The neutron beam facility at the Australian replacement research reactor

International Nuclear Information System (INIS)

Hunter, B.; Kennedy, S.

1999-01-01

Full text: The Australian federal government gave ANSTO final approval to build a research reactor to replace HIFAR on August 25th 1999. The replacement reactor is to be a multipurpose reactor with a thermal neutron flux of 3 x 10 14 n.cm -2 .s -1 and having improved capabilities for neutron beam research and for the production of radioisotopes for pharmaceutical, scientific and industrial use. The replacement reactor will commence operation in 2005 and will cater for Australian scientific, industrial and medical needs well into the 21st century. The scientific capabilities of the neutron beams at the replacement reactor are being developed in consultation with representatives from academia, industry and government research laboratories to provide a facility for condensed matter research in physics, chemistry, materials science, life sciences, engineering and earth sciences. Cold, thermal and hot neutron sources are to be installed, and neutron guides will be used to position most of the neutron beam instruments in a neutron guide hall outside the reactor confinement building. Eight instruments are planned for 2005, with a further three to be developed by 2010. A conceptual layout for the neutron beam facility is presented including the location of the planned suite of neutron beam instruments. The reactor and all the associated infrastructure, with the exception of the neutron beam instruments, is to be built by an accredited reactor builder in a turnkey contract. Tenders have been called for December 1999, with selection of contractor planned by June 2000. The neutron beam instruments will be developed by ANSTO and other contracted organisations in consultation with the user community and interested overseas scientists. The facility will be based, as far as possible, around a neutron guide hall that is be served by three thermal and three cold neutron guides. Efficient transportation of thermal and cold neutrons to the guide hall requires the use of modern super

Polarization of fast neutrons in VVR-M reactor

International Nuclear Information System (INIS)

Garusov, E.A.; Lifshits, E.P.; Petrov, Yu.V.

1987-01-01

Neutron polarization in the reactor leads to circular polarization of γ quanta emitted both in radiational capture of neutrons and in the transition of nuclei excited as a result of inelastic scattering to the ground state. This may be used to determine the polarization of reactor neutrons. The circular polarization of γ quanta at light-water and graphite targets at the center of the active zone of the VVR-M reactor at the B.P. Konstantinov Leningrad Institute of Nuclear Physics was recently measured. A simplified experimental scheme is shown. Fast neutrons leaving the active zone of the reactor were excited in the inelastic scattering at the target nuclei. The polarization of the γ quanta emitted by nuclei in transitions to the ground state was measured by a polarimeter positioned above the active zone. The reason for the circular polarization of γ quanta may also be nonconservation of P parity on account of weak interaction in the capture of a neutron by hydrogen

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

Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

Energy Technology Data Exchange (ETDEWEB)

Mohammed, Abdul Aziz, E-mail: azizM@uniten.edu.my; Rahman, Shaik Mohmmed Haikhal Abdul [Universiti Tenaga Nasional. Jalan Ikram-UNITEN, 43000 Kajang, Selangor (Malaysia); Pauzi, Anas Muhamad, E-mail: anas@uniten.edu.my; Zin, Muhamad Rawi Muhammad; Jamro, Rafhayudi; Idris, Faridah Mohamad [Malaysian Nuclear Agency, Bangi, 43000 Kajang, Selangor (Malaysia)

2016-01-22

In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 ({sup 233}U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.

NEUTRONIC REACTOR STRUCTURE

Science.gov (United States)

Weinberg, A.M.; Vernon, H.C.

1961-05-30

A neutronic reactor is described. It has a core consisting of natural uranium and heavy water and having a K-factor greater than unity which is surrounded by a reflector consisting of natural uranium and ordinary water having a Kfactor less than unity.

Performance Test for Neutron Detector and Associated System using Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Yang, Seongwoo; Park, Sung Jae; Cho, Man Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Oh, Se Hyun [USERS, Daejeon (Korea, Republic of); Shin, Ho Cheol [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

SPND (Self-Powered Neutron Detector) has been developed to extend its lifespan. ENFMS (Ex-Core Flux Monitoring System) of pressurized water reactor has been also improved. After the development and improvement, their performance must be verified under the neutron irradiation environment. We used a research reactor for the performance verification of neutron detector and associated system because the research reactor can meet the neutron flux level of commercial nuclear reactor. In this paper, we report the performance verification method and result for the SPND and ENFMS using the research reactor. The performance tests for the SPND and ENFMS were conducted using UCI TRIGA reactor. The test environment of commercial reactor’s neutron flux level must be required. However, it is difficult to perform the test in the commercial rector due to the constraint of time and space. The research reactor can be good alternative neutron source for the test of neutron detectors and associated system.

Sulfide toxicity kinetics of a uasb reactor

Directory of Open Access Journals (Sweden)

D. R. Paula Jr.

2009-12-01

Full Text Available The effect of sulfide toxicity on kinetic parameters of anaerobic organic matter removal in a UASB (up-flow anaerobic sludge blanket reactor is presented. Two lab-scale UASB reactors (10.5 L were operated continuously during 12 months. The reactors were fed with synthetic wastes prepared daily using glucose, ammonium acetate, methanol and nutrient solution. One of the reactors also received increasing concentrations of sodium sulfide. For both reactors, the flow rate of 16 L.d-1 was held constant throughout the experiment, corresponding to a hydraulic retention time of 15.6 hours. The classic model for non-competitive sulfide inhibition was applied to the experimental data for determining the overall kinetic parameter of specific substrate utilization (q and the sulfide inhibition coefficient (Ki. The application of the kinetic parameters determined allows prediction of methanogenesis inhibition and thus the adoption of operating parameters to minimize sulfide toxicity in UASB reactors.

Neutron beam facilities at the Replacement Research Reactor, ANSTO

International Nuclear Information System (INIS)

Kim, S.

2003-01-01

The exciting development for Australia is the construction of a modern state-of-the-art 20-MW Replacement Research Reactor which is currently under construction to replace the aging reactor (HIFAR) at ANSTO in 2006. To cater for advanced scientific applications, the replacement reactor will provide not only thermal neutron beams but also a modern cold-neutron source moderated by liquid deuterium at approximately -250 deg C, complete with provision for installation of a hot-neutron source at a later stage. The latest 'supermirror' guides will be used to transport the neutrons to the Reactor Hall and its adjoining Neutron Guide Hall where a suite of neutron beam instruments will be installed. These new facilities will expand and enhance ANSTO's capabilities and performance in neutron beam science compared with what is possible with the existing HIFAR facilities, and will make ANSTO/Australia competitive with the best neutron facilities in the world. Eight 'leading-edge' neutron beam instruments are planned for the Replacement Research Reactor when it goes critical in 2006, followed by more instruments by 2010 and beyond. Up to 18 neutron beam instruments can be accommodated at the Replacement Research Reactor, however, it has the capacity for further expansion, including potential for a second Neutron Guide Hall. The first batch of eight instruments has been carefully selected in conjunction with a user group representing various scientific interests in Australia. A team of scientists, engineers, drafting officers and technicians has been assembled to carry out the Neutron Beam Instrument Project to successful completion. Today, most of the planned instruments have conceptual designs and are now being engineered in detail prior to construction and procurement. A suite of ancillary equipment will also be provided to enable scientific experiments at different temperatures, pressures and magnetic fields. This paper describes the Neutron Beam Instrument Project and gives

A conceptual design of neutron tumor therapy reactor facility with a YAYOI based fast neutron source reactor

International Nuclear Information System (INIS)

Wakabayashi, Hiroaki; An, Shigehiro.

1983-01-01

Fast neutron is known as one of useful radiations for radiation therapy of tumors. Boron neutron capture therapy (BNCT) of tumors which makes use of 10 B(n, α) 7 Li reaction of 10 B compounds selectively attached to tumor cells with thermal and intermediate neutrons is another way of neutron based radiation therapy which is, above all, attractive enough to kill tumor cells selectively sparing normal tissue. In Japan, BNCT has already been applied and leaned to be effective. After more than a decade operational experiences and the specific experiments designed for therapeutical purposes, in this paper, a conceptual design of a special neutron therapy reactor facility based on YAYOI - fast neutron source reactor of Nuclear Engineering Research Laboratory, Faculty of Engineering, the University of Tokyo - modified to provide an upward beam of fast and intermediate neutrons is presented. Emphasis is placed on the in-house nature of facility and on the coordinating capability of biological and physical researches as well as maintenances of the facility. (author)

Status of neutron beam utilization at the Dalat nuclear research reactor

International Nuclear Information System (INIS)

Dien, Nguyen Nhi; Hai, Nguyen Canh

2003-01-01

The 500-kW Dalat nuclear research reactor was reconstructed from the USA-made 250-kW TRIGA Mark II reactor. After completion of renovation and upgrading, the reactor has been operating at its nominal power since 1984. The reactor is used mainly for radioisotope production, neutron activation analysis, neutron beam researches and reactor physics study. In the framework of the reconstruction and renovation project of the 1982-1984 period, the reactor core, the control and instrumentation system, the primary and secondary cooling systems, as well as other associated systems were newly designed and installed by the former Soviet Union. Some structures of the reactor, such as the reactor aluminum tank, the graphite reflector, the thermal column, horizontal beam tubes and the radiation concrete shielding have been remained from the previous TRIGA reactor. As a typical configuration of the TRIGA reactor, there are four neutron beam ports, including three radial and one tangential. Besides, there is a large thermal column. Until now only two-neutron beam ports and the thermal column have been utilized. Effective utilization of horizontal experimental channels is one of the important research objectives at the Dalat reactor. The research program on effective utilization of these experimental channels was conducted from 1984. For this purpose, investigations on physical characteristics of the reactor, neutron spectra and fluxes at these channels, safety conditions in their exploitation, etc. have been carried out. The neutron beams, however, have been used only since 1988. The filtered thermal neutron beams at the tangential channel have been extracted using a single crystal silicon filter and mainly used for prompt gamma neutron activation analysis (PGNAA), neutron radiography (NR) and transmission experiments (TE). The filtered quasi-monoenergetic keV neutron beams using neutron filters at the piercing channel have been used for nuclear data measurements, study on

Commissioning of the Opal reactor cold neutron source

International Nuclear Information System (INIS)

Thiering, R.; Lu, W.; Ullah, R.

2006-01-01

Full text: At OPAL, Australia's first cold neutron facility will form an essential part of the reactor's research programs. Fast neutrons, born in the core of a reactor, interact with a cryogenic material, in this case liquid deuterium, to give them very low energies ( 1 0 m eV). A cold neutron flux of 1.4 1 0 E 1 4 n /cm 2/ s is expected, with a peak in the energy spectrum at 4.2m eV. The cold neutron source reached cryogenic conditions for the first time in late 2005. The cold neutron source operates with a sub-cooled liquid Deuterium moderator at 24 K. The moderator chamber, which contains the deuterium, has been constructed from AlMg 5. The thermosiphon and moderator chamber are cooled by helium gas, in a natural convection thermosiphon loop. The helium refrigeration system utilises the Brayton cycle, and is fully insulated within a high vacuum environment. Despite the proximity of the cold neutron source to the reactor core, it has been considered as effectively separate to the reactor system, due to the design of its special vacuum containment vessel. As OPAL is a multipurpose research reactor, used for beam research as well as radiopharmaceutical production and industrial irradiations, the cold neutron source has been designed with a stand-by mode, to maximise production. The stand-by mode is a warm operating mode using only gaseous deuterium at ambient temperatures (∼ 3 00 K ), allowing for continued reactor operations whilst parts of the cold source are unavailable or in maintenance. This is the first time such a stand-by feature has been incorporated into a cold source facility

Measurement of delayed neutron-emitting fission products in nuclear reactor coolant water during reactor operation

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

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

Kartini Research Reactor prospective studies for neutron scattering application

International Nuclear Information System (INIS)

Widarto

1999-01-01

The Kartini Research Reactor (KRR) is located in Yogyakarta Nuclear Research Center, Yogyakarta - Indonesia. The reactor is operated for 100 kW thermal power used for research, experiments and training of nuclear technology. There are 4 beam ports and 1 column thermal are available at the reactor. Those beam ports have thermal neutron flux around 10 7 n/cm 2 s each other and used for sub critical assembly, neutron radiography studies and Neutron Activation Analysis (NAA). Design of neutron collimator has been done for piercing radial beam port and the calculation result of collimated neutron flux is around 10 9 n/cm 2 s. This paper describes experiment facilities and parameters of the Kartini research reactor, and further more the prospective studies for neutron scattering application. The purpose of this paper is to optimize in utilization of the beam ports facilities and enhance the manpower specialty. The special characteristic of the beam ports and preliminary studies, pre activities regarding with neutron scattering studies for KKR is presented. (author)

Neutron beam utilization at the TRIGA Mark II reactor Vienna

International Nuclear Information System (INIS)

Villa, M.; Boeck, H.; Ismail, S.; Koerner, S.; Baron, M.; Hainbuchner, M.; Badurek, G.; Buchelt, R.J.

1999-01-01

A review is given about the research activities around the 250 kw TRIGA reactor Vienna, which are adequate to other neutron sources of comparable or bigger size. The topics selected for presentation range from neutron radiography, materials irradiation, neutron small-angle scattering, neutron activation analysis, neutron polarization to neutron interferometry. It is the aim of this presentation to stimulate programs for more efficient use around TRIGA research reactors with neutron flux densities of 1013 cm-2a-1 at the center of the reactor core. We briefly describe the experimental facilities installed at the 250 kw TRIGA reactor of the Austrian Universities in Vienna and present a great part of the current research activities performed with them. We believe that most of the techniques and experiments presented here are adequate for implementation to other reactors of similar or even higher power. Those technologies which require extremely specialized know-how not generally available at every research Inst.e will not be treated here or are just mentioned without any further details.(author)

Transient cases analyses of the TRIGA IPR-R1 using thermal hydraulic and neutron kinetic coupled codes

Energy Technology Data Exchange (ETDEWEB)

Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Scari, Maria E., E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br, E-mail: melizabethscari@yahoo.com [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq), Belo Horizonte (Brazil); Miro, Rafael; Verdu, Gumersindo, E-mail: rmiro@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Universidad Politecnica de Valencia (Spain). Departamento de Ingenieria Quimica y Nuclear

2015-07-01

Simulations and analyses of nuclear reactors have been improved by utilization of coupled thermal-hydraulic (TH) and neutron kinetics (NK) system codes especially to simulate transients that involve strong feedback effects between NK and TH. The TH-NK coupling technique was initially developed and used to simulate the behavior of power reactors; however, several coupling methodologies are now being applied for research reactors. This work presents the coupling methodology application between RELAP5 and PARCS codes using as a model the TRIGA IPR-R1 research reactor. Analyses of steady state and transient conditions and comparisons with results from simulations using only the RELAP5 code are being presented in this paper. (author)

Maximum neutron flux in thermal reactors

International Nuclear Information System (INIS)

Strugar, P.V.

1968-12-01

Direct approach to the problem is to calculate spatial distribution of fuel concentration if the reactor core directly using the condition of maximum neutron flux and comply with thermal limitations. This paper proved that the problem can be solved by applying the variational calculus, i.e. by using the maximum principle of Pontryagin. Mathematical model of reactor core is based on the two-group neutron diffusion theory with some simplifications which make it appropriate from maximum principle point of view. Here applied theory of maximum principle are suitable for application. The solution of optimum distribution of fuel concentration in the reactor core is obtained in explicit analytical form. The reactor critical dimensions are roots of a system of nonlinear equations and verification of optimum conditions can be done only for specific examples

OBJECT KINETIC MONTE CARLO SIMULATIONS OF RADIATION DAMAGE IN TUNGSTEN SUBJECTED TO NEUTRON FLUX WITH PKA SPECTRUM CORRESPONDING TO THE HFIR

Energy Technology Data Exchange (ETDEWEB)

Nandipati, Giridhar; Setyawan, Wahyu; Heinisch, Howard L.; Roche, Kenneth J.; Kurtz, Richard J.; Wirth, Brian D.

2015-12-31

The objective of this work is to study the damage accumulation in pure tungsten (W) subjected to neutron bombardment with a primary knock-on atom (PKA) spectrum corresponding to the High Flux Isotope Reactor (HFIR), using the object kinetic Monte Carlo (OKMC) method.

Safety reviews of the Brazilian multipurpose reactor

International Nuclear Information System (INIS)

Soares, Humberto Vitor

2014-01-01

This work presents a model developed for thermal hydraulic (TH) simulation of the Multipurpose Brazilian Reactor (RMB), whose Brazilian proposal for design, construction and operation was established in 2007. This reactor has as main proposed the production of radioisotopes for use in exams of nuclear medicine, material tests and utilization of neutrons beams. Besides of the TH modeling and safety analysis of the reactor, the application of a methodology to perform coupled calculation thermal-hydraulic/neutron kinetic (TH/NK) is also presented. Initially, the RMB was modeled in the safety analysis RELAP5 code. This code performs the thermal hydraulic calculation using point kinetics. Subsequently, the model was adapted and verified to the RELAP5-3D© code. This code performs the process of internal coupling through the option of nodal neutron kinetics calculation using the NESTLE code which solves the neutron diffusion equation. To generate the neutronic group constants, which are macroscopic cross sections that serve as input data for the neutronic codes, it was used the WIMSD-5B cell calculation code. The neutron analysis code PARCS was also used to model the 3D RMB core in order to compare the results of radial and axial average power distribution with the results generated by RELAP5-3D© code and with the available results of the CITATION neutron kinetic code. The safety analyses demonstrated safe behavior of the reactor through situations of possible transients. The 3D coupled calculations to the steady state operation also showed expected behavior, as well as the RMB neutronic analyzes performed with the codes NESTLE and PARCS.(author)

Study on coupling of three-dimension space time neutron kinetics model and RELAP5 and improvement of RELAP5

International Nuclear Information System (INIS)

Gui Xuewen; Cai Qi; Luo Bangqi

2007-01-01

A two-group three-dimension space-time neutron kinetics model is applied to the RELAP5 code, which replaces the point reactor kinetics model. A visual operation interface is designed to convenience interactive operation between operator and computer. The calculation results and practical applications indicate that the functions and precision of improved RELAP5 are enhanced and can be easily used. The improved RELAP5 has a good application perspective in nuclear power plant simulation. (authors)

Construction of the neutron beam facility at Australia's OPAL research reactor

International Nuclear Information System (INIS)

Kennedy, Shane J.

2006-01-01

Australia's new research reactor, OPAL, has been designed principally for neutron beam science and radioisotope production. It has a capacity for 18 neutron beam instruments, located at the reactor face and in a neutron guide hall. The neutron beam facility features a 20 l liquid deuterium cold neutron source and cold and thermal supermirror neutron guides. Nine neutron beam instruments are under development, of which seven are scheduled for completion in early 2007. The project is approaching the hot-commissioning stage, when criticality will be demonstrated. Installation of the neutron beam transport system and neutron beam instruments in the neutron guide hall and at the reactor face is underway, and the path to completion of this project is relatively clear. This paper will outline the key features of the OPAL reactor, and will describe the neutron beam facility in particular. The status of the construction and a forecast of the program to completion, including commissioning and commencement of routine operation in 2007 will also be discussed

Development and verification of an efficient spatial neutron kinetics method for reactivity-initiated event analyses

International Nuclear Information System (INIS)

Ikeda, Hideaki; Takeda, Toshikazu

2001-01-01

A space/time nodal diffusion code based on the nodal expansion method (NEM), EPISODE, was developed in order to evaluate transient neutron behavior in light water reactor cores. The present code employs the improved quasistatic (IQS) method for spatial neutron kinetics, and neutron flux distribution is numerically obtained by solving the neutron diffusion equation with the nonlinear iteration scheme to achieve fast computation. A predictor-corrector (PC) method developed in the present study enabled to apply a coarse time mesh to the transient spatial neutron calculation than that applicable in the conventional IQS model, which improved computational efficiency further. Its computational advantage was demonstrated by applying to the numerical benchmark problems that simulate reactivity-initiated events, showing reduction of computational times up to a factor of three than the conventional IQS. The thermohydraulics model was also incorporated in EPISODE, and the capability of realistic reactivity event analyses was verified using the SPERT-III/E-Core experimental data. (author)

Background determination for the neutron-neutron scattering experiment at the reactor YAGUAR

Energy Technology Data Exchange (ETDEWEB)

Muzichka, A.Yu. [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Furman, W.I. [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Lychagin, E.V. [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Krylov, A.R. [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Nekhaev, G.V. [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Sharapov, E.I. [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Shvetsov, V.N. [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Strelkov, A.V. [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Levakov, B.G. [Russian Federal Nuclear Center-All-Russian Research Institute of Technical Physics, PO Box 245, 456770 Snezhinsk (Russian Federation); Lyzhin, A.E. [Russian Federal Nuclear Center-All-Russian Research Institute of Technical Physics, PO Box 245, 456770 Snezhinsk (Russian Federation); Chernukhin, Yu.I. [Russian Federal Nuclear Center-All-Russian Research Institute of Technical Physics, PO Box 245, 456770 Snezhinsk (Russian Federation); Kandiev, Ya.Z. [Russian Federal Nuclear Center-All-Russian Research Institute of Technical Physics, PO Box 245, 456770 Snezhinsk (Russian Federation); Howell, C.R. [Duke University and Triangle Universities Nuclear Laboratory, Durham, NC 27708-0308 (United States); Mitchell, G.E. [North Carolina State University, Raleigh, NC 27695-8202 (United States); Triangle Universities Nuclear Laboratory, Durham, NC 27708-0308 (United States); Crawford, B.E. [Gettysburg College, Box 405, Gettysburg, PA 17325 (United States); Stephenson, S.L. [Gettysburg College, Box 405, Gettysburg, PA 17325 (United States)]. E-mail: sstephen@gettysburg.edu; Tornow, W. [Duke University and Triangle Universities Nuclear Laboratory, Durham, NC 27708-0308 (United States)

2007-06-01

The motivation and design is outlined for the experiment to measure the neutron-neutron singlet scattering length directly with thermal neutrons at the pulsed reactor YAGUAR. A statistical accuracy of 3% can be reached, though achieving the goal of an overall accuracy of 3-5% for the nn-scattering length depends on the background level. Possible sources of background are discussed in depth and the results of extensive modeling of the background are presented. Measurements performed at YAGUAR to test these background calculations are described. The experimental results indicate an anticipated background level up to 30% relative to the expected nn effect at the maximal energy burst of the reactor. The conclusion is made that the nn experiment at YAGUAR is feasible to produce the first directly measured value for the neutron-neutron scattering length.

Neutron dosimetry. Environmental monitoring in a BWR type reactor; Dosimetria de neutrones. Monitoreo ambiental en un reactor del tipo BWR

Energy Technology Data Exchange (ETDEWEB)

Tavera D, L; Camacho L, M E

1991-01-15

The measurements carried out on reactor dosimetry are applied mainly to the study on the effects of the radiation in 108 materials of the reactor; little is on the environmental dosimetry outside of the primary container of BWR reactors. In this work the application of a neutron spectrometer formed by plastic detectors of nuclear traces manufactured in the ININ, for the environmental monitoring in penetrations around the primary container of the unit I of the Laguna Verde central is presented. The neutron monitoring carries out with purposes of radiological protection, during the operational tests of the reactor. (Author)

Limitations for qualitative and quantitative neutron activation analysis using reactor neutrons

International Nuclear Information System (INIS)

El-Abbady, W.H.; El-Tanahy, Z.H.; El-Hagg, A.A.; Hassan, A.M.

1999-01-01

In this work, the most important limitations for qualitative and quantitative analysis using reactor neutrons for activation are reviewed. Each limitation is discussed using different examples of activated samples. Photopeak estimation, nuclear reactions interference and neutron flux measurements are taken into consideration. Solutions for high accuracy evaluation in neutron activation analysis applications are given. (author)

Gamma compensated pulsed ionization chamber wide range neutron/reactor power measurement system

International Nuclear Information System (INIS)

Ellis, W.H.

1975-01-01

An improved method and system of pulsed mode operation of ionization chambers is described in which a single sensor system with gamma compensation is provided by sampling, squaring, automatic gate selector, and differential amplifier circuit means, employed in relation to chambers sensitized to neutron plus gamma and gamma only to subtract out the gamma component, wherein squaring functions circuits, a supplemental high performance pulse rate system, and operational and display mode selection and sampling gate circuits are utilized to provide automatic wide range linear measurement capability for neutron flux and reactor power. Neon is employed as an additive in the ionization chambers to provide independence of ionized gas kinetics temperature effects, and the pulsed mode of operation provide independence of high temperature insulator leakage effects. (auth)

On the Modeling of Local Neutronically-Coupled Flow-Induced Oscillations in Advanced Boiling Water Reactors

International Nuclear Information System (INIS)

Aniel-Buchheit, Sylvie; Podowski, Michael Z.

2006-01-01

The purpose of this paper is to discuss the development in progress of a complete space- and time-dependent model of the coupled neutron kinetic and reactor thermal-hydraulics. The neutron kinetics model is based on two-group diffusion equations with Doppler and void reactivity feedback effects. This model is coupled with the model of two-phase flow and heat transfer in parallel coolant channels. The modeling concepts considered for this purpose include one-dimensional drift flux and two-fluid models, as well a CFD model implemented in the NPHASE advanced computational multiphase fluid dynamics (CMFD) computer code. Two methods of solution for the overall model are proposed. One is based on direct numerical integration of the spatially-discretized governing equations. The other approach is based on a quasi-analytical modal approach to the neutronics model, in which a complete set of eigenvectors is found for step-wise temporal changes of the cross-sections of core materials (fuel and coolant/moderator). The issues investigated in the paper include details of model formulation, as well as the results of calculations for neutronically-coupled density-wave oscillations. (authors)

Neutron flux measurements in PUSPATI Triga Reactor

International Nuclear Information System (INIS)

Gui Ah Auu; Mohamad Amin Sharifuldin Salleh; Mohamad Ali Sufi.

1983-01-01

Neutron flux measurement in the PUSPATI TRIGA Reactor (PTR) was initiated after its commissioning on 28 June 1982. Initial measured thermal neutron flux at the bottom of the rotary specimen rack (rotating) and in-core pneumatic terminus were 3.81E+11 n/cm 2 sec and 1.10E+12n/cm 2 sec respectively at 100KW. Work to complete the neutron flux data are still going on. The cadmium ratio, thermal and epithermal neutron flux are measured in the reactor core, rotary specimen rack, in-core pneumatic terminus and thermal column. Bare and Cadmium covered gold foils and wires are used for the above measurement. The activities of the irradiated gold foils and wires are determined using Ge(Li) and hyperpure germinium detectors. (author)

Application of the Wiener-Hermite functional method to point reactor kinetics driven by random reactivity fluctuations

International Nuclear Information System (INIS)

Behringer, K.; Pineyro, J.; Mennig, J.

1990-06-01

The Wiener-Hermite functional (WHF) method has been applied to the point reactor kinetic equation excited by Gaussian random reactivity noise under stationary conditions. Delayed neutrons and any feedback effects are disregarded. The neutron steady-state value and the power spectral density (PSD) of the neutron flux have been calculated in a second order (WHF-2) approximation. Two cases are considered: in the first case, the noise source is low-pass white noise. In both cases the WHF-2 approximation of the neutron PSDs leads to relatively simple analytical expressions. The accuracy of the approach is determined by comparison with exact solutions of the problem. The investigations show that the WHF method is a powerful approximative tool for studying the nonlinear effects in the stochastic differential equation. (author) 5 figs., 29 refs

AUS, Neutron Transport and Gamma Transport System for Fission Reactors and Fusion Reactors

International Nuclear Information System (INIS)

1990-01-01

1 - Description of program or function: AUS is a neutronics code system which may be used for calculations of a wide range of fission reactors, fusion blankets and other neutron applications. The present version, AUS98, has a nuclear cross section library based on ENDF/B-VI and includes modules which provide for reactor lattice calculations, one-dimensional transport calculations, multi-dimensional diffusion calculations, cell and whole reactor burnup calculations, and flexible editing of results. Calculations of multi-region resonance shielding, coupled neutron and photon transport, energy deposition, fission product inventory and neutron diffusion are combined within the one code system. The major changes from the previous release, AUS87, are the inclusion of a cross-section library based on ENDF/B-VI, the addition of the POW3D multi-dimensional diffusion module, the addition of the MICBURN module for controlling whole reactor burnup calculations, and changes to the system as a consequence of moving from IBM mainframe computers to UNIX workstations. 2 - Method of solution: AUS98 is a modular system in which the modules are complete programs linked by a path given in the input stream. A simple path is simply a sequence of modules, but the path is actually pre-processed and compiled using the Fortran 77 compiler. This provides for complex module linking if required. Some of the modules included in AUS98 are: MIRANDA Cross-section generation in a multi-region resonance subgroup calculation and preliminary group condensation. ANAUSN One-dimensional discrete ordinates calculation. ICPP Isotropic collision probability calculation in one dimension and for rod clusters. POW3D Multi-dimensional neutron diffusion calculation including feedback-free kinetics. AUSIDD One-dimensional diffusion calculation. EDITAR Reaction-rate editing and group collapsing following a transport calculation. CHAR Lattice and global burnup calculation. MICBURN Control of global burnup

Spatial neutronics modelling to evaluate the temperature reactivity feedbacks in a lead-cooled fast reactor - 15288

International Nuclear Information System (INIS)

Lorenzi, S.; Cammi, A.; Luzzi, L.

2015-01-01

The qualitative and quantitative assessment of the thermal reactivity feedbacks occurring in a nuclear reactor is a crucial issue for the time-dependent evolution of the system and, in turn, it has a great impact on the development and validation of advanced control techniques. In the present work, in order to overcome the limitations of the classic Point Kinetics adopted in the control simulators, a spatial neutronics model, representing the neutron flux as sum of a spatial basis weighted by time-dependent coefficients, has been considered. The reference reactor is ALFRED, the European demonstrator of the Lead-cooled Fast Reactor technology. Average cross-sections for each fuel assembly, calculated by means of a Monte Carlo code, have been used to solve the partial differential equations of the neutron diffusion, exploiting the capabilities of the COMSOL software. Once obtained the spatial functions, the set of equations for studying the reactivity effects has been implemented in the MATLAB environment. Among the several temperature reactivity feedbacks, specific attention has been paid to the Doppler effect in the fuel and to the lead density effect. Several spatial bases have been calculated and their capability of representing the reactivity variation have been assessed. (authors)

maximum neutron flux at thermal nuclear reactors

International Nuclear Information System (INIS)

Strugar, P.

1968-10-01

Since actual research reactors are technically complicated and expensive facilities it is important to achieve savings by appropriate reactor lattice configurations. There is a number of papers, and practical examples of reactors with central reflector, dealing with spatial distribution of fuel elements which would result in higher neutron flux. Common disadvantage of all the solutions is that the choice of best solution is done starting from the anticipated spatial distributions of fuel elements. The weakness of these approaches is lack of defined optimization criteria. Direct approach is defined as follows: determine the spatial distribution of fuel concentration starting from the condition of maximum neutron flux by fulfilling the thermal constraints. Thus the problem of determining the maximum neutron flux is solving a variational problem which is beyond the possibilities of classical variational calculation. This variational problem has been successfully solved by applying the maximum principle of Pontrjagin. Optimum distribution of fuel concentration was obtained in explicit analytical form. Thus, spatial distribution of the neutron flux and critical dimensions of quite complex reactor system are calculated in a relatively simple way. In addition to the fact that the results are innovative this approach is interesting because of the optimization procedure itself [sr

Neutron personnel dosimetry considerations for fusion reactors

International Nuclear Information System (INIS)

Barton, T.P.; Easterly, C.E.

1979-07-01

The increasing development of fusion reactor technology warrants an evaluation of personnel neutron dosimetry systems to aid in the concurrent development of a radiation protection program. For this reason, current state of knowledge neutron dosimeters have been reviewed with emphasis placed on practical utilization and the problems inherent in each type of dosimetry system. Evaluations of salient parameters such as energy response, latent image instability, and minimum detectable dose equivalent are presented for nuclear emulsion films, track etch techniques, albedo and other thermoluminescent dosimetry techniques, electrical conductivity damage effects, lyoluminescence, thermocurrent, and thermally stimulated exoelectron emission. Brief summaries of dosimetry regulatory requirements and intercomparison study results help to establish compliance and recent trends, respectively. Spectrum modeling data generated by the Neutron Physics Division of Oak Ridge National Laboratory for the Princeton Tokamak Fusion Test Reactor (TFTR) Facility have been analyzed by both International Commission on Radiological Protection fluence to dose conversion factors and an adjoint technique of radiation dosimetry, in an attempt to determine the applicability of current neutron dosimetry systems to deuterium and tritium fusion reactor leakage spectra. Based on the modeling data, a wide range of neutron energies will probably be present in the leakage spectra of the TFTR facility, and no appreciable risk of somatic injury to occupationally exposed workers is expected. The relative dose contributions due to high energy and thermal neutrons indicate that neutron dosimetry will probably not be a serious limitation in the development of fusion power

Condensed matter and materials research using neutron diffraction and spectroscopy: reactor and pulsed neutron sources

International Nuclear Information System (INIS)

Bisanti, Paola; Lovesey, S.W.

1987-05-01

The paper provides a short, and partial view of the neutron scattering technique applied to condensed matter and materials research. Reactor and accelerator-based neutron spectrometers are discussed, together with examples of research projects that illustrate the puissance and modern applications of neutron scattering. Some examples are chosen to show the range of facilities available at the medium flux reactor operated by Casaccia ENEA, Roma and the advanced, pulsed spallation neutron source at the Rutherford Appleton Laboratory, Oxfordshire. (author)

Simplified methodology for control cell constant calculations of the reactor cores for the space kinetics

International Nuclear Information System (INIS)

Santos, Rubens Souza dos; Martinez, Aquilino Senra; Alvim, Antonio Carlos Marques

2002-01-01

In this work is presented a methodology which focuses the distribution of neutron absorber rods in nuclear reactor power plants, for utilizing in space kinetic calculations, principally in the cluster ejection transients of control rods. A numerical model for macroscopic constant calculations based on the knowledge of the neutron flux without the control rods is proposed, as alternative to the analytical models, based on the hypothesis of the null current on the cell super boundaries. The proposed model in this work has itself showed adequate to deal with problems with strong space dependence, once that the model showed consistence in the global average built in the analytical model. (author)

Progress on RMC: a Monte Carlo neutron transport code for reactor analysis

International Nuclear Information System (INIS)

Wang, Kan; Li, Zeguang; She, Ding; Liu, Yuxuan; Xu, Qi; Shen, Huayun; Yu, Ganglin

2011-01-01

This paper presents a new 3-D Monte Carlo neutron transport code named RMC (Reactor Monte Carlo code), specifically intended for reactor physics analysis. This code is being developed by Department of Engineering Physics in Tsinghua University and written in C++ and Fortran 90 language with the latest version of RMC 2.5.0. The RMC code uses the method known as the delta-tracking method to simulate neutron transport, the advantages of which include fast simulation in complex geometries and relatively simple handling of complicated geometrical objects. Some other techniques such as computational-expense oriented method and hash-table method have been developed and implemented in RMC to speedup the calculation. To meet the requirements of reactor analysis, the RMC code has the calculational functions including criticality calculation, burnup calculation and also kinetics simulation. In this paper, comparison calculations of criticality problems, burnup problems and transient problems are carried out using RMC code and other Monte Carlo codes, and the results show that RMC performs quite well in these kinds of problems. Based on MPI, RMC succeeds in parallel computation and represents a high speed-up. This code is still under intensive development and the further work directions are mentioned at the end of this paper. (author)

Construction of the Neutron Beam Facility at Australia's OPAL Research Reactor

International Nuclear Information System (INIS)

Kennedy, J.S.

2005-01-01

Full text: Australia's new research reactor, OPAL, has been designed for high quality neutron beam science and radioisotope production. It has a capacity for eighteen neutron beam instruments to be located at the reactor face and in a neutron guide hall. The new neutron beam facility features a 20 litre liquid deuterium cold neutron source and supermirror neutron reflecting guides for intense cold and thermal neutron beams. Nine neutron beam instruments are under development, of which seven are scheduled for completion in early 2007. The project is approaching the hot-commissioning stage, where criticality will be demonstrated. Installation of the neutron beam transport system and neutron beam instruments in the neutron guide hall and at the reactor face is underway, and the path to completion of this project is relatively clear. The lecture will outline Australia's aspirations for neutron science at the OPAL reactor, and describe the neutron beam facility under construction. The status of this project and a forecast of the program to completion, including commissioning and commencement of routine operation in 2007 will also be discussed. This project is the culmination of almost a decade of effort. We now eagerly anticipate catapulting Australia's neutron beam science capability to meet the best in the world today. (author)

Updated neutron spectrum characterization of SNL baseline reactor environments

International Nuclear Information System (INIS)

Griffin, P.J.; Kelly, J.G.; Vehar, D.W.

1994-04-01

This document provides SAND-II and MANIPULATE output listings from calculations used to derive the new spectrum-averaged integral parameters that were reported in volume 1. When used in conjunction with volume 1, this document provides an audit trail for the neutron radiation field characterization and supports current quality assurance initiatives. This document provides detailed information on the neutron spectrum characteristics of the primary Sandia National Laboratories' (SNL) reactor environments. The information in this volume is not intended for the casual user of the SNL reactor facilities. This detailed characterization of the neutron and gamma environments at the Sandia Pulsed Reactor (SPR) and the Annular Core Research Reactor (ACRR) is provided to aid the users who wish to convert the information given in the Radiation Metrology Laboratory (RML) dosimetry reports into other (non-silicon) measures of neutron damage. The spectra provided in these appendices can be used as a source term for Monte Carlo radiation transport calculations to study the impact of experimenter's test package on the neutron environment

Utilization of cold neutron beams at intermediate flux reactors

International Nuclear Information System (INIS)

Clark, D.D.

1992-01-01

With the advent of cold neutron beam (CNB) facilities at U.S. reactors [National Institute of Standards and Technology (NIST) in 1991; Cornell University and the University of Texas at Austin, anticipated in 1992], it is appropriate to reexamine the types of research for which they are likely to be best suited or uniquely suited. With the exception of a small-angle neutron scattering facility at Brookhaven National Laboratory, there has been no prior experience in the United States with such beams, but they have been extensively used at European reactors where cold neutron sources and neutron guides were developed some years age. This paper does not discuss specialized cases such as ultracold neutrons or very high flux facilities such as the Institute Laue-Langevin ractor and the proposed advanced neutron source. Instead, it concentrates on potential utilization of CNBs at intermediate-flux reactors such as at Cornell and Texas, i.e., in the 1-MW range and operated <24 h a day

EL-2 reactor: Thermal neutron flux distribution; EL-2: Repartition du flux de neutrons thermiques

Energy Technology Data Exchange (ETDEWEB)

Rousseau, A; Genthon, J P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

The flux distribution of thermal neutrons in EL-2 reactor is studied. The reactor core and lattices are described as well as the experimental reactor facilities, in particular, the experimental channels and special facilities. The measurement shows that the thermal neutron flux increases in the central channel when enriched uranium is used in place of natural uranium. However the thermal neutron flux is not perturbed in the other reactor channels by the fuel modification. The macroscopic flux distribution is measured according the radial positioning of fuel rods. The longitudinal neutron flux distribution in a fuel rod is also measured and shows no difference between enriched and natural uranium fuel rods. In addition, measurements of the flux distribution have been effectuated for rods containing other material as steel or aluminium. The neutron flux distribution is also studied in all the experimental channels as well as in the thermal column. The determination of the distribution of the thermal neutron flux in all experimental facilities, the thermal column and the fuel channels has been made with a heavy water level of 1825 mm and is given for an operating power of 1000 kW. (M.P.)

Experimental determination of neutron temperature distribution in reactor cell

International Nuclear Information System (INIS)

Bosevski, T.

1965-12-01

This paper describes theoretical preparation of the experiment for measuring neutron temperature distribution at the RB reactor by activation foils. Due to rather low neutron flux Cu and Lu foil were irradiated for 4 days. Special natural uranium fuel element was prepared to enable easy removal of foils after irradiation. Experimental device was placed in the reactor core at half height in order to measure directly the mean neutron density. Experimental data of neutron temperature distribution for square lattice pitch 16 cm are presented with mean values of neutron temperature in the moderator, in the fuel and on the fuel element surface

Critical fluctuations of the number of neutrons in a reactor

International Nuclear Information System (INIS)

Ryazanov, V.V.; Lakoza, E.L.; Sysoev, V.M.

1995-01-01

The nuclear chain reaction is the most important physical process in a reactor. The theory of nuclear chain reaction fluctuations (neutron noise), developed in and other studies, has given results that are important for reactor physics and reactor practice (correlation analysis of neutron noise for measurement of the physical characteristics and reactor monitoring, stability of the critical state, etc.). Here we propose to study these problems by applying the methods of continuous phase transitions and synergetics and using the analogy with chemical chain reactions and the general laws of critical phenomena. The optimal reactor operating conditions are critical. To predict how a critical reactor will behave it is necessary to reveal those features of the neutron laws that are universal in some way, i.e., do not depend on the details of the individual acts of neutron motion and transformation that occur in reactors of different types. The similarity between chemical and nuclear chain reactions was noted long ago. Consequently, a universal theory of continuous phase transition was developed for systems of diverse physical nature

TAR-1 A programme for the determination of time behaviour of neutron density on a thermal reactor; TAR-1. Programa para la determinacion del comportamiento temporal de la densidad neutronica en reactor termico

Energy Technology Data Exchange (ETDEWEB)

Torres Vida, J

1963-07-01

This programme, written for the UNIVAC-UCT of J.E.N., obtain the time behaviour of neutron density as a function of both positive and negative step change in reactivity. These results are obtained from solutions of the space-independent kinetic equations of a bare thermal reactor based on the Fermi continuous slowing down model and using six groups of delayed neutrons. (Author) 3 refs.

Neutronics comparative analysis between MNSR and slowpoke-II reactors

International Nuclear Information System (INIS)

Khamis, I.; Khattab, K.

1999-01-01

Neutronics analysis of both MNSR and Slowpoke reactors were made. Calculations including flux distribution, power estimation, excess and shutdown reactivity margins, flooding effects of irradiation sites, and initial investigation of fuel conversion from high to low enriched uranium were discussed. A neutronic 3-D model, dedicated mainly for the MNSR, has been developed to perform such neutronic calculations for both reactors. Well-known cell and core calculation codes such as WIMSD4 and CITATIONS have been used. It was found out that it is possible to lower the fuel enrichment of the Miniature Neutron Source Reactor (MNSR) to 20% using U O 2 as fuel instead of U Al 4 . The number of fuel elements required for the new core is 199. The use of double thickness of the bottom reflector in Slowpoke reactor made it possible to load the reactor with lower enriched fuel compared to MNSR. Values of reactivity flooding effects for single or combination of inner irradiation sites were obtained accurately. Results show good agreement with reported data for MNSR. (author)

Four energy group neutron flux distribution in the Syrian miniature neutron source reactor using the WIMSD4 and CITATION code

International Nuclear Information System (INIS)

Khattab, K.; Omar, H.; Ghazi, N.

2009-01-01

A 3-D (R, θ , Z) neutronic model for the Miniature Neutron Source Reactor (MNSR) was developed earlier to conduct the reactor neutronic analysis. The group constants for all the reactor components were generated using the WIMSD4 code. The reactor excess reactivity and the four group neutron flux distributions were calculated using the CITATION code. This model is used in this paper to calculate the point wise four energy group neutron flux distributions in the MNSR versus the radius, angle and reactor axial directions. Good agreement is noticed between the measured and the calculated thermal neutron flux in the inner and the outer irradiation site with relative difference less than 7% and 5% respectively. (author)

Thermal neutron flux distribution in ET-RR-2 reactor thermal column

Directory of Open Access Journals (Sweden)

Imam Mahmoud M.

2002-01-01

Full Text Available The thermal column in the ET-RR-2 reactor is intended to promote a thermal neutron field of high intensity and purity to be used for following tasks: (a to provide a thermal neutron flux in the neutron transmutation silicon doping, (b to provide a thermal flux in the neutron activation analysis position, and (c to provide a thermal neutron flux of high intensity to the head of one of the beam tubes leading to the room specified for boron thermal neutron capture therapy. It was, therefore, necessary to determine the thermal neutron flux at above mentioned positions. In the present work, the neutron flux in the ET-RR-2 reactor system was calculated by applying the three dimensional diffusion depletion code TRITON. According to these calculations, the reactor system is composed of the core, surrounding external irradiation grid, beryllium block, thermal column and the water reflector in the reactor tank next to the tank wall. As a result of these calculations, the thermal neutron fluxes within the thermal column and at irradiation positions within the thermal column were obtained. Apart from this, the burn up results for the start up core calculated according to the TRITION code were compared with those given by the reactor designer.

Empiric model for mean generation time adjustment factor for classic point kinetics equations

Energy Technology Data Exchange (ETDEWEB)

Goes, David A.B.V. de; Martinez, Aquilino S.; Goncalves, Alessandro da C., E-mail: david.goes@poli.ufrj.br, E-mail: aquilino@lmp.ufrj.br, E-mail: alessandro@con.ufrj.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Departamento de Engenharia Nuclear

2017-11-01

Point reactor kinetics equations are the easiest way to observe the neutron production time behavior in a nuclear reactor. These equations are derived from the neutron transport equation using an approximation called Fick's law leading to a set of first order differential equations. The main objective of this study is to review classic point kinetics equation in order to approximate its results to the case when it is considered the time variation of the neutron currents. The computational modeling used for the calculations is based on the finite difference method. The results obtained with this model are compared with the reference model and then it is determined an empirical adjustment factor that modifies the point reactor kinetics equation to the real scenario. (author)

Empiric model for mean generation time adjustment factor for classic point kinetics equations

International Nuclear Information System (INIS)

Goes, David A.B.V. de; Martinez, Aquilino S.; Goncalves, Alessandro da C.

2017-01-01

Point reactor kinetics equations are the easiest way to observe the neutron production time behavior in a nuclear reactor. These equations are derived from the neutron transport equation using an approximation called Fick's law leading to a set of first order differential equations. The main objective of this study is to review classic point kinetics equation in order to approximate its results to the case when it is considered the time variation of the neutron currents. The computational modeling used for the calculations is based on the finite difference method. The results obtained with this model are compared with the reference model and then it is determined an empirical adjustment factor that modifies the point reactor kinetics equation to the real scenario. (author)

Aspects of the physics and chemistry of water radiolysis by fast neutrons and fast electrons in nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

McCracken, D.R. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Tsang, K.T. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Laughton, P.J

1998-09-01

Detailed radiation physics calculations of energy deposition have been done for the coolant of CANDU reactors and Pressurized Water Reactors (PWRs). The geometry of the CANDU fuel channel was modelled in detail. Fluxes and energy-deposition rates for neutrons, recoil ions, photons, and fast electrons have been calculated using MCNP4B, WIMS-AECL, and specifically derived energy-transfer factors. These factors generate the energy/flux spectra of recoil ions from fast-neutron energy/flux spectra. The energy spectrum was divided into 89 discrete ranges (energy bins).The production of oxidizing species and net coolant radiolysis can be suppressed by the addition of hydrogen to the coolant of nuclear reactors. It is argued that the net dissociation of coolant by gamma rays is suppressed by lower levels of excess hydrogen than when dissociation is by ion recoils. This has consequences for the modelling of coolant radiolysis by homogeneous kinetics. More added hydrogen is required to stop water radiolysis by recoil ions acting alone than if recoil ions and gamma rays acted concurrently in space and time. Homogeneous kinetic models and experimental data suggest that track overlap is very inefficient in providing radicals from gamma-ray tracks to recombine molecular products in ion-recoil tracks. An inhomogeneous chemical model is needed that incorporates ionizing-particle track structure and track overlap. Such a model does not yet exist, but a number of limiting cases using homogeneous kinetics are discussed. There are sufficient uncertainties and contradictions in the data relevant to the radiolysis of reactor coolant that the relatively high CHC's (critical hydrogen concentration) observed in NRU reactor experiments (compared to model predictions) may be explainable by errors in fundamental data and understanding of water radiolysis under reactor conditions. The radiation chemistry program at CRL has been focused to generate quantitative water-radiolysis data in a

Neutron Fluence And DPA Rate Analysis In Pebble-Bed HTR Reactor Vessel Using MCNP

Science.gov (United States)

Hamzah, Amir; Suwoto; Rohanda, Anis; Adrial, Hery; Bakhri, Syaiful; Sunaryo, Geni Rina

2018-02-01

In the Pebble-bed HTR reactor, the distance between the core and the reactor vessel is very close and the media inside are carbon and He gas. Neutron moderation capability of graphite material is theoretically lower than that of water-moderated reactors. Thus, it is estimated much more the fast neutrons will reach the reactor vessel. The fast neutron collisions with the atoms in the reactor vessel will result in radiation damage and could be reducing the vessel life. The purpose of this study was to obtain the magnitude of neutron fluence in the Pebble-bed HTR reactor vessel. Neutron fluence calculations in the pebble-bed HTR reactor vessel were performed using the MCNP computer program. By determining the tally position, it can be calculated flux, spectrum and neutron fluence in the position of Pebble-bed HTR reactor vessel. The calculations results of total neutron flux and fast neutron flux in the reactor vessel of 1.82x108 n/cm2/s and 1.79x108 n/cm2/s respectively. The fast neutron fluence in the reactor vessel is 3.4x1017 n/cm2 for 60 years reactor operation. Radiation damage in stainless steel material caused by high-energy neutrons (> 1.0 MeV) will occur when it has reached the neutron flux level of 1.0x1024 n/cm2. The neutron fluence results show that there is no radiation damage in the Pebble-bed HTR reactor vessel, so it is predicted that it will be safe to operate at least for 60 years.

The spatial kinetic analysis of accelerator-driven subcritical reactor

International Nuclear Information System (INIS)

Takahashi, H.; An, Y.; Chen, X.

1998-02-01

The operation of the accelerator driven reactor with subcritical condition provides a more flexible choice of the reactor materials and of design parameters. A deep subcriticality is chosen sometime from the analysis of point kinetics. When a large reactor is operated in deep subcritical condition by using a localized spallation source, the power distribution has strong spatial dependence, and point kinetics does not provide proper analysis for reactor safety. In order to analyze the spatial and energy dependent kinetic behavior in the subcritical reactor, the authors developed a computation code which is composed of two parts, the first one is for creating the group cross section and the second part solves the multi-group kinetic diffusion equations. The reactor parameters such as the cross section of fission, scattering, and energy transfer among the several energy groups and regions are calculated by using a code modified from the Monte Carlo codes MCNPA and LAHET instead of the usual analytical method of ANISN, TWOTRAN codes. Thus the complicated geometry of the accelerator driven reactor core can be precisely taken into account. The authors analyzed the subcritical minor actinide transmutor studied by Japan Atomic Energy Research Institute (JAERI) using the code

Utilization of RP-10 reactor for neutron therapy

International Nuclear Information System (INIS)

Paucar, R.; Nieto, M.; Parreno, F.; Vela, M.; Pozo, Z.

1997-01-01

In the Nuclear Energy Peruvian Institute, IPEN, a research area has established of Neutron Radiotherapy, know as NCT. This research joins the physics of particles (Neutrons and photons) and Medical Physics, and this one is an applied investigation where in considering the construction of a treatment hall in Huarangal (Peru) Reactor's irradiation facility, it can treat patients with brain tumors. In Neutron Therapy (NCT), it tries to use neutrons to destroy tumor cells where other therapeutic techniques are not effective. This process consist on to incise a neutrons beam of adequate characteristics over the tumor area of the patient. The neutrons used are of thermal energy and therefore irradiations are developed in experimental reactors. For this one, it is used horizontal channels prepared suitably. Before the irradiation, it is injected to the patient a substance which is absorbed by tumoral tissue. The substance components will be B-10, nuclide with an absorption cross section high to thermal neutrons (3837 b). The B-10 irradiate with thermal neutrons produce alpha particles of short reach (10 μm. on soft tissue) and with LET values (lineal energy transference) very high. The result is a cell preferential destruction which have absorbed the substance and it's next neighbors, like the cell size is 10 μm. This process as know as Boron Neutron Capture Therapy (BNCT). This work describes Peruvian RP-10 reactor and recently efforts to assess the design and feasibility of the medical neutron irradiation facility for NCT. (author). 22 refs., 6 tabs

A new neutron noise technique for fast reactors

International Nuclear Information System (INIS)

Zhuo Fengguan; Jin Manyi; Yao Shigui; Su Zhuting

1987-12-01

This paper gives a new neutron noise technique for fast reactors, which is known as thermalization measurement technique of the neutron noise. The theoretical formulas of the technique were developed, and a digital delayed coincidence time analyzer consisted of TTL integrated circuits was constructed for the study of this technique. The technique has been tested and applied practically at Df-VI fast zero power reactor. It was shown that the provided technique in this work has a number of significant advantages in comparison with the conventional neutron noise method

Neutron fluence determination for light water reactor pressure vessels

International Nuclear Information System (INIS)

Gold, R.

1994-01-01

A general description of limitations that exist in pressure vessel neutron fluence determinations for commercial light water reactors is presented. Complexity factors that arise in light water reactor pressure vessel neutron fluence calculations are identified and used to analyze calculational limitations. Two broad categories of calculational limitations are introduced, namely benchmark field limitations and deep penetration limitations. Explicit examples of limitations that can arise in each of these two broad categories are presented. These limitations are used to show that the recent draft regulatory guide for the determination of pressure vessel neutron fluence, developed by the Nuclear Regulatory Commission, is based upon procedures and assumptions that are not valid. To eliminate the complexity and limitations of calculational methods, it is recommended that the determination of light water reactor pressure vessel neutron fluence be based upon experiment. Recommendations for improved methods of pressure vessel surveillance neutron dosimetry are advanced

Coupling of unidimensional neutron kinetics to thermal hydraulics in parallel channels

International Nuclear Information System (INIS)

Cecenas F, M.; Campos G, R.M.

2003-01-01

In this work the dynamic behavior of a consistent system in fifteen channels in parallel that represent the reactor core of a BWR type, coupled of a kinetic neutronic model in one dimension is studied by means of time series. The arrangement of channels is obtained collapsing the assemblies that it consists the core to an arrangement of channels prepared in straight lines, and it is coupled to the unidimensional solution of the neutron diffusion equation. This solution represents the radial power distribution, and initially the static solution is obtained to verify that the one modeling core is critic. The coupled set nuclear-thermal hydraulics it is solved numerically by means of a net of CPUs working in the outline teacher-slave by means of Parallel Virtual Machine (PVM), subject to the restriction that the pressure drop is equal for each channel, which is executed iterating on the refrigerant distribution. The channels are dimensioned according to the one Stability Benchmark of the Ringhals swedish plant, organized by the Nuclear Energy Agency in 1994. From the information of this benchmark it is obtained the axial power profile for each channel, which is assumed as invariant in the time. To obtain the time series, the system gets excited with white noise (sequence that statistically obeys to a normal distribution with zero media), so that the power generated in each channel it possesses the same ones characteristics of a typical signal obtained by means of the acquisition of those signals of neutron flux in a BWR reactor. (Author)

SHOVAV-JUEL. A one dimensional space-time kinetic code for pebble-bed high-temperature reactors with temperature and Xenon feedback

International Nuclear Information System (INIS)

Nabbi, R.; Meister, G.; Finken, R.; Haben, M.

1982-09-01

The present report describes the modelling basis and the structure of the neutron kinetics-code SHOVAV-Juel. Information for users is given regarding the application of the code and the generation of the input data. SHOVAV-Juel is a one-dimensional space-time-code based on a multigroup diffusion approach for four energy groups and six groups of delayed neutrons. It has been developed for the analysis of the transient behaviour of high temperature reactors with pebble-bed core. The reactor core is modelled by horizontal segments to which different materials compositions can be assigned. The temperature dependence of the reactivity is taken into account by using temperature dependent neutron cross sections. For the simulation of transients in an extended time range the time dependence of the reactivity absorption by Xenon-135 is taken into account. (orig./RW)

Kinetic parameters of a material test research reactor fueled with various low enriched uranium dispersion fuels

International Nuclear Information System (INIS)

Muhammad, Farhan; Majid, Asad

2009-01-01

The effects of using different low enriched uranium fuels, having same uranium density, on the kinetic parameters of a material test research reactor were studied. For this purpose, the original aluminide fuel (UAl x -Al) containing 4.40 gU/cm 3 of an MTR was replaced with silicide (U 3 Si-Al and U 3 Si 2 -Al) and oxide (U 3 O 8 -Al) dispersion fuels having the same uranium density as of the original fuel. Simulations were carried out to calculate prompt neutron generation time, effective delayed-neutron fraction, core excess reactivity and neutron flux spectrum. Nuclear reactor analysis codes including WIMS-D4 and CITATION were used to carry out these calculations. It was observed that both the silicide fuels had the same prompt neutron generation time 0.02% more than that of the original aluminide fuel, while the oxide fuel had a prompt neutron generation time 0.05% less than that of the original aluminide fuel. The effective delayed-neutron fraction decreased for all the fuels; the decrease was maximum at 0.06% for U 3 Si 2 -Al followed by 0.03% for U 3 Si-Al, and 0.01% for U 3 O 8 -Al fuel. The U 3 O 8 -Al fueled reactor gave the maximum ρ excess at BOL which was 21.67% more than the original fuel followed by U 3 Si-Al which was 2.55% more, while that of U 3 Si 2 -Al was 2.50% more than the original UAl x -Al fuel. The neutron flux of all the fuels was more thermalized, than in the original fuel, in the active fuel region of the core. The thermalization was maximum for U 3 O 8 -Al followed by U 3 Si-Al and then U 3 Si 2 -Al fuel.

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

Small Angle Neutron Scattering instrument at Malaysian TRIGA reactor

Energy Technology Data Exchange (ETDEWEB)

Mohd, Shukri; Kassim, Razali; Mahmood, Zal Uyun [Malaysian Inst. for Nuclear Technology Research (MINT), Bangi, Kajang (Malaysia); Radiman, Shahidan

1998-10-01

The TRIGA MARK II Research reactor at the Malaysian Institute for Nuclear Research (MINT) was commissioned in July 1982. Since then various works have been performed to utilise the neutrons produced from this steady state reactor. One of the project involved the Small Angle Neutron Scattering (SANS). (author)

Analysis of neutronics and dynamic characteristics with reactivity injection in LBE cooled sub-critical reactor

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)

Application of a two-region kinetic model for reflected reactors to experimental data

International Nuclear Information System (INIS)

Busch, R.D.; Spriggs, G.D.; Williams, J.G.

1996-01-01

Reflected reactors constitute one of the most important classes of nuclear reactors. Yet, during the past 50 yr, a plethora of experimental data involving reflected systems has been reported in the literature that cannot be satisfactorily explained using the open-quotes standardclose quotes (i.e., one-region) point-kinetic model. In particular, many have observed that the prompt-decay a curves obtained from Rossi-α and pulsed-neutron experiments can exhibit multiple decay modes in the vicinity near delayed critical in some types of reflected systems. When analyzed using theories based on the standard point-kinetic model, these data yielded system lifetimes that do not always agree well with the lifetimes predicted by numerical solutions of the multigroup, multidimensional diffusion or transport equations. In several cases, when the longest lived decay mode (i.e., the dominant root) was plotted as a function of reactivity, the a curve intercepted the reactivity axis at a reactivity significantly greater than 1$. Brunson dubbed this seemingly inexplicable behavior as the open-quotes dollar discrepancy.close quotes Furthermore, it has also been observed that the kinetic behavior of some reflected, fast-burst assemblies exhibits a very pronounced nonlinear relationship between reactivity and the initial inverse period for reactivity insertions > 1 $

EPRTM Reactor neutron instrumentation

International Nuclear Information System (INIS)

Pfeiffer, Maxime; SALA, Stephanie

2013-06-01

The core safety during operation is linked, in particular, to the respect of criteria related to the heat generated in fuel rods and to the heat exchange between the rods and the coolant. This local power information is linked to the power distribution in the core. In order to evaluate the core power distribution, the EPR TM reactor relies on several types of neutron detectors: - ionization chambers located outside the vessel and used for protection and monitoring - a fixed in-core instrumentation based on Cobalt Self Powered Neutron Detectors used for protection and monitoring - a mobile reference in-core instrumentation based on Vanadium aero-balls This document provides a description of this instrumentation and its use in core protection, limitation, monitoring and control functions. In particular, a description of the detectors and the principles of their signal generation is supplied as well as the description of the treatments related to these detectors in the EPR TM reactor I and C systems (including periodical calibration). (authors)

Personal neutron dosimetry at a research reactor facility

International Nuclear Information System (INIS)

Kamenopoulou, V.; Carinou, E.; Stamatelatos, I.E.

2001-01-01

Individual neutron monitoring presents several difficulties due to the differences in energy response of the dosemeters. In the present study, an individual dosemeter (TLD) calibration approach is attempted for the personnel of a research reactor facility. The neutron energy response function of the dosemeter was derived using the MCNP code. The results were verified by measurements to three different neutron spectra and were found to be in good agreement. Three different calibration curves were defined for thermal, intermediate and fast neutrons. At the different working positions around the reactor, neutron spectra were defined using the Monte Carlo technique and ambient dose rate measurements were performed. An estimation of the neutrons energy is provided by the ratio of the different TLD pellets of each dosemeter in combination with the information concerning the worker's position; then the dose equivalent is deduced according to the appropriate calibration curve. (author)

Neutrons down-under: Australia's research reactor review

International Nuclear Information System (INIS)

Murray, Allan

1995-01-01

Australian research reactor review commenced in September 1992, the Review had the following Terms of Reference: Whether, on review of the benefits and costs for scientific, commercial, industrial and national interest reasons, Australia has a need for a new reactor; a review of the present reactor, HIFAR, to include: an assessment of national and commercial benefits and costs of operations, its likely remaining useful life and its eventual closure and decommissioning; if Australia has a need for a new nuclear research reactor, the Review will consider: possible locations for a new reactor, its environmental impact at alternative locations, recommend a preferred location, and evaluate matters associated with regulation of the facility and organisational arrangements for reactor-based research. From the Review findings the following recommendations were stated: keep HIFAR going; commission a PRA to ascertain HIFAR's remaining life and refurbishment possibilities; identify and establish a HLW repository; accept that neither HIFAR nor a new reactor can be completely commercial; any decision on a new neutron source must rest primarily on benefits to science and Australia's national interest; make a decision on a new neutron source in about five years' time (1998). Design Proposals for a New Reactor are specified

Application of reactors for testing neutron-induced upsets in commercial SRAMs

International Nuclear Information System (INIS)

Griffin, P.J.; Luera, T.F.; Sexton, F.W.; Cooper, P.J.; Karr, S.G.; Hash, G.L.; Fuller, E.

1997-01-01

Reactor neutron environments can be used to test/screen the sensitivity of unhardened commercial SRAMs to low-LET neutron-induced upset. Tests indicate both thermal/epithermal (< 1 keV) and fast neutrons can cause upsets in unhardened parts. Measured upset rates in reactor environments can be used to model the upset rate for arbitrary neutron spectra

Rod drop in the LR-0 reactor core comprising 55 fuel assemblies

International Nuclear Information System (INIS)

Hadek, J.; Grundmann, U.

1989-09-01

Data from the third stage of kinetic measurements on the LR-0 reactor, performed in 1988, were employed for additional calculations using the 3-dimensional neutron kinetics code HEXDYN3D. The reactor consists of subassemblies similar to those in the WWER-1000 (PWR) reactor. The theoretical and experimental results are compared for the time behavior of the neutron flux caused by drop of the control rod cluster in various subassemblies of the reactor. The results demonstrate that the HEXDYN3D code is well suited to the treatment of the space-time behavior of the neutron flux. (author). 21 figs., 2 tabs., 16 refs

Neutron mean annihilation time and inverse of the mean annihilation rate in nuclear reactor

International Nuclear Information System (INIS)

Hayashi, Masatoshi

1999-01-01

There is a dogma in nuclear reactor theory that in a critical reactor the mean annihilation time of neutron is equal to the mean generation time. The author insists that this is a dogma from the basic reexamination of the mean annihilation time of neutron. There are two kinds of neutrons, one participating in chain reactions and the other not participating in chain reactions. The mean annihilation time of neutron is the mean time of the time to annihilation of all neutrons generated in the reactor. The 'prompt neutron life' as a dynamic characteristic parameter proper to nuclear reactor can not be understood as the mean time of neutron to annihilation. The author explains the logic quantitatively with two kinds of nuclear reactors, a bare reactor and an infinite reactor, for which two different mean neutron annihilation times can be defined. Thus, (1) the inverse of the annihilation rate can not simply be considered as the mean annihilation time, (2) the mean annihilation time of a critical reactor is not necessarily equal to the mean generation time, and (3) the prompt neutron life used as a dynamic characteristic parameter of a nuclear reactor can not be understood as the mean time of neutron to annihilation. (M.M.)

Neutron spectrometry for reactor applications: status, limitations, and future directions

International Nuclear Information System (INIS)

Gold, R.

1975-08-01

The ability of ''state-of-the-art'' reactor neutron spectrometry to provide definitive environmental results required for high fluence radiation damage experiments is reviewed. A formal definition of the neutron component is presented as well as general considerations which accrue from both this definition and the existence of the mixed radiation field generally encountered in reactors. A description of four selected methods of reactor neutron spectrometry is included, namely Proton Recoil (PR) methods, Time-Of-Flight (TOF) methods, the 6 Li(n,α) 3 H coincidence method, and Multiple Foil Activation (MFA) methods. These selected methods are compared. Future requirements and directions for reactor neutron spectrometry are discussed. In particular, the needs of future CTR research are stressed and the He 4 - recoil proportional counter spectroscopy method is advanced as a means of meeting these future requirements. 50 references. (auth)

Conversion of reactor neutrons by lithium deuteride; Konverzija reaktorskih neutrona pomocu litijumdeuterida

Energy Technology Data Exchange (ETDEWEB)

Strugar, P; Altiparmakov, D [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

1979-07-01

Nuclear reactors are powerful neutron sources. But for many purposes, neutrons of higher than fission energy are needed. Such neutrons may be produced by the conversion of reactor neutrons into 14 MeV neutrons using lithium deuteride converter. For that converter is generally employed {sup 6}LiD, the usual material for the thermonuclear weapons, and therefore hardly accessible in needed quantity. That was the reason to analyse such converter made of LiD with the lithium of natural isotopic content. This analysis starts with the basic conversion relations, takes into account neutron absorption and tritium generation, and finally, estimates the 14 MeV neutron flux and the heat generated in proposed converter, when the converter was coupled to each of three Yugoslav nuclear reactors. Results show that the converter made of LiD with the natural lithium is 50% less efficient than the converter of {sup 6}LiD. Intensity of 14 MeV neutrons is within limits 5. 10{sup 5} - 10{sup 10} (n/cm{sup 2}.s) for the converter used either as external converter with reactor RB, within the thermal column in the reactor TRIGA or as a 'fuel' segment at the reactor RA. (author)

Absorbing rods for nuclear fast neutron reactor absorbing assembly

International Nuclear Information System (INIS)

Aji, M.; Ballagny, A.; Haze, R.

1986-01-01

The invention proposes a neutron absorber rod for neutron absorber assembly of a fast neutron reactor. The assembly comprises a bundle of vertical rods, each one comprising a stack of pellets made of a neutron absorber material contained in a long metallic casing with a certain radial play with regard to this casing; this casing includes traps for splinters from the pellets which may appear during reactor operation, at the level of contact between adjacent pellets. The present invention prevents the casing from rupture involved by the disintegration of the pellets producing pieces of boron carbide of high hardness [fr

Solving the uncommon reactor core neutronics problems

International Nuclear Information System (INIS)

Vondy, D.R.; Fowler, T.B.

1980-01-01

The common reactor core neutronics problems have fundamental neutron space, energy spectrum solutions. Typically the most positive eigenvalue is associated with an all-positive flux for the pseudo-steady-state condition (k/sub eff/), or the critical state is to be effected by selective adjustment of some variable such as the fuel concentration. With sophistication in reactor analysis has come the demand for solutions of other, uncommon neutronics problems. Importance functionss are needed for sensitivity and uncertainty analyses, as for ratios of intergral reaction rates such as the fuel conversion (breeding) ratio. The dominant higher harmonic solution is needed in stability analysis. Typically the desired neutronics solution must contain negative values to qualify as a higher harmonic or to satisfy a fixed source containing negative values. Both regular and adjoint solutions are of interest as are special integrals of the solutions to support analysis

Device for detecting neutron flux in nuclear reactor. [BWR

Energy Technology Data Exchange (ETDEWEB)

Bessho, Y; Nishizawa, Y

1976-07-30

The object of the invention is to ensure accuracy in the operation of the nuclear reactor by reducing the difference that results between the readings of a Traversing Incore Probe (TIP) and a Local Power Range Monitor (LPRM) when the neutron flux distribution undergoes a change. In an apparatus for detecting neutrons in a nuclear reactor, an LPRM sensor comprising a layer containing a substance capable of nuclear fission, a section filled with argon gas and a collector is constructed so as to surround a TIP within a TIP guide tube at the height of the reactor axis. In this way, the LPRM detects the average value of neutron distribution in the region surrounding the TIP, so that no great difference between the readings of both the sensors is produced even if the neutron flux distribution is changed.

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)

Reflected kinetics model for nuclear space reactor kinetics and control scoping calculations

Energy Technology Data Exchange (ETDEWEB)

Washington, K.E.

1986-05-01

The objective of this research is to develop a model that offers an alternative to the point kinetics (PK) modelling approach in the analysis of space reactor kinetics and control studies. Modelling effort will focus on the explicit treatment of control drums as reactivity input devices so that the transition to automatic control can be smoothly done. The proposed model is developed for the specific integration of automatic control and the solution of the servo mechanism problem. The integration of the kinetics model with an automatic controller will provide a useful tool for performing space reactor scoping studies for different designs and configurations. Such a tool should prove to be invaluable in the design phase of a space nuclear system from the point of view of kinetics and control limitations.

Reflected kinetics model for nuclear space reactor kinetics and control scoping calculations

International Nuclear Information System (INIS)

Washington, K.E.

1986-05-01

The objective of this research is to develop a model that offers an alternative to the point kinetics (PK) modelling approach in the analysis of space reactor kinetics and control studies. Modelling effort will focus on the explicit treatment of control drums as reactivity input devices so that the transition to automatic control can be smoothly done. The proposed model is developed for the specific integration of automatic control and the solution of the servo mechanism problem. The integration of the kinetics model with an automatic controller will provide a useful tool for performing space reactor scoping studies for different designs and configurations. Such a tool should prove to be invaluable in the design phase of a space nuclear system from the point of view of kinetics and control limitations

Simulating Neutronic Core Parameters in a Research and Test Reactor

International Nuclear Information System (INIS)

Selim, H.K.; Amin, E.A.; Koutb, M.E.

2011-01-01

The present study proposes an Artificial Neural Network (ANN) modeling technique that predicts the control rods positions in a nuclear research reactor. The neutron, flux in the core of the reactor is used as the training data for the neural network model. The data used to train and validate the network are obtained by modeling the reactor core with the neutronic calculation code: CITVAP. The type of the network used in this study is the feed forward multilayer neural network with the backpropagation algorithm. The results show that the proposed ANN has good generalization capability to estimate the control rods positions knowing neutron flux for a research and test reactor. This method can be used to predict critical control rods positions to be used for reactor operation after reload

The development of ex-core neutron flux monitoring system for integral reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, J. K.; Kwon, H. J.; Park, H. Y.; Koo, I. S

2004-12-01

Due to the arrangement of major components within the reactor vessel, the integral reactor has relatively long distance between the core support barrel and the reactor vessel when compared with the currently operating plants. So, a neutron flux leakage at the ex-vessel represents a relatively low flux level which may generate some difficulties in obtaining a wide range of neutron flux information including the source range one. This fact may have an impact upon the design and fabrication of an ex-core neutron flux detector. Therefore, it is required to study neutron flux detectors that are suitable for the installation location and characteristics of an integral reactor. The physical constraints of an integral reactor should be considered when one designs and develops the ex-core neutron flux monitoring detectors and their systems. As a possible installation location of the integral reactor ex-core neutron flux detector assembly, two candidate locations are considered, that is, one is between the core support barrel and the reactor vessel and the other is within the Internal Shielding Tank(IST). And, for these locations, some factors such as the environmental requirements and geometrical restrictions are investigated In the case of considering the inside of the IST as a ex-core neutron flux detector installation position, an electrical insulation problem and a low neutron flux measurement problem arose and when considering the inside of the reactor vessel, a detector's sensitivity variation problem, an electrical insulation problem, a detector's insertion and withdrawal problem, and a high neutron flux measurement problem were encountered. Through a survey of the detector installation of the currently operating plants and detector manufacturer's products, the proposed structure and specifications of an ex-core neutron flux detector are suggested. And, the joint ownership strategy for a proposed detector model is also depicted. At the end, by studying

The development of ex-core neutron flux monitoring system for integral reactor

International Nuclear Information System (INIS)

Lee, J. K.; Kwon, H. J.; Park, H. Y.; Koo, I. S.

2004-12-01

Due to the arrangement of major components within the reactor vessel, the integral reactor has relatively long distance between the core support barrel and the reactor vessel when compared with the currently operating plants. So, a neutron flux leakage at the ex-vessel represents a relatively low flux level which may generate some difficulties in obtaining a wide range of neutron flux information including the source range one. This fact may have an impact upon the design and fabrication of an ex-core neutron flux detector. Therefore, it is required to study neutron flux detectors that are suitable for the installation location and characteristics of an integral reactor. The physical constraints of an integral reactor should be considered when one designs and develops the ex-core neutron flux monitoring detectors and their systems. As a possible installation location of the integral reactor ex-core neutron flux detector assembly, two candidate locations are considered, that is, one is between the core support barrel and the reactor vessel and the other is within the Internal Shielding Tank(IST). And, for these locations, some factors such as the environmental requirements and geometrical restrictions are investigated In the case of considering the inside of the IST as a ex-core neutron flux detector installation position, an electrical insulation problem and a low neutron flux measurement problem arose and when considering the inside of the reactor vessel, a detector's sensitivity variation problem, an electrical insulation problem, a detector's insertion and withdrawal problem, and a high neutron flux measurement problem were encountered. Through a survey of the detector installation of the currently operating plants and detector manufacturer's products, the proposed structure and specifications of an ex-core neutron flux detector are suggested. And, the joint ownership strategy for a proposed detector model is also depicted. At the end, by studying the ex

A neutron radiography facility on the IRT-2000 reactor

International Nuclear Information System (INIS)

Khadduri, I.Y.

1976-01-01

A neutron radiography facility has been constructed on the thermal neutron channel of the IRT-2000 reactor. A collimated thermal neutron beam exposure area of 10 cm diameter is obtained with an L/D ratio of 48.8. The film used is cellulose nitrate coated with lithium tetraborate which is insensitive to gamma and beta radiation. Some pictures with good contrast and resolution have been obtained. Pictures of parts of an IRT-2000 reactor fuel pin have also been recorded. (orig) [de

An ultracold neutron source at the NC State University PULSTAR reactor

Science.gov (United States)

Korobkina, E.; Wehring, B. W.; Hawari, A. I.; Young, A. R.; Huffman, P. R.; Golub, R.; Xu, Y.; Palmquist, G.

2007-08-01

Research and development is being completed for an ultracold neutron (UCN) source to be installed at the PULSTAR reactor on the campus of North Carolina State University (NCSU). The objective is to establish a university-based UCN facility with sufficient UCN intensity to allow world-class fundamental and applied research with UCN. To maximize the UCN yield, a solid ortho-D 2 converter will be implemented coupled to two moderators, D 2O at room temperature, to thermalize reactor neutrons, and solid CH 4, to moderate the thermal neutrons to cold-neutron energies. The source assembly will be located in a tank of D 2O in the space previously occupied by the thermal column of the PULSTAR reactor. Neutrons leaving a bare face of the reactor core enter the D 2O tank through a 45×45 cm cross-sectional area void between the reactor core and the D 2O tank. Liquid He will cool the disk-shaped UCN converter to below 5 K. Independently, He gas will cool the cup-shaped CH 4 cold-neutron moderator to an optimum temperature between 20 and 40 K. The UCN will be transported from the converter to experiments by a guide with an inside diameter of 16 cm. Research areas being considered for the PULSTAR UCN source include time-reversal violation in neutron beta decay, neutron lifetime determination, support measurements for a neutron electric-dipole-moment search, and nanoscience applications.

Materials research with neutron beams from a research reactor

Energy Technology Data Exchange (ETDEWEB)

Root, J.; Banks, D. [Canadian Neutron Beam Centre, Chalk River Laboratories, Chalk River, Ontario (Canada)

2015-03-15

Because of the unique ways that neutrons interact with matter, neutron beams from a research reactor can reveal knowledge about materials that cannot be obtained as easily with other scientific methods. Neutron beams are suitable for imaging methods (radiography or tomography), for scattering methods (diffraction, spectroscopy, and reflectometry) and for other possibilities. Neutron-beam methods are applied by students and researchers from academia, industry and government to support their materials research programs in several disciplines: physics, chemistry, materials science and life science. The arising knowledge about materials has been applied to advance technologies that appear in everyday life: transportation, communication, energy, environment and health. This paper illustrates the broad spectrum of materials research with neutron beams, by presenting examples from the Canadian Neutron Beam Centre at the NRU research reactor in Chalk River. (author)

Neutronic analysis for conversion of the Ghana Research Reactor-1 facility using Monte Carlo methods and UO{sub 2} LEU fuel

Energy Technology Data Exchange (ETDEWEB)

Anim-Sampong, S.; Akaho, E.H.K.; Maakuu, B.T.; Gbadago, J.K. [Ghana Research Reactor-1 Centre, Dept. of Nuclear Engineering and Materials Science, National Nuclear Research Institute, Ghana Atomic Energy Commission, Legon, Accra (Ghana); Andam, A. [Kwame Nkrumah Univ. of Science and Technology, Dept. of Physics (Ghana); Liaw, J.J.R.; Matos, J.E. [Argonne National Lab., RERTR Programme, Div. of Nuclear Engineering (United States)

2007-07-01

Monte Carlo particle transport methods and software (MCNP) have been applied to the modelling, simulation and neutronic analysis for the conversion of the HEU-fuelled (high enrichment uranium) core of the Ghana Research Reactor-1 (GHARR-1) facility. The results show that the MCNP model of the GHARR-1 facility, which is a commercial version of the Miniature Neutron Source Reactor (MNSR) is good as the simulated neutronic and other reactor physics parameters agree with very well with experimental and zero power results. Three UO{sub 2} LEU (low enrichment uranium) fuels with different enrichments (12.6% and 19.75%), core configurations, core loadings were utilized in the conversion studies. The nuclear criticality and kinetic parameters obtained from the Monte Carlo simulation and neutronic analysis using three UO{sub 2} LEU fuels are in close agreement with results obtained for the reference 90.2% U-Al HEU core. The neutron flux variation in the core, fission chamber and irradiation channels for the LEU UO{sub 2} fuels show the same trend as the HEU core as presented in the paper. The Monte Carlo model confirms a reduction (8% max) in the peak neutron fluxes simulated in the irradiation channels which are utilized for experimental and commercial activities. However, the reductions or 'losses' in the flux levels neither affects the criticality safety, reactor operations and safety nor utilization of the reactor. Employing careful core loading optimization techniques and fuel loadings and enrichment, it is possible to eliminate the apparent reductions or 'losses' in the neutron fluxes as suggested in this paper. Concerning neutronics, it can be concluded that all the 3 LEU fuels qualify as LEU candidates for core conversion of the GHARR-1 facility.

Generalized treatment of point reactor kinetics driven by random reactivity fluctuations via the Wiener-Hermite functional method

International Nuclear Information System (INIS)

Behringer, K.

1991-02-01

In a recent paper by Behringer et al. (1990), the Wiener-Hermite Functional (WHF) method has been applied to point reactor kinetics excited by Gaussian random reactivity noise under stationary conditions, in order to calculate the neutron steady-state value and the neutron power spectral density (PSD) in a second-order (WHF-2) approximation. For simplicity, delayed neutrons and any feedback effects have been disregarded. The present study is a straightforward continuation of the previous one, treating the problem more generally by including any number of delayed neutron groups. For the case of white reactivity noise, the accuracy of the approach is determined by comparison with the exact solution available from the Fokker-Planck method. In the numerical comparisons, the first-oder (WHF-1) approximation of the PSD is also considered. (author) 4 figs., 10 refs

Neutron dosimetry. Environmental monitoring in a BWR type reactor

International Nuclear Information System (INIS)

Tavera D, L.; Camacho L, M.E.

1991-01-01

The measurements carried out on reactor dosimetry are applied mainly to the study on the effects of the radiation in 108 materials of the reactor; little is on the environmental dosimetry outside of the primary container of BWR reactors. In this work the application of a neutron spectrometer formed by plastic detectors of nuclear traces manufactured in the ININ, for the environmental monitoring in penetrations around the primary container of the unit I of the Laguna Verde central is presented. The neutron monitoring carries out with purposes of radiological protection, during the operational tests of the reactor. (Author)

Design study of a medical reactor for Boron Neutron Capture Therapy

International Nuclear Information System (INIS)

Sasaki, M.; Hirota, J.; Tamao, S.; Kanda, K.; Mishima, Y.

1992-01-01

A new design study of a medical reactor for Boron Neutron Capture Therapy (BNCT) has been carried out. The reactor is to be used exclusively for the treatment of malignant melanoma and other cancers as well as for the further biomedical research. Main specifications of the reactor are as follows; thermal power of 2 MW, water cooling by natural convection, semitight core of triangular lattice, UO 2 fuel rod of 9.5 mm diameter and no refueling in the reactor-life. Three horizontal and one vertical neutron beam hole are to be provided to deliver thermal and epithermal neutrons. N-γ coupling Sn transport calculations indicate that the patient treatment period will be about 30 minutes with minimal fast neutron and gamma contaminants. (author)

Burnup influence on the VVER-1000 reactor vessel neutron fluence evaluation

International Nuclear Information System (INIS)

Panayotov, I.; Mihaylov, N.; Ilieva, K.; Kirilova, D.; Manolova, M.

2009-01-01

The neutron fluence of the vessels of the reactors is determined regularly accordingly the RPV Surveillance Program of the Kozloduy NPP Unit 5 and 6 in order to assess the state of the metal vessel and their radiation damaging. The calculations are carried out by the method of discrete ordinates used in the TORT program for operated reactor cycles. An average reactor spectrum corresponding to fresh U-235 fuel is used as an input neutron source. The impact of the burn up of the fuel on the neutron fluence of VVER-1000 reactor vessel is evaluated. The calculations of isotopic concentrations of U-235 and Pu-239 corresponding to 4 years burn up were performed by the module SAS2H of the code system SCALE 4.4. Since fresh fuel or 4 years burn up fuel assembly are placed in periphery of reactor core the contribution of Pu-239 of first year burn up and of 4 years burn up is taken in consideration. Calculations of neutron fluence were performed with neutron spectrum for fresh fuel, for 1 year and for 4 years burn up fuel. Correction factors for neutron fluence at the inner surface of the reactor vessel, in 1/4 depth of the vessel and in the air behind the vessel were obtained. The correction coefficient could be used when the neutron fluence is assessed so in verification when the measured activity of ex-vessel detectors is compared with calculated ones. (authors)

Burnup influence on the WWER-1000 reactor vessel neutron fluence evaluation

International Nuclear Information System (INIS)

Panayotov, I.; Mihaylov, N.; Ilieva, K.; Kirilova, D.; Manolova, M.

2009-01-01

The neutron fluence of the vessels of the reactors is determined regularly accordingly the RPV Surveillance Program of Kozloduy NPP Unit 5 and 6 in order to assess the state of the metal vessel and their radiation damaging. The calculations are carried out by the method of discrete ordinates used in the TORT program for operated reactor cycles. An average reactor spectrum corresponding to fresh U-235 fuel is used as an input neutron source. The impact of the burn up of the fuel on the neutron fluence of WWER-1000 reactor vessel is evaluated. The calculations of isotopic concentrations of U-235 and Pu-239 corresponding to 4 years burn up were performed by the module SAS2H of the code system SCALE 4.4. Since fresh fuel or 4 years burn up fuel assembly are placed in periphery of reactor core the contribution of Pu-239 of first year burn up and of 4 years burn up is taken in consideration. Calculations of neutron fluence were performed with neutron spectrum for fresh fuel, for 1 year and for 4 years burn up fuel. Correction factors for neutron fluence at the inner surface of the reactor vessel, in ? depth of the vessel and in the air behind the vessel were obtained. The correction coefficient could be used when the neutron fluence is assessed so in verification when the measured activity of ex-vessel detectors is compared with calculated ones. (Authors)

Lifetime measurement of prompt neutrons using the neutronic noise analysis

International Nuclear Information System (INIS)

Ortiz Servin, J.J.

1992-01-01

The purpose of this work is to estimate the life of the prompt neutrons, i, of a nuclear reactor utilizing the neutron noise analysis. This technique carry to development of mathematical model that is valid for lower powers reactor. The equation resulting convey to the observation about power spectrum behaviour respect to the frecquency. In this case, the reactor in study is the Triga Mark III of Nuclear Center of Mexico that it was provided of fission chambers for register the neutron fluxes. These fluxes was digitized and storage in computer disc as signals dependents of time, for later apply the Fourier Transformation and obtain the spectras. The spectras measured to different reactor powers were adjusted to the development equation before, using the method of square minimum and so estimate the parameter i. The analysis of results throw a value of 22.73 +/- 0.92 μs. On the other hand, the calculate value to the resolve the kinetic equation of reactor defer in lower than 4 % about the estimate. Of this, it concludes that the model utilized is trusty with a good mistake margin, moreover of that the technique of Neutron Noise analysis demonstrate be competitive (Author)

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

Energy Technology Data Exchange (ETDEWEB)

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

1998-03-01

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

Analysis of calculated neutron flux response at detectors of G.A. Siwabessy multipurpose reactor (RSG-GAS Reactor)

International Nuclear Information System (INIS)

Taryo, Taswanda

2002-01-01

Multi Purpose Reactor G.A. Siwabessy (RSG-GAS) reactor core possesses 4 fission-chamber detectors to measure intermediate power level of RSG-GAS reactor. Another detector, also fission-chamber detector, is intended to measure power level of RSG-GAS reactor. To investigate influence of space to the neutron flux values for each detector measuring intermediate and power levels has been carried out. The calculation was carried out using combination of WIMS/D4 and CITATION-3D code and focused on calculation of neutron flux at different detector location of RSG-GAS typical working core various scenarios. For different scenarios, all calculation results showed that each detector, located at different location in the RSG-GAS reactor core, causes different neutron flux occurred in the reactor core due to spatial time effect

Thermal neutron filter design for the neutron radiography facility at the LVR-15 reactor

Energy Technology Data Exchange (ETDEWEB)

Soltes, Jaroslav [Research Centre Rez Ltd., Husinec - Rez 130, 250 68 Rez, (Czech Republic); Faculty of Nuclear Sciences and Physical Engineering, CTU in Prague, (Czech Republic); Viererbl, Ladislav; Lahodova, Zdena; Koleska, Michal; Vins, Miroslav [Research Centre Rez Ltd., Husinec - Rez 130, 250 68 Rez, (Czech Republic)

2015-07-01

In 2011 a decision was made to build a neutron radiography facility at one of the unused horizontal channels of the LVR-15 research reactor in Rez, Czech Republic. One of the key conditions for operating an effective radiography facility is the delivery of a high intensity, homogeneous and collimated thermal neutron beam at the sample location. Additionally the intensity of fast neutrons has to be kept as low as possible as the fast neutrons may damage the detectors used for neutron imaging. As the spectrum in the empty horizontal channel roughly copies the spectrum in the reactor core, which has a high ratio of fast neutrons, neutron filter components have to be installed inside the channel in order to achieve desired beam parameters. As the channel design does not allow the instalment of complex filters and collimators, an optimal solution represent neutron filters made of large single-crystal ingots of proper material composition. Single-crystal silicon was chosen as a favorable filter material for its wide availability in sufficient dimensions. Besides its ability to reasonably lower the ratio of fast neutrons while still keeping high intensities of thermal neutrons, due to its large dimensions, it suits as a shielding against gamma radiation from the reactor core. For designing the necessary filter dimensions the Monte-Carlo MCNP transport code was used. As the code does not provide neutron cross-section libraries for thermal neutron transport through single-crystalline silicon, these had to be created by approximating the theory of thermal neutron scattering and modifying the original cross-section data which are provided with the code. Carrying out a series of calculations the filter thickness of 1 m proved good for gaining a beam with desired parameters and a low gamma background. After mounting the filter inside the channel several measurements of the neutron field were realized at the beam exit. The results have justified the expected calculated values

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)

FX2-TH: a two-dimensional nuclear reactor kinetics code with thermal-hydraulic feedback

International Nuclear Information System (INIS)

Shober, R.A.; Daly, T.A.; Ferguson, D.R.

1978-10-01

FX2-TH is a two-dimensional, time-dependent nuclear reactor kinetics program with thermal and hydraulic feedback. The neutronics model used is multigroup neutron diffusion theory. The following geometry options are available: x, r, x-y, r-z, theta-r, and triangular. FX2-TH contains two basic thermal and hydraulic models: a simple adiabatic fuel temperature calculation, and a more detailed model consisting of an explicit representation of a fuel pin, gap, clad, and coolant. FX2-TH allows feedback effects from both fuel temperature (Doppler) and coolant temperature (density) changes. FX2-TH will calculate a consistent set of steady state conditions by iterating between the neutronics and thermal-hydraulics until convergence is reached. The time-dependent calculation is performed by the use of the improved quasistatic method. A disk editing capability is available. FX2-TH is operational on IBM system 360 or 370 computers and on the CDC 7600

Extension of the AUS reactor neutronics system for application to fusion blanket neutronics

International Nuclear Information System (INIS)

Robinson, G.S.

1984-03-01

The AUS modular code scheme for reactor neutronics computations has been extended to apply to fusion blanket neutronics. A new group cross-section library with 200 neutron groups, 37 photon groups and kerma factor data has been generated from ENDF/B-IV. The library includes neutron resonance subgroup parameters and temperature-dependent data for thermal neutron scattering matrices. The validity of the overall calculation system for fusion applications has been checked by comparison with a number of published conceptual design studies

The implication of sensitivity analysis on the safety and delayed-neutron parameters for fast breeder reactors

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)

AIREKMOD-RR, Reactivity Transients in Nuclear Research Reactors

International Nuclear Information System (INIS)

Baggoura, B.; Mazrou, H.

2001-01-01

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

Experimental determination of the neutron source for the Argonauta reactor subcritical assembly

Energy Technology Data Exchange (ETDEWEB)

Renke, Carlos A.C.; Furieri, Rosanne C.A.A.; Pereira, Joao C.S.; Voi, Dante L.; Barbosa, Andre L.N., E-mail: renke@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

2011-07-01

The utilization of a subcritical assembly for the determination of nuclear parameters in a multiplier medium requires a well defined neutron source to carry out the experiments necessary for the acquisition of the desired data. The Argonauta research reactor installed at the Instituto de Engenharia Nuclear has a subcritical assembly, under development, to be coupled at the upper part of the reactor core that will provide the needed neutrons emerging from its internal thermal column made of graphite. In order to perform neutronic calculations to compare with the experimental results, it is necessary a precise knowledge of the emergent neutron flux that will be used as neutron source in the subcritical assembly. In this work, we present the thermal neutron flux profile determined experimentally via the technique of neutron activation analysis, using dysprosium wires uniformly distributed at the top of the internal thermal neutron column of the Argonauta reactor and later submitted to a detection system using Geiger-Mueller detector. These experimental data were then compared with those obtained through neutronic calculation using HAMMER and CITATION codes in order to validate this calculation system and to define a correct neutron source distribution to be used in the subcritical assembly. This procedure avoids a coupled neutronic calculation of the subcritical assembly and the reactor core. It has also been determined the dimension of the graphite pedestal to be used in the bottom of the subcritical assembly tank in order to smooth the emergent neutron flux at the reactor top. Finally, it is estimated the thermal neutron flux inside the assembly tank when filled with water. (author)

Conceptual design of neutron diagnostic systems for fusion experimental reactor

International Nuclear Information System (INIS)

Iguchi, T.; Kaneko, J.; Nakazawa, M.

1994-01-01

Neutron measurement in fusion experimental reactors is very important for burning plasma diagnostics and control, monitoring of irradiation effects on device components, neutron source characterization for in-situ engineering tests, etc. A conceptual design of neutron diagnostic systems for an ITER-like fusion experimental reactor has been made, which consists of a neutron yield monitor, a neutron emission profile monitor and a 14-MeV spectrometer. Each of them is based on a unique idea to meet the required performances for full power conditions assumed at ITER operation. Micro-fission chambers of 235 U (and 238 U) placed at several poloidal angles near the first wall are adopted as a promising neutron yield monitor. A collimated long counter system using a 235 U fission chamber and graphite neutron moderators is also proposed to improve the calibration accuracy of absolute neutron yield determination

Performance of non-conventional factorization approaches for neutron kinetics

International Nuclear Information System (INIS)

Bulla, S.; Nervo, M.

2013-01-01

The use of factorization techniques provides a interesting option for the simulation of the time-dependent behavior of nuclear systems with a reduced computational effort. While point kinetics neglects all spatial and spectral effects, quasi-statics and multipoint kinetics allow to produce results with a higher accuracy for transients involving relevant modifications of the neutron distribution. However, in some conditions these methods can not work efficiently. In this paper, we discuss some possible alternative formulations for the factorization process for neutron kinetics, leading to mathematical models of reduced complications that can allow an accurate simulation of transients involving spatial and spectral effects. The performance of these innovative approaches are compared to standard techniques for some test cases, showing the benefits and shortcomings of the method proposed. (authors)

Reference Neutron Radiographs of Nuclear Reactor Fuel

DEFF Research Database (Denmark)

Domanus, Joseph Czeslaw

1986-01-01

Reference neutron radiographs of nuclear reactor fuel were produced by the Euraton Neutron Radiography Working Group and published in 1984 by the Reidel Publishing Company. In this collection a classification is given of the various neutron radiographic findings, that can occur in different parts...... of pelletized, annular and vibro-conpacted nuclear fuel pins. Those parts of the pins are shown where changes of appearance differ from those for the parts as fabricated. Also radiographs of those as fabricated parts are included. The collection contains 158 neutron radiographs, reproduced on photographic paper...

Conversion of the RB reactor neutrons by highly enriched uranium fuel and lithium deuteride

International Nuclear Information System (INIS)

Strugar, P.; Sotic, O.; Ninkovic, M.; Pesic, M.; Altiparmakov, D.

1981-01-01

A thermal-to-fast-neutron converter has been constructed at the RB reactor. The material used for the conversion of thermal neutrons is highly enriched uranium fuel of Soviet production applied in Yugoslav heavy water experimental reactors RA and RB. Calculations and preliminary measurements show that the spectrum of converted neutrons only slightly differs from that of fission neutrons. The basic characteristics of converted neutrons can be expressed by the neutron radiation dose of 800 rad (8 Gy) for 1 h of reactor operation at a power level of 1 kW. This dose is approximately 10 times higher than the neutron dose at the same place without converter. At the same time, thermal neutron and gamma radiation doses are negligible. The constructed neutron converter offers wide possibilities for applications in reactor and nuclear physics and similar disciplines, where neutron spectra of high energies are required, as well as in the domain of neutron dosimetry and biological irradiations in homogeneous fields of larger dimensions. The possibility of converting thermal reactor neutrons with energies of about 14 MeV with the aid of lithium deuteride from natural lithium has been considered too. (author)

Neutron irradiation facilities for fission and fusion reactor materials studies

International Nuclear Information System (INIS)

Rowcliffe, A.F.

1985-01-01

The successful development of energy-conversion machines based upon nuclear fission or fusion reactors is critically dependent upon the behavior of the engineering materials used to construct the full containment and primary heat extraction systems. The development of radiation damage-resistant materials requires irradiation testing facilities which reproduce, as closely as possible, the thermal and neutronic environment expected in a power-producing reactor. The Oak Ridge National Laboratory (ORNL) reference core design for the Center for Neutron Research (CNR) reactor provides for instrumented facilities in regions of both hard and mixed neutron spectra, with substantially higher fluxes than are currently available. The benefits of these new facilities to the development of radiation damage resistant materials are discussed in terms of the major US fission and fusion reactor programs

Implosion anisotropy of neutron kinetic energy distributions as measured with the neutron time-of-flight diagnostics at the National Ignition Facility

Science.gov (United States)

Hartouni, Edward; Eckart, Mark; Field, John; Grim, Gary; Hatarik, Robert; Moore, Alastair; Munro, David; Sayer, Daniel; Schlossberg, David

2017-10-01

Neutron kinetic energy distributions from fusion reactions are characterized predominantly by the excess energy, Q, of the fusion reaction and the variance of kinetic energy which is related to the thermal temperature of the plasma as shown by e.g. Brysk. High statistics, high quality neutron time-of-flight spectra obtained at the National Ignition Facility provide a means of measuring small changes to the neutron kinetic energy due to the spatial and temporal distribution of plasma temperature, density and velocity. The modifications to the neutron kinetic energy distribution as described by Munro include plasma velocity terms with spatial orientation, suggesting that the neutron kinetic energy distributions could be anisotropic when viewed by multiple lines-of-sight. These anisotropies provide a diagnostic of burn averaged plasma velocity distributions. We present the results of measurements made for a variety of DT implosions and discuss their possible physical interpretations. This work was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under contract DE-AC52-07NA27344. Lawrence Livermore National Security, LLC.

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)

Neutron noise analysis for malfunction diagnosis at sodium cooled reactors

International Nuclear Information System (INIS)

Hoppe, P.

1978-09-01

For the investigation of the potential use of neutron noise analysis at sodium cooled power reactors, measurements have been performed at the KNK I reactor over a period of 18 month under different operational conditions. The signal fluctuations of the following tranducers have been recorded: In-core and Ex-core neutron detectors, temperature-, flow-, pressure-, vibration- and acoustic sensors. These extensive measurements have been analyzed in the frequency range from 0,001 Hz to 1000 Hz with all currently known methods for the identification of noise sources. The following results have been found: - Neutron noise for f 20 Hz the white detection noise prevails. In the region from 1 Hz to 20 Hz the vibrations of core components contribute to neutron noise. - Neutron noise is influenced by the state of the plant. - The contributions to neutron noise due to the fluctuations of coolant flow and inlet temperature are small compared to those produced by the movements of the control rod initiated by the reactor control system. The quantitatively unidentifiable amount of reactivity fluctuations (0,6 time-dependent thermal bowing of the core. With respect to these results and by calculation of the neutron noise patterns to be expected for the SNR 300, the following possible applications for neutron noise analysis have been found: By means of neutron noise analysis only reactivity fluctuations can be identified and supervised which are produced by time dependent changes of the core geometry. Furthermore neutron noise analysis is well suited for a sensitive detection of control rod vibrations and of local sodium boiling. Finally it can be used for the surveillance of the proper functioning of the reactor control system and of the control rod drive mechanism. (orig./HP) 891 HP [de

Intermediate-energy neutron beams from reactors for NCT

International Nuclear Information System (INIS)

Brugger, R.M.; Less, T.J.; Passmore, G.G.

1986-01-01

This paper discusses ways that a beam of intermediate-energy neutrons might be extracted from a nuclear reactor. The challenge is to suppress the fast-neutron component and the gamma-ray component of the flux while leaving enough of the intermediate-energy neutrons in the beam to be able to perform neutron capture therapy in less than an hour exposure time. Moderators, filters, and reflectors are considered. 11 references, 7 figures, 3 tables

Neutronics of a mixed-flow gas-core reactor

International Nuclear Information System (INIS)

Soran, P.D.; Hansen, G.E.

1977-11-01

The study was made to investigate the neutronic feasibility of a mixed-flow gas-core reactor. Three reactor concepts were studied: four- and seven-cell radial reactors and a seven-cell scallop reactor. The reactors were fueled with UF 6 (either U-233 or U-235) and various parameters were varied. A four-cell reactor is not practical nor is the U-235 fueled seven-cell radial reactor; however, the 7-cell U-233 radial and scallop reactors can satisfy all design criteria. The mixed flow gas core reactor is a very attractive reactor concept and warrants further investigation

Transport calculation of neutron flux distribution in reflector of PW reactor

International Nuclear Information System (INIS)

Remec, I.

1982-01-01

Two-dimensional transport calculation of the neutron flux and spectrum in the equatorial plain of PW reactor, using computer program DOT 3, is presented. Results show significant differences between neutron fields in which test samples and reactor vessel are exposed. (author)

Change of neutron flow sensors effectiveness in the course of reactor experiments

International Nuclear Information System (INIS)

Kurpesheva, A.M.; Kotov, V.M.; Zhotabaev, Zh.R.

2007-01-01

Full text: IGR reactor is a reactor of thermal capacity type. During the operation, uranium-graphite core can be heated up to 1500 deg. C and reactivity can be changed considerably. Core dimensions are comparatively small. Amount of control rods, providing required reactivity, is not big as well. Increasing of core temperature leads to the rise of neutrons path length in its basic material - graphite. Change of temperature is not even. All this causes the non-conservation of neutron flows ratio in irradiated sample and in the place of reactor power sensors installation. Deviations in this ratio were registered during the number of reactor experiments. Empiric corrections can be introduced in order to decrease influence of change of neutron flow effectiveness upon provision of required parameters of investigated matters load. However, dependence of these corrections upon many factors can lead to the increasing of instability of process control. Previous experiment-calculated experiments showed inequality of neutron field in the place of sensors location (up to tens of percent), low effectiveness of experimental works, carried out without access to the individual reactor laying elements. Imperfection during the experiment was an idea of possibility to connect distribution of out of reactor neutron flow and control rods position. Subsequent analysis showed that for the development of representative phenomenon model it is necessary to take into account reactor operation dynamic subject to unevenness of heating of individual laying parts. Elemental calculations showed that temperature laying effects in the change of neutron outer field are great. Algorithm of calculations for the change of outer filed and field of investigated fabrication includes calculation of neutron-physic reactor characteristics interlacing with calculations of thermal-physic reactor characteristics, providing correlation of temperature fields for neutron-physic calculations. In the course of such

Utilization of boron irradiation filters in reactor neutron activation via epithermal (n,γ) and fast neutron reactions

International Nuclear Information System (INIS)

Chisela, F.

1986-01-01

The technique of instrumental neutron activation analysis based on irradiation with reactor epithermal and fast neutrons has been described and evaluated. Important characteristics of boron neutron absorbers used to remove thermal neutrons from the reactor neutron spectrum have been examined and compared with those of cadmium. Three boron compound shields, have been designed and constructed at the BER II 5MW reactor for use in epithermal neutron activation analysis of biological materials. The major advantages offered by these filters in this application include the flexibility of varying the filter thickness, the low radioactivity induced in the filters during irradiation, ease of fabrication and the relatively low cost of the filter materials. The radiation heating due to the 10 B(n,α) 7 Li-reaction has been experimentally investigated for the filters used and the results obtained confirm the necessity for efficient cooling of these filters during irradiation. Three irradiation facilities have been characterized with respect to the neutron flux density and the flux spatial distribution. An experiment has been designed and carried out to compensate the flux inhomogeneity in two irradiation positions of the DBV facility caused by the reactor geometry. Several biological samples including well characterized reference materials have been analysed after epithermal activation and the results compared with those obtained with the classical thermal neutron activation method. Improved sensitivity of determination has been found for elements with high resonance integral to thermal neutron cross section ratios (RI/σ 0 ). The range of elements that can be determined instrumentally is extended and the time scale of analysis is considerably reduced. (orig.) [de

Research and development of the software for visualizing nuclear reactor and neutronics analysis

International Nuclear Information System (INIS)

Okui, Shota; Sekimoto, Hiroshi

2009-01-01

It is not easy to image three-dimensional construct of a nuclear reactor with only its two-dimensional figure because it contains a number of structures and its construction is very complicated. Several visualization softwares for the nuclear reactor or some other plant exist, but require high skills and their operation is not simple. In this study, we developed nuclear reactor visualization software, called 'Visual Reactor (VR)', which does not require specific skills. We added the neutronics analysis code to that software. This code executes cell calculation, neutron diffusion calculation and nuclide burnup calculation by itself without any other codes. We tried to treat simple physics model in order to perform these calculation in a short time. Neutronics characteristics, such as neutron flux and power density distribution, are visualized on structure of nuclear reactor. Target operating system is Microsoft Windows XP or Vista. VR is utilized to figure out the structure of nuclear reactor and whole picture of neutronics characteristics. (author)

Determination of neutron flux densities in WWR-S reactor core

International Nuclear Information System (INIS)

Tomasek, F.

1989-04-01

The method is described of determining neutron flux densities and neutron fluences using activation detectors. The basic definitions and relations for determining reaction rates, fluence and neutron flux as well as the characteristics of some reactions and of sitable activation detectors are reported. The flux densities were determined of thermal and fast neutrons and of gamma quanta in the WWR-S reactor core. The data measured in the period 1984-1987 are tabulated. Cross sections for the individual reactions were determined from spectra measurements processed using program SAND-II and cross section library ENDF-B IV. Neutron flux densities were also measured for the WWR-S reactor vertical channels. (E.J.). 10 figs., 8 tabs., 111 refs

Neutron polarizing set-up of the Sofia IRT research reactor

International Nuclear Information System (INIS)

Krezhov, K.; Mikhajlova, V.; Okorokov, A.

1990-01-01

Neutron polarizing set-up of one of the horizontal beam tubes of the IRT-200 research reactor of the Bulgarian Institute of Nuclear Research and Nuclear Energy is presented. Neutron mirrors are extensively used in an effort to compensate the moderate reactor beam intensity by the high reflected intensity and wide-band transmittance of the mirror neutron guides. Time-to-flight technique using a slotted neutron absorbing chopper with a horizontal rotation axis has been applied to obtain the exit neutron spectra. Beam polarization and flipping ratios have been determined. Cadmium ratio in the polarized beam has been found almost 10 4 and the average polarization has been measured to be higher than 96%. 3 figs, 3 refs

Solution of the reactor point kinetics equations by MATLAB computing

Directory of Open Access Journals (Sweden)

Singh Sudhansu S.

2015-01-01

Full Text Available The numerical solution of the point kinetics equations in the presence of Newtonian temperature feedback has been a challenging issue for analyzing the reactor transients. Reactor point kinetics equations are a system of stiff ordinary differential equations which need special numerical treatments. Although a plethora of numerical intricacies have been introduced to solve the point kinetics equations over the years, some of the simple and straightforward methods still work very efficiently with extraordinary accuracy. As an example, it has been shown recently that the fundamental backward Euler finite difference algorithm with its simplicity has proven to be one of the most effective legacy methods. Complementing the back-ward Euler finite difference scheme, the present work demonstrates the application of ordinary differential equation suite available in the MATLAB software package to solve the stiff reactor point kinetics equations with Newtonian temperature feedback effects very effectively by analyzing various classic benchmark cases. Fair accuracy of the results implies the efficient application of MATLAB ordinary differential equation suite for solving the reactor point kinetics equations as an alternate method for future applications.

Effective neutron temperature measurements in well moderated reactor by the reactivity coefficient method

International Nuclear Information System (INIS)

Raisic, N.; Klinc, T.

1968-11-01

The ratio of the reactivity changes of a nuclear reactor produced by successive introduction of two different neutron absorbers in the reactor core, has been measured and information on effective neutron temperature at a particular point obtained. Boron was used as a l/v absorber and cadmium as an absorber sensiti ve to neutron temperature. Effective neutron temperature distribution has been deduced by moving absorbers across the reactor core and observing the corresponding reactivity changes. (author)

Calculation of the neutron parameters of fast thermal reactor

International Nuclear Information System (INIS)

Kukuleanu, V.; Mocioiu, D.; Drutse, E.; Konstantinesku, E.

1975-01-01

The system of neutron calculation for fast reactors is given. This system was used for estimation of physical parameters of fast thermal reactors investigated. The results obtained and different specific problems of the reactors of this type are described. (author)

The advanced neutron source - A world-class research reactor facility

International Nuclear Information System (INIS)

Thompson, P.B.; Meek, W.E.

1993-01-01

The advanced neutron source (ANS) is a new facility being designed at the Oak Ridge National Laboratory that is based on a heavy-water-moderated reactor and extensive experiment and user-support facilities. The primary purpose of the ANS is to provide world-class facilities for neutron scattering research, isotope production, and materials irradiation in the United States. The neutrons provided by the reactor will be thermalized to produce sources of hot, thermal, cold, very cold, and ultracold neutrons usable at the experiment stations. Beams of cold neutrons will be directed into a large guide hall using neutron guide technology, greatly enhancing the number of research stations possible in the project. Fundamental and nuclear physics, materials analysis, and other research pro- grams will share the neutron beam facilities. Sufficient laboratory and office space will be provided to create an effective user-oriented environment

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

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)

The effect of temperature and the control rod position on the spatial neutron flux distribution in the Syrian Miniature Neutron Source Reactor

International Nuclear Information System (INIS)

Khattab, K.; Omar, H.; Ghazi, N.

2007-01-01

The effect of water and fuel temperature increase and changes in the control rod positions on the spatial neutron flux distribution in the Syrian Miniature Neutron Source Reactor (MNSR) is discussed. The cross sections of all the reactor components at different temperatures are generated using the WIMSD4 code. These group constants are used then in the CITATION code to calculate the special neutron flux distribution using four energy groups. This work shows that water and fuel temperature increase in the reactor during the reactor daily operating time does not affect the spatial neutron flux distribution in the reactor. Changing the control rod position does not affect as well the spatial neutron flux distribution except in the region around the control rod position. This stability in the spatial neutron flux distribution, especially in the inner and outer irradiation sites, makes MNSR as a good tool for the neutron activation analysis (NAA) technique and production of radioisotopes with medium or short half lives during the reactor daily operating time. (author)

Measuring neutron flux density in near-vessel space of a commercial WWER-1000 reactor

International Nuclear Information System (INIS)

Borodkin, G.I.; Eremin, A.N.; Lomakin, S.S.; Morozov, A.G.

1987-01-01

Distribution of neutron flux density in two experimental channels on the reactor vessel external surface and in ionization chamber channel of a commercial WWER-1000 reactor, is measured by the activation detector technique. Azimuthal distributions of fast and thermal neutron fluxes and height distributions of fast neutron flux density within energy range >1.2 and 2.3 MeV are obtained. Conclusion is made, that reactor core state and its structural peculiarities in the measurement range essentially affect space and energy distribution of neutron field near the vessel. It should be taken into account when determining permissible neutron fluence for the reactor vessel

Fast neutron reactors: the safety point of view

International Nuclear Information System (INIS)

Laverie, M.; Avenas, M.

1984-01-01

All versions of nuclear reactors present favourable and unfavourable characteristics from the point of view of safety. The safety of the installations is obtained by making efforts to utilize in the best possible way those which are favourable and by taking proper steps in the face of those which are unfavourable. The present article shows how this general principle has been applied as regards the fast neutron reactors of integrated design which have been developped in France, taking into account the specific features of this version. A qualitative method to compare the safety of this version with that of pressurized water reactors which has been widely put to the test commercially all over the world is presented. These analyses make, generally speaking, several positive characteristics stand out for these fast neutron reactors from the safety aspects [fr

Ozone disintegration kinetics in the reactor for tyres decomposition

International Nuclear Information System (INIS)

Golota, V.I.; Manujlenko, O.V.; Taran, G.V.; Pis'menetskij, A.S.; Zamuriev, A.A.

2010-01-01

The results of theoretical and experimental research of ozone disintegration kinetics in the chemical reactor which is developed for decomposition of tyres in the ozone-air environment are presented. Analytical expression for dependence of ozone concentration in the reactor from time and from parameters of the task, such as volume speed of ozone-air mixture feed on a reactor input, concentration of ozone on the input to the reactor, volume speed of output of the used mixture, reactor size, and square of its internal surface is obtained. It is shown that at the same speed of ozone-air mixture pro rolling through the reactor, with growth of ozone concentration on the input, value of stationary concentration in the reactor grows, remaining always less than concentration on the input. It is also shown that at the same ozone concentration on the input, with growth of speed of ozone-air mixture pro rolling through the reactor, value of stationary ozone concentration in the reactor also grows, remaining always less than ozone concentration on the input. The ozone disintegration kinetics in the reactor in a wide range of speed of ozone-air mixture pro rolling through the reactor (0.15, 0.30, 0.45, 0.60 m3/hour) and various ozone concentration on the input (5, 10, 15, 20 g/m3) is experimentally studied. It is shown that experimental results with good accuracy coincide with the theoretical. Direct experiment showed the essential influence of the internal surface of the reactor on the ozone disintegration kinetics.

Design of neutron 'fluse-trop' in Tehran Research Reactor

International Nuclear Information System (INIS)

Khalafi, H.

1999-01-01

In this project, a Neutron Flux-Trap intended for 5MW Tehran Research Reactor was designed. Fuel conversion from HEU to LEU in research reactors usually deprives the core from the high neutron flux. Therefore one has to look for a remedy in such situations otherwise radioisotope production, especially for those neutron demanding ones, falls down dramatically. The initiations of Neutron Flux-Trap idea comes true to face up this problem and provide an appropriate place inside or outside the core with sufficient neutron flux higher than the normal average level. To implement such a design, a number of codes and calculational tools have been used. At preliminary stage, WIMSD and EXTERMINATOR-II were used and then at later stages CITATION and MCNP codes were used for final design. Furthermore, SAND-II and ORIGEN were also employed for spectral analysis and radioisotope production calculations. Good agreements resulted with experiments and especially the case that Neutron Flux-Trap filled with ordinary water

Vanadium Beta Emission Detectors for Reactor In-Core Neutron Monitoring

Energy Technology Data Exchange (ETDEWEB)

Andersson, I Oe; Soederlund, B

1969-06-15

In-core flux measurements are becoming increasingly important in both power reactors and test reactors. In particular power distribution measurements in large power reactors have to be performed with a great number of neutron detectors capable of withstanding high integrated flux values. This report presents a summary of the development and application of a new type of nuclear radiation sensor, a beta emission detector, for measurements at high neutron flux levels. The work has been carried out at the Section for Instrumentation and has been the basis for a type of neutron detector employed in the Marviken in-core system as well as for other types. The report describes the design and principle of operation, experiments and tests. Also included are the results and comments from a long-term irradiation of some detectors in the Halden reactor.

Neutron Radiography at the RP-10 reactor

International Nuclear Information System (INIS)

Hinostrosa, H.; Ravello, Y.; Cornejo, N.; Mendoza, M.; Montoya, M.

1992-01-01

The facility of neutron radiography at the RP-10 peruvian research reactor is described. The factor of collimation L/D is 149; the Cadmium ratio for the gold in the inspection's area is 4.5, and the thermal neutrons flux on the sample is 3,14 x 10 6 n/cm 2 s (author). 5 refs. 5 fig

OPTIMIZATION OF THE EPITHERMAL NEUTRON BEAM FOR BORON NEUTRON CAPTURE THERAPY AT THE BROOKHAVEN MEDICAL RESEARCH REACTOR.

Energy Technology Data Exchange (ETDEWEB)

HU,J.P.; RORER,D.C.; RECINIELLO,R.N.; HOLDEN,N.E.

2002-08-18

Clinical trials of Boron Neutron Capture Therapy for patients with malignant brain tumor had been carried out for half a decade, using an epithermal neutron beam at the Brookhaven's Medical Reactor. The decision to permanently close this reactor in 2000 cut short the efforts to implement a new conceptual design to optimize this beam in preparation for use with possible new protocols. Details of the conceptual design to produce a higher intensity, more forward-directed neutron beam with less contamination from gamma rays, fast and thermal neutrons are presented here for their potential applicability to other reactor facilities. Monte Carlo calculations were used to predict the flux and absorbed dose produced by the proposed design. The results were benchmarked by the dose rate and flux measurements taken at the facility then in use.

Neutron streaming evaluation for the DREAM fusion power reactor

International Nuclear Information System (INIS)

Seki, Yasushi; Nishio, Satoshi; Ueda, Shuzo; Kurihara, Ryoichi

2000-01-01

Aiming at high degree of safety and benign environmental effect, we have proposed a tokamak fusion reactor concept called DREAM, which stands for DRastically EAsy Maintenance Reactor. The blanket structure of the reactor is made from very low activation SiC/SiC composites and cooled by non-reactive helium gas. High net thermal efficiency of about 50% is realized by 900 C helium gas and high plant availability is possible with simple maintenance scheme. In the DREAM Reactor, neutron streaming is a big problem because cooling pipes with diameter larger than 80 cm are used for blanket heat removal. Neutron streaming through the cooling pipes could cause hot spots in the superconducting magnets adjacent to the cooling pipes to shorten the magnet lifetime or increase cryogenic cooling requirement. Neutron streaming could also activate components such as gas turbine further away from the fusion plasma. The effect of neutron streaming through the helium cooling pipes was evaluated for the two types of cooling pipe extraction scheme. The result of a preliminary calculation indicates the gas turbine activation prohibits personnel access in the case of inboard pipe extraction while with additional shielding measures, limited contact maintenance is possible in the case of outboard extraction. (author)

Nuclear reactor ex-core startup neutron detector

International Nuclear Information System (INIS)

Wyvill, J.R.

1980-01-01

A sensitive ex-core neutron detector is needed to monitor the power level of reactors during startup. The neutron detector of this invention has a photomultiplier with window resistant to radiation darkening at the input end and an electrical connector at the output end. The photomultiplier receives light signals from a neutron-responsive scintillator medium, typically a cerium-doped lithium silicate glass, that responds to neutrons after they have been thermalized by a silicone resin moderator. Enclosing and shielding the photmultiplier, the scintillator medium and the moderator is a combined lead and borated silicone resin housing

Intermediate and fast neutron absorbed doses in fast neutron field at the RB reactor

International Nuclear Information System (INIS)

Sokcic-Kostic, M.; Pesic, M.; Antic, D.

1987-10-01

The experimental fuel channel EFC is created as one of the fast neutron fields at the RB reactor. The intermediate and fast neutron spectra in EFC are measured by activation technique. The intermediate and fast neutron absorbed doses are computed on the basis of these experimental results. At the end the obtained doses are compared. (author)

Application of Extreme Learning Machines to inverse neutron kinetics

International Nuclear Information System (INIS)

Picca, Paolo; Furfaro, Roberto

2017-01-01

Highlights: • The paper applies the Extreme Learning Machines (ELMs) to inverse reactor problems. • Multi-group transport model is used for the inversion as opposed to point kinetics. • ELMs are compared against Artificial Neural Networks (ANNs). • Various options are tested to improve the reliability of the estimation. • Results highlight the potential of the ELM approach. - Abstract: The paper presents the application of Extreme Leaning Machines (ELMs) for inverse reactor kinetic applications. ELMs were proposed by Huang and co-workers (2004, 2006a,b, 2015), which showed their enhances capabilities in terms of training speed and generalization with respect to classical Artificial Neural Networks (ANNs). ELMs are here implemented for reactivity determination as an alternative to ANNs (e.g. Picca et al. (2008)) and Gaussian Processes (Picca and Furfaro, 2012). After a review of the main features of ELMs, their application to inverse kinetic problems is proposed. The ELMs performance is tested on a typical accelerator drive system configuration (Yalina reactor) and the inversion is carried out on an accurate kinetic model (multi-group transport).

Monochromatic neutron beam production at Brazilian nuclear research reactors

Science.gov (United States)

Stasiulevicius, Roberto; Rodrigues, Claudio; Parente, Carlos B. R.; Voi, Dante L.; Rogers, John D.

2000-12-01

Monochomatic beams of neutrons are obtained form a nuclear reactor polychromatic beam by the diffraction process, suing a single crystal energy selector. In Brazil, two nuclear research reactors, the swimming pool model IEA-R1 and the Argonaut type IEN-R1 have been used to carry out measurements with this technique. Neutron spectra have been measured using crystal spectrometers installed on the main beam lines of each reactor. The performance of conventional- artificial and natural selected crystals has been verified by the multipurpose neutron diffractometers installed at IEA-R1 and simple crystal spectrometer in operator at IEN- R1. A practical figure of merit formula was introduced to evaluate the performance and relative reflectivity of the selected planes of a single crystal. The total of 16 natural crystals were selected for use in the neutron monochromator, including a total of 24 families of planes. Twelve of these natural crystal types and respective best family of planes were measured directly with the multipurpose neutron diffractometers. The neutron spectrometer installed at IEN- R1 was used to confirm test results of the better specimens. The usually conventional-artificial crystal spacing distance range is limited to 3.4 angstrom. The interplane distance range has now been increased to approximately 10 angstrom by use of naturally occurring crystals. The neutron diffraction technique with conventional and natural crystals for energy selection and filtering can be utilized to obtain monochromatic sub and thermal neutrons with energies in the range of 0.001 to 10 eV. The thermal neutron is considered a good tool or probe for general applications in various fields, such as condensed matter, chemistry, biology, industrial applications and others.

INR TRIGA Research Reactors: A Neutron Source for Radioisotopes and Materials Investigation

International Nuclear Information System (INIS)

Barbos, D.; Ciocanescu, M.; Paunoiu, C.; Bucsa, A.F.

2013-01-01

At the INR there are 2 high intensity neutron sources. These sources are in fact the two nuclear TRIGA reactors: TRIGA SSR 14 MW and TRIGA ACPR. TRIGA stationary reactor is provided with several in-core irradiation channels. Other several out-of-core irradiation channels are located in the vertical channels in the beryllium reflector blocks. The maximum value of the thermal neutron flux (E 14 cm -2 s -1 and of fast neutron flux (E>1 MeV) is 6.89×10 13 cm -2 s -1 . For neutron activation analysis both reactors are used and k0-NAA method has been implemented. At INR Pitesti a prompt gamma ray neutron activation analysis devices has been designed, manufactured ant put into operation. For nuclear materials properties investigation neutron radiography methods was developed in INR. For these purposes two neutron radiography devices were manufacture, one of them underwater and other one dry. The neutron beams are used for investigation of materials properties and components produced or under development for applications in the energy sector (fission and fusion). At TRIGA 14 MW reactor a neutron difractormeter and a SANS devices are available for material residual stress and texture measurements. TRIGA 14 MW reactor is used for medical and industrial radioisotopes production ( 131 I, 125 I, 192 Ir, etc) and a method for 99 Mo- 99 Tc production from fission is under developing. At INR Pitesti several special programmes for new types of nuclear fuel behavior characterization are under development. (author)

Neuro-diffuse algorithm for neutronic power identification of TRIGA Mark III reactor; Algoritmo neuro-difuso para la identificacion de la potencia neutronica del reactor Triga Mark III

Energy Technology Data Exchange (ETDEWEB)

Rojas R, E.; Benitez R, J. S. [Instituto Tecnologico de Toluca, Division de Estudios de Posgrado e Investigacion, Av. Tecnologico s/n, Ex-Rancho La Virgen, 50140 Metepec, Estado de Mexico (Mexico); Segovia de los Rios, J. A.; Rivero G, T. [ININ, Gerencia de Ciencias Aplicadas, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)], e-mail: jorge.benitez@inin.gob.mx

2009-10-15

In this work are presented the results of design and implementation of an algorithm based on diffuse logic systems and neural networks like method of neutronic power identification of TRIGA Mark III reactor. This algorithm uses the punctual kinetics equation as data generator of training, a cost function and a learning stage based on the descending gradient algorithm allow to optimize the parameters of membership functions of a diffuse system. Also, a series of criteria like part of the initial conditions of training algorithm are established. These criteria according to the carried out simulations show a quick convergence of neutronic power estimated from the first iterations. (Author)

Multicomponent activation detector measurements of reactor neutron spectra

International Nuclear Information System (INIS)

Sandberg, J.; Aarnio, P. A.; Routti, J. T.

1984-01-01

Information on the neutron flux is required in many applications of research reactors, such as activation analysis or radiation damage measurements. Flux spectrum measurements are commonly carried out with activation foils. The reaction types used are threshold reactions in the fast energy region, resonance reactions in the intermediate region and neutron capture reactions with l/v-cross section in the thermal region. It has been shown that it is possible to combine several detector elements into homogeneous multicomponent detectors. The activities of all detector reaction products can be determined with a single gamma spectrum measurement. The multicomponent principle sets some restrictions on the choice of detector reactions, for example, each product nuclide may be produced in one reaction only. Separate multicomponent threshold and resonance detectors were designed for the fast and intermediate regions, respectively. The detectors were fabricated in polyethylene irradiation capsules or quartz glass ampoules, and they were irradiated in a cadmium cover. The detectors were succesfully used in the irradiation ring and in the core of a Triga reactor. The intermediate and fast neutron spectrum was unfolded with the least-squares unfolding program LOUHI. According to the preliminary results multicomponent activation detectors might constitute a convenient means for carrying out routine neutron spectrum measurements in research reactors. (orig.)

FORE-2, Thermohydraulics and Space-Independent Reactor Kinetics for Transients

International Nuclear Information System (INIS)

Fox, J.N.; Lawler, B.E.; Butz, H.R.; Heames, T.J.

1984-01-01

1 - Description of problem or function: FORE2 is a coupled thermal hydraulics-point kinetics digital computer code designed to calculate significant reactor parameters under steady-state conditions, or as functions of time during transients. The transients may result from a programmed reactivity insertion or a power change. Variable inlet coolant flow rate and temperature are considered. The code calculates the reactor power, the individual reactivity feedbacks, and the temperature of coolant, cladding, fuel, structure, and additional material for up to seven axial positions in three channel types which represent radial zones of the reactor. The heat of fusion, accompanying fuel melting, the liquid metal voiding reactivity, and the spatial and the time variation of the fuel cladding gap coefficient due to changes in gap size are considered. 2 - Method of solution: FORE2 input consists of property data, geometry, power and flow distribution factors, external time varying functions, experimental coefficients, and termination data. The differential equations for fluid flow, heat transfer, and point neutronics are solved by explicit finite-difference procedures. 3 - Restrictions on the complexity of the problem: Reactor excursions which can be calculated are restricted to those transients in which the reactor is not substantially destroyed. As a general rule, changes in reactor geometry and composition during an excursion are limited to those cases in which the reactivity effects of the changes may be considered as small perturbations of the initial system. Thus, accidents involving large-scale disassembly and bulk meltdown of a core are not covered by FORE2. FORE2 is valid only while the core retains its initial geometry

Neutron shielding studies on an advanced molten salt fast reactor design

International Nuclear Information System (INIS)

Merk, Bruno; Konheiser, Jörg

2014-01-01

Highlights: • Material damage due to irradiation has already been discovered at the MSRE. • Neutronic analysis of MSFR with curved blanket wall geometry. • Neutron fluence limit at the wall of the outer vessel can be kept for 80 years. • Shielded MSFR core will be of same dimension than a SFR core. - Abstract: The molten salt reactor technology has gained some new interest. In contrast to the historic molten salt reactors, the current projects are based on designing a molten salt fast reactor. Thus the shielding becomes significantly more challenging than in historic concepts. One very interesting and innovative result of the most recent EURATOM project on molten salt reactors – EVOL – is the fluid flow optimized design of the inner reactor vessel using curved blanket walls. The developed structure leads to a very uniform flow distribution. The design avoids all internal structures. Based on this new geometry a model for neutron physics calculation is presented. The major steps are: the modeling of the curved geometry in the unstructured mesh neutron transport code HELIOS and the determination of the real neutron flux and power distribution for this new geometry. The developed model is then used for the determination of the neutron fluence distribution in the inner and outer wall of the system. Based on these results an optimized shielding strategy is developed for the molten salt fast reactor to keep the fluence in the safety related outer vessel below expected limit values. A lifetime of 80 years can be assured, but the size of the core/blanket system will be comparable to a sodium cooled fast reactor. The HELIOS results are verified against Monte-Carlo calculations with very satisfactory agreement for a deep penetration problem

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

Development of SiC Neutron Detector Assembly to Measure the Neutron Flux of the Reactor Core

Energy Technology Data Exchange (ETDEWEB)

Park, Se Hwan; Park, June Sic; Shin, Hee Sung; Kim, Ho Dong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Yong Kyun [Hanyang University, Seoul (Korea, Republic of)

2012-05-15

At present, the conventional detector to measure the neutron at harsh environment is a Self Powered Neutron Detector (SPND). Rhodium(Rh)-103 is in the SPND. When neutron is incident on the Rhodium, the neutron capture reaction occurs, and the Rh-103 is converted to Rh-104. The Rh-104 is decayed to Pd-104 by {beta}-decay, and electrons are generated as the decay products. Because of the half life of Rh-104, approximately 5 minutes are required for the SPND output to reach the equilibrium condition. Therefore the on-line monitoring of the nuclear reactor state is limited if the neutron flux in the reactor core is monitored with the SPND. Silicon carbide (SiC) has the possibility to be developed as neutron detector at harsh environment, because the SiC can be operative at high temperature and high neutron flux conditions. Previously, the basic operation properties of the SiC detector were studied. Also, the radiation response of the SiC detector was studied at high neutron and gamma dose rate. The measurement results for an ex-core neutron flux monitor or a neutron flux monitor of the spent fuel were published. The SiC detector was also developed as neutron detector to measure the fissile material with active interrogation method. However, the studies about the development of SiC detector are still limited. In the present work, the radiation damage effect of the SiC detector was studied. The detector structure was determined based on the study, and a neutron detector assembly was made with the SiC detectors. The neutron and gamma-ray response of the detector assembly is presented in this paper. The detector assembly was positioned in the HANARO research reactor core, the performance test was done. The preliminary results are also included in this paper

High Conduction Neutron Absorber to Simulate Fast Reactor Environment in an Existing Test Reactor

Energy Technology Data Exchange (ETDEWEB)

Guillen, Donna; Greenwood, Lawrence R.; Parry, James

2014-06-22

A need was determined for a thermal neutron absorbing material that could be cooled in a gas reactor environment without using large amounts of a coolant that would thermalize the neutron flux. A new neutron absorbing material was developed that provided high conduction so a small amount of water would be sufficient for cooling thereby thermalizing the flux as little as possible. An irradiation experiment was performed to assess the effects of radiation and the performance of a new neutron absorbing material. Neutron fluence monitors were placed inside specially fabricated holders within a set of drop-in capsules and irradiated for up to four cycles in the Advanced Test Reactor. Following irradiation, the neutron fluence monitor wires were analyzed by gamma and x-ray spectrometry to determine the activities of the activation products. The adjusted neutron fluences were calculated and grouped into three bins – thermal, epithermal and fast to evaluate the spectral shift created by the new material. Fluence monitors were evaluated after four different irradiation periods to evaluate the effects of burn-up in the absorbing material. Additionally, activities of the three highest activity isotopes present in the specimens are given.

The IAEA collaborating centre for neutron activation based methodologies of research reactors

International Nuclear Information System (INIS)

Bode, P.

2010-01-01

The Reactor Institute Delft of the Delft University of Technology houses the Netherlands' only academic nuclear research reactor, with associated instrumentation and laboratories, for scientific education and research with ionizing radiation. The Institute's swimming pool type research reactor reached first criticality in 1963 and is currently operated at 2MW thermal powers on a 100 h/week basis. The reactor is equipped with neutron mirror guides serving ultra modern neutron beam physics instruments and with a very bright positron facility. Fully automated gamma-ray spectrometry systems are used by the laboratory for neutron activation analysis, providing large scale services under an ISO/IEC 17025:2005 compliant management system, being (since 1993) the first accredited laboratory of its kind in the world. Already for several years, this laboratory is sustainable by rendering these services to both the public and the private sector. The prime user of the Institute's fac ilities is the scientific Research Department of Radiation, Radionuclide and Reactors of the Faculty of Applied Sciences, housed inside the building. All reactor facilities are also made available for use by or for services to, external clients (industry, government, private sector, other (international research institutes and universities). The Reactor Institute Delft was inaugurated in May 2009 as a new lAEA Collaborating Centre for Neutron Activation Based Methodologies of Research Reactors. The collaboration involves education, research and development in (I) Production of reactor-produced, no-carrier added radioisotopes of high specific activity via neutron activation; (II) Neutron activation analysis with emphasis on automation as well as analysis of large samples, and radiotracer techniques and as a cross-cutting activity, (IIl) Quality assurance and management in research and application of research reactor based techniques and in research reactor operations. This c ollaboration will

Thai Research Reactor (TRR-1/M1) Neutron Beam Measurements

International Nuclear Information System (INIS)

Ratanatongchai, Wichian

2009-07-01

Full text: Neutron beam tube of neutron radiography facility at Thai Research Reactor (TRR-1/M1) Thailand Institute of Nuclear Technology (public organization) is a divergent beam. The rectangular open-end of the beam tube is 16 cm x 17 cm while the inner-end is closed to the reactor core. The neutron beam size was measured using 20 cm x 40 cm neutron imaging plate. The measurement at the position 100 cm from the end of the collimator has shown that the beam size was 18.2 cm x 19.0 cm. Gamma ray in neutron the beam was also measured by the identical position using industrial X ray film. The area of gamma ray was 27.8 cm x 31.1 cm with the highest intensity found to be along the neutron beam circumference

Neutron collar calibration for assay of LWR [light-water reactor] fuel assemblies

International Nuclear Information System (INIS)

Menlove, H.O.; Pieper, J.E.

1987-03-01

The neutron-coincidence collar is used for the verification of the uranium content in light-water reactor fuel assemblies. An AmLi neutron source is used to give an active interrogation of the fuel assembly to measure the 235 U content, and the 238 U content is verified from a passive neutron-coincidence measurement. This report gives the collar calibration data of pressurized-water reactor and boiling-water reactor fuel assemblies. Calibration curves and correction factors are presented for neutron absorbers (burnable poisons) and different fuel assembly sizes. The data were collected at Exxon Nuclear, Franco-Belge de Fabrication de Combustibles, ASEA-Atom, and other nuclear fuel fabrication facilities

Optimization study of ultracold neutron sources at TRIGA reactors using MCNP

International Nuclear Information System (INIS)

Pokotilovskij, Yu.N.; Rogov, A.D.

1997-01-01

Monte Carlo simulation for the optimization of ultracold and very cold neutron sources for TRIGA reactors is performed. The calculations of thermal and cold neutron fluxes from the TRIGA reactor for different positions and configurations of a very cold solid methane moderator were performed with using the MCNP program. The production of neutrons in the ultracold and very cold energy range was calculated for the most promising final moderators (converters): very cold solid deuterium and heavy methane. The radiation energy deposition was calculated for the optimized solid methane-heavy methane cold neutron moderator

Research reactor of the future: The advanced neutron source

International Nuclear Information System (INIS)

Appleton, B.; West, C.

1994-01-01

Agents for cancer detection and treatment, stronger materials, better electronic gadgets, and other consumer and industrial products - these are assured benefits of a research reactor project proposed for Oak Ridge. Just as American companies have again assumed world leadership in producing semiconductor chips as well as cars and trucks, the United States is poised to retake the lead in neutron science by building and operating the $2.9 billion Advanced Neutron Source (ANS) research reactor by the start of the next century. In 1985, the neutron community, led by ORNL researchers, proposed a pioneering project, later called the ANS. Scheduled to begin operation in 2003, the ANS is seen not only as a replacement for the aging HFIR and HFBR but also as the best laboratory in the world for conducting neutron-based research

Studying the effect of xenon poisoning on the power of the Syrian miniature neutron source reactor

International Nuclear Information System (INIS)

Khamis, I.; Khattab, K.

1999-07-01

The uranium 235 is often used as a fuel to produce the energy in nuclear reactors. Uranium nuclei are fissioned with thermal neutrons and produce energy plus a number of neutrons. A fraction of such fission neutrons is involved in other fission with new nuclei to sustain the fission reactions. The remain fraction of the neutrons is lost from the reactor in two ways: escaped from the reactor, or absorbed with other nuclei that exist in the reactor before or produced from fission. Fission nuclei which absorb neutrons heavily are called p oison , such as Xe 135. Because Xe 135 absorbs neutrons heavily, it reduces the number of neutrons in the reactor. Hence, Xe 135 is studied explicitly in the MNSR reactor, and calculation of its negative reactivity is presented in this research during the operation, equilibrium, and after the shutting down of the reactor. (author)

RA-0 reactor. New neutronic calculations

International Nuclear Information System (INIS)

Rumis, D.; Leszczynski, F.

1990-01-01

An updating of the neutronic calculations performed at the RA-0 reactor, located at the Natural, Physical and Exact Sciences Faculty of Cordoba National University, are herein described. The techniques used for the calculation of a reactor like the RA-0 allows prediction in detail of the flux behaviour in the core's interior and in the reflector, which will be helpful for experiments design. In particular, the use of WIMSD4 code to make calculations on the reactor implies a novelty in the possible applications of this code to solve the problems that arise in practice. (Author) [es

Cold neutron PGAA facility developments at university research reactors in the USA

International Nuclear Information System (INIS)

Uenlue, K.; Rios-Martinez, C.

2005-01-01

The PGAA applications can be enhanced by using subthermal neutrons, cold neutrons at university research reactors. Only two cold neutron beam facilities were developed at the U.S. university research reactors, namely at Cornell University and the University of Texas at Austin. Both facilities used mesitylene moderator. The mesitylene moderator in the Cornell Cold Neutron Beam Facility (CNBF) was cooled by a helium cryorefrigerator via copper cold fingers to maintain the moderator below 30 K at full power reactor operation. Texas Cold Neutron Source (TCNS) also uses mesitylene moderator that is cooled by a cryorefrigerator via a neon thermosiphon. The operation of the TCNS is based on a helium cryorefrigerator, which liquefies neon gas in a 3-m long thermosiphon. The thermosiphon cools and maintains mesitylene moderator at about 30 K in a chamber. Neutrons streaming through the mesitylene chamber are moderated and thus reduce their energy to produce a cold neutron distribution. (author)

Characterization of the Annular Core Research Reactor (ACRR Neutron Radiography System Imaging Plane

Directory of Open Access Journals (Sweden)

Kaiser Krista

2016-01-01

Full Text Available The Annular Core Research Reactor (ACRR at Sandia National Laboratories (SNL is an epithermal pool-type research reactor licensed up to a thermal power of 2.4 MW. The ACRR facility has a neutron radiography facility that is used for imaging a wide range of items including reactor fuel and neutron generators. The ACRR neutron radiography system has four apertures (65:1, 125:1, 250:1, and 500:1 available to experimenters. The neutron flux and spectrum as well as the gamma dose rate were characterized at the imaging plane for the ACRR's neutron radiography system for the 65:1, 125:1 and 250:1 apertures.

Safety reviews of the Brazilian multipurpose reactor; Avaliacoes de seguranca do reator multiproposito brasileiro

Energy Technology Data Exchange (ETDEWEB)

Soares, Humberto Vitor

2014-08-01

This work presents a model developed for thermal hydraulic (TH) simulation of the Multipurpose Brazilian Reactor (RMB), whose Brazilian proposal for design, construction and operation was established in 2007. This reactor has as main proposed the production of radioisotopes for use in exams of nuclear medicine, material tests and utilization of neutrons beams. Besides of the TH modeling and safety analysis of the reactor, the application of a methodology to perform coupled calculation thermal-hydraulic/neutron kinetic (TH/NK) is also presented. Initially, the RMB was modeled in the safety analysis RELAP5 code. This code performs the thermal hydraulic calculation using point kinetics. Subsequently, the model was adapted and verified to the RELAP5-3D© code. This code performs the process of internal coupling through the option of nodal neutron kinetics calculation using the NESTLE code which solves the neutron diffusion equation. To generate the neutronic group constants, which are macroscopic cross sections that serve as input data for the neutronic codes, it was used the WIMSD-5B cell calculation code. The neutron analysis code PARCS was also used to model the 3D RMB core in order to compare the results of radial and axial average power distribution with the results generated by RELAP5-3D© code and with the available results of the CITATION neutron kinetic code. The safety analyses demonstrated safe behavior of the reactor through situations of possible transients. The 3D coupled calculations to the steady state operation also showed expected behavior, as well as the RMB neutronic analyzes performed with the codes NESTLE and PARCS.(author)

Feasibility study of Self Powered Neutron Detectors in Fast Reactors for detecting local change in neutron flux distribution

International Nuclear Information System (INIS)

Jammes, Christian; Filliatre, Philippe; Verma, Vasudha; Hellesen, Carl; Jacobsson Svard, Staffan

2015-01-01

Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor system. Diverse possibilities of detector systems installation have to be investigated with respect to practicality and feasibility according to the detection parameters. In this paper, we demonstrate the feasibility of using self powered neutron detectors as in-core detectors in fast reactors for detecting local change in neutron flux distribution. We show that the gamma contribution from fission products decay in the fuel and activation of structural materials is very small compared to the fission gammas. Thus, it is possible for the in-core SPND signal to follow changes in local neutron flux as they are proportional to each other. This implies that the signal from an in-core SPND can provide dynamic information on the neutron flux perturbations occurring inside the reactor core. (authors)

Feasibility study of Self Powered Neutron Detectors in Fast Reactors for detecting local change in neutron flux distribution

Energy Technology Data Exchange (ETDEWEB)

Jammes, Christian; Filliatre, Philippe [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St Paul-Lez-Durance, (France); Verma, Vasudha; Hellesen, Carl; Jacobsson Svard, Staffan [Division of Applied Nuclear Physics, Uppsala University, SE-75120 Uppsala, (Sweden)

2015-07-01

Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor system. Diverse possibilities of detector systems installation have to be investigated with respect to practicality and feasibility according to the detection parameters. In this paper, we demonstrate the feasibility of using self powered neutron detectors as in-core detectors in fast reactors for detecting local change in neutron flux distribution. We show that the gamma contribution from fission products decay in the fuel and activation of structural materials is very small compared to the fission gammas. Thus, it is possible for the in-core SPND signal to follow changes in local neutron flux as they are proportional to each other. This implies that the signal from an in-core SPND can provide dynamic information on the neutron flux perturbations occurring inside the reactor core. (authors)

Study of the RP-10 reactor neutron beam applied to the neutron radiography

International Nuclear Information System (INIS)

Zegarra, Manuel; Lopez, Alcides

2013-01-01

We have studied the RP-10 reactor radial neutron beam No. 3, which is used for neutron radiographies, by comparing radiograph's with and without the inner duct, and neutron flux determination with in flakes along the external duct, being the presence of photons creating signals at comparable levels of neutron effects, which reduce the quality of the analysis, values around 10 6 and 10 4 n/cm 2 s for thermal and epithermal flux were obtained respectively. It is recommended evaluate the design of the internal duct which presents strong photon emission. (authors).

Proceedings of 2. Yugoslav symposium on reactor physics, Part 1, Herceg Novi (Yugoslavia), 27-29 Sep 1966

International Nuclear Information System (INIS)

1966-01-01

This Volume 1 of the Proceedings of 2. Yugoslav symposium on reactor physics includes nine papers dealing with the following topics: reactor kinetics, reactor noise, neutron detection, methods for calculating neutron flux spatial and time dependence in the reactor cores of both heavy and light water moderated experimental reactors, calculation of reactor lattice parameters, reactor instrumentation, reactor monitoring systems; measuring methods of reactor parameters; reactor experimental facilities

Progress report on reactor physics research program, January 1963 - February 1964

Energy Technology Data Exchange (ETDEWEB)

NONE

1964-02-15

This progress report is a part of the annual report of the department of reactor physics prepared for the Boris Kidric Institute of nuclear sciences. It is a review of research activities in the field of theoretical and experimental reactor physics in the year 1973. A part of this program was included in the NPY Cooperative program in reactor physics. The topics covered by this report are as follows: Calculations of the thermal neutron distribution and reaction rate in a reactor cell and comparison with experiments; buckling measurements; thermalization and slowing down of neutrons; pulsed neutron source techniques; and reactor kinetics.

Progress report on reactor physics research program, January 1963 - February 1964

International Nuclear Information System (INIS)

1964-02-01

This progress report is a part of the annual report of the department of reactor physics prepared for the Boris Kidric Institute of nuclear sciences. It is a review of research activities in the field of theoretical and experimental reactor physics in the year 1973. A part of this program was included in the NPY Cooperative program in reactor physics. The topics covered by this report are as follows: Calculations of the thermal neutron distribution and reaction rate in a reactor cell and comparison with experiments; buckling measurements; thermalization and slowing down of neutrons; pulsed neutron source techniques; and reactor kinetics

Cold neutron source conceptual designing for Tehran Research Reactor

International Nuclear Information System (INIS)

Khajvand, N.; Mirvakili, S.M.; Faghihi, F.

2016-01-01

Highlights: • Cold neutron source conceptual designing for Tehran research reactor is carried out. • Type and geometry of moderator and dimensions of cold neutron source are analyzed. • Liquid hydrogen with more ortho-concentration can be better option as moderator. - Abstract: A cold neutron source (CNS) conceptual designing for the Tehran Research Reactor (TRR) were carried out using MCNPX code. In this study, a horizontal beam tube of the core which has appropriate the highest thermal flux is selected and parametric analysis to choose the type and geometry of the moderator, and the required CNS dimensions for maximizing the cold neutron production was performed. In this design the moderator cell has a spherical annulus structure, and the cold neutron flux and its brightness are calculated together with the nuclear heat load of the CNS for a variety of materials including liquid hydrogen, liquid deuterium, and solid methane. Based on our study, liquid hydrogen with more ortho-concentration than para and solid methane are the best options.

Reactor physics for non-nuclear engineers

International Nuclear Information System (INIS)

Lewis, E.E.

2011-01-01

A one-term undergraduate course in reactor physics is described. The instructional format is strongly influenced by its intended audience of non-nuclear engineering students. In contrast to legacy treatments of the subject, the course focuses on the physics of nuclear power reactors with no attempt to include instruction in numerical methods. The multi-physics of power reactors is emphasized highlighting the close interactions between neutronic and thermal phenomena in design and analysis. Consequently, the material's sequencing also differs from traditional treatments, for example treating kinetics before the neutron diffusion is introduced. (author)

Neutron spectra in two beam ports of the TRIGA Mark III reactor

International Nuclear Information System (INIS)

Vega C, H. R.; Hernandez D, V. M.; Aguilar, F.; Paredes, L.; Rivera M, T.

2013-10-01

The neutron spectra have been measured in two beam ports, radial and tangential, of the TRIGA Mark III nuclear reactor from the National Institute of Nuclear Research. Measurements were carried out with the core with mixed fuel (Leu 8.5/20 and Flip Heu 8.5/70). Two reactor powers, 5 and 10 W, were used during neutron spectra measurements using a Bonner sphere spectrometer with a 6 Lil(Eu) scintillator and 2, 3, 5, 8, 10 and 12 inches-diameter high density polyethylene spheres. The neutron spectra were unfolded using the NSDUAZ unfolding code; from each spectrum the total neutron flux, the neutron mean energy and the neutron ambient dose equivalent dose were determined. Measured spectra show fission (E≥ 0.1 MeV), epithermal (from 0.4 eV up to 0.1 MeV) and thermal neutrons (E≤ 0.4 eV). For both reactor powers the spectra in the radial beam port have similar features which are different to the neutron spectrum characteristics in the tangential beam port. (Author)

Neutron spectra in two beam ports of the TRIGA Mark III reactor

Energy Technology Data Exchange (ETDEWEB)

Vega C, H. R.; Hernandez D, V. M. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98060 Zacatecas (Mexico); Aguilar, F.; Paredes, L. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Rivera M, T., E-mail: fermineutron@yahoo.com [IPN, Centro de Investigacion en Ciencia Aplicada y Tecnologia Avanzada, Unidad Legaria, Av. Legaria 694, 11500 Mexico D. F. (Mexico)

2013-10-15

The neutron spectra have been measured in two beam ports, radial and tangential, of the TRIGA Mark III nuclear reactor from the National Institute of Nuclear Research. Measurements were carried out with the core with mixed fuel (Leu 8.5/20 and Flip Heu 8.5/70). Two reactor powers, 5 and 10 W, were used during neutron spectra measurements using a Bonner sphere spectrometer with a {sup 6}Lil(Eu) scintillator and 2, 3, 5, 8, 10 and 12 inches-diameter high density polyethylene spheres. The neutron spectra were unfolded using the NSDUAZ unfolding code; from each spectrum the total neutron flux, the neutron mean energy and the neutron ambient dose equivalent dose were determined. Measured spectra show fission (E≥ 0.1 MeV), epithermal (from 0.4 eV up to 0.1 MeV) and thermal neutrons (E≤ 0.4 eV). For both reactor powers the spectra in the radial beam port have similar features which are different to the neutron spectrum characteristics in the tangential beam port. (Author)

Study of the multiplication and kinetic effects of salt mixtures and salt blanket micromodels on thermal neutron spectra of heavy water MAKET facility

International Nuclear Information System (INIS)

Titarenko, Yu.E.; Batyaev, V.F.; Borovlev, S.P.; Gladkikh, N.G.; Igumnov, M.M.; Legostaev, V.O.; Karpikhin, E.I.; Konev, V.N.; Kushnerev, Yu.T.; Ryazhsky, V.I.; Spiridonov, V.G.; Chernyavsky, E.V.; Shvedov, O.V.

2009-10-01

The main goal of the Project is to study and evaluate nuclear characteristics of materials and isotopes involved in processes of irradiated nuclear fuel transmutation. This principal task is subdivided into 9 subtasks subject to the neutron or proton source used, the type of the nuclear process under study, isotope collection, characteristics of which are to be investigated, etc. In the presented extract of the Project Activity report the measurements there were used the MAKET zero-power heavy-water reactor in the measurements there was employed a large set of minor actinide samples highly enriched with the main isotope. The samples were obtained with mass-separator SM-2 (VNIIEF). At the heavy-water reactor MAKET (ITEP) there were measured multiplying and kinetic characteristics of salt mixtures basing on the spectra of fast and thermal neutrons. The salt mixtures of zirconium and sodium fluorides were available in salt blanket models (SBM) of cylindrical shape. There were measured the neutron spectra formed by this micro-model as well as the effective fission cross-sections of neptunium, plutonium, americium and curium isotopes caused by SBM neutrons. The neutron spectra in the measurement positions were determined from activation reaction rates. (author)

Neutron Field Characterization of Irradiation Locations Applied to the Slovenian TRIGA Reactor

International Nuclear Information System (INIS)

Barbot, Loic; Domergue, Christophe; Breaud, Stephane; Destouches, Christophe; Villard, Jean-Francois; Snoj, Luka; Stancar, Ziga; Radulovic, Vladimir; Trkov, Andrej

2013-06-01

This work deals with several neutron flux measurement instruments and particle transport calculations combined in a method to assess the neutron field in experimental locations in nuclear reactor core or reflector. First test of this method in the TRIGA Mark II of Slovenia led to the assessment of three energy groups neutron fluxes in central irradiation locations within reactor core. (authors)

DETECTION OF COATING FAILURES IN A NEUTRONIC REACTOR

Science.gov (United States)

Snell, A.H.; Allison, S.K.

1958-02-11

This patent relates to water-cooled reactor systems and discloses a means to detect leaks in the jackets of jacketed fuel elements comprising a neutron detector located in the cooling water discharge pipe,the pipe being provided with an enlarged portion for housing the detector so that the latter is completely surrounded by the water in its passage through the pipe, said enlarged portion and detector being shielded from the reactor for the purpose of detecting only those delayed neutrons emitted in the cooling water and due to the latter picking up fission fragments from the defective fuel elements.

Neutronics issues in fusion-fission hybrid reactor design

International Nuclear Information System (INIS)

Liu Chengan

1995-01-01

The coupled neutron and γ-ray transport equations and nuclear number density equations, and its computer program systems concerned in fusion-fission hybrid reactor design are briefly described. The current status and focal point for coming work of nuclear data used in fusion reactor design are explained

Identification of nuclear reactor characteristics by the reactor noise analysis

International Nuclear Information System (INIS)

Yashima, Hideyuki

1980-01-01

Reactor noise analysis method was applied to TRIGA II Research Reactor (Atomic Research Laboratory, Musashi Institute of Technology) and computed power spectral density (PSD) from the CIC current record. PSD has provided many valuable informations regarding to the reactor kinetics, including the effect of control rods vibration. Another information of neutron physics parameters were obtained and this result was compared with the parameter which was formerly measured by the Feynman-α experiment. Through these experiments we could find overall frequency characteristics of TRIGA II Reactor. (author)

Transient Analysis of Generation IV quick reactors; Analisis de Transitorios en Reactores Rapidos de Generacion IV

Energy Technology Data Exchange (ETDEWEB)

Vazquez, M.; Martin-Fuertes, F.

2013-07-01

As a complement to the attached code 3D neutron-CIEMAT thermohydraulic added a module to simulate transient. Temporary kinetics is resolved by factoring flow in a spatial part and another storm. MCNP provides the reactivity and updated spatial function and COBRA-IV calculates the temperature distribution. Temporary dependence of amplitude is calculated using time delayed neutron Kinetic equations. As an example of application, examines a transient loss of flow in MYRRHA, a lead-cooled experimental reactor.

Physical model study of neutron noise induced by vibration of reactor internals

International Nuclear Information System (INIS)

Liu Jinhui; Gu Fangyu

1999-01-01

The author presents a physical model of neutron noise induced by reactor internals vibration in frequency domain. Based on system control theory, the reactor dynamic equations are coupled with random vibration equation, and non-linear terms are also taken into accounted while treating the random vibration. Experiments carried out on a zero-power reactor show that the model can be used to describe dynamic character of neutron noise induced by internals' vibration. The model establishes a method to help to determine internals'vibration features, and to diagnosis anomalies through neutron noise

Solving the uncommon nuclear reactor core neutronics problems

International Nuclear Information System (INIS)

Vondy, D.R.; Fowler, T.B.

1983-01-01

Calculational procedures have been implemented for solving importance and higher harmonic neutronics problems. Solutions are obtained routinely to support analysis of reactor core performance, treating up to three space coordinates with the multigroup diffusion theory approximation to neutron transport. The techniques used and some of the calculational difficulties are discussed

Resonance self-shielding effect in uncertainty quantification of fission reactor neutronics parameters

International Nuclear Information System (INIS)

Chiba, Go; Tsuji, Masashi; Narabayashi, Tadashi

2014-01-01

In order to properly quantify fission reactor neutronics parameter uncertainties, we have to use covariance data and sensitivity profiles consistently. In the present paper, we establish two consistent methodologies for uncertainty quantification: a self-shielded cross section-based consistent methodology and an infinitely-diluted cross section-based consistent methodology. With these methodologies and the covariance data of uranium-238 nuclear data given in JENDL-3.3, we quantify uncertainties of infinite neutron multiplication factors of light water reactor and fast reactor fuel cells. While an inconsistent methodology gives results which depend on the energy group structure of neutron flux and neutron-nuclide reaction cross section representation, both the consistent methodologies give fair results with no such dependences.

Bayesian calibration of reactor neutron flux spectrum using activation detectors measurements: Application to CALIBAN reactor

International Nuclear Information System (INIS)

Cartier, J.; Casoli, P.; Chappert, F.

2013-01-01

In this paper, we present calibration methods in order to estimate reactor neutron flux spectrum and its uncertainties by using integral activation measurements. These techniques are performed using Bayesian and MCMC framework. These methods are applied to integral activation experiments in the cavity of the CALIBAN reactor. We estimate the neutron flux and its related uncertainties. The originality of this work is that these uncertainties take into account measurements uncertainties, cross-sections uncertainties and model error. In particular, our results give a very good approximation of the total flux and indicate that neutron flux from MCNP simulation for energies above about 5 MeV seems to overestimate the 'real flux'. (authors)

Radiation protection commissioning of neutron beam instruments at the OPAL research reactor

International Nuclear Information System (INIS)

Parkes, Alison; Saratsopoulos, John; Deura, Michael; Kenny, Pat

2008-01-01

The neutron beam facilities at the 20 MW OPAL Research Reactor were commissioned in 2007 and 2008. The initial suite of eight neutron beam instruments on two thermal neutron guides, two cold neutron guides and one thermal beam port located at the reactor face, together with their associated shielding were progressively installed and commissioned according to their individual project plans. Radiation surveys were systematically conducted as reactor power was raised in a step-wise manner to 20 MW in order to validate instrument shielding design and performance. The performance of each neutron guide was assessed by neutron energy spectrum and flux measurements. The activation of beam line components, decay times assessments and access procedures for Bragg Institute beam instrument scientists were established. The multiple configurations for each instrument and the influence of operating more than one instrument or beamline simultaneously were also tested. Areas of interest were the shielding around the secondary shutters, guide shield and bunker shield interfaces and monochromator doors. The shielding performance, safety interlock checks, improvements, radiation exposures and related radiation protection challenges are discussed. This paper discusses the health physics experience of commissioning the OPAL Research Reactor neutron beam facilities and describes health physics results, actions taken and lessons learned during commissioning. (author)

Development of a generalized stochastic model for the analysis of monoenergetic space-time nuclear factor Kinetics

International Nuclear Information System (INIS)

Pham, Nhu Viet Ha

2011-02-01

To predict the space-time dependent behavior of a nuclear reactor, the conventional space-dependent kinetics equations are widely used for treating the spatial variables. However, the solutions of such deterministic space-dependent kinetics equations, which give only the mean values of the neutron population and the delayed neutron precursor concentrations, do not offer sufficient insight into the actual dynamic processes within a reactor, where the interacting populations vary randomly with space and time. It is also noted that at high power levels, the random behavior of a reactor is negligible but at low power levels, such as at start-up, random fluctuations in population dynamics can be significant. To mathematically describe the evolution of the state of a nuclear reactor using a set of stochastic kinetics equations, the forward stochastic model (FSM) in stochastic kinetics theory is devised through the concept of reactor transition probability and its probability generating function as the spatial domain of a reactor is partitioned into a number of space cells. Nevertheless, the FSM equations for the mean value of neutron and precursor distribution are deterministic-like. Furthermore, the numerical treatment of the FSM equations for the means, variances, and covariances is quite complicated and time-consuming. In the present study, a generalized stochastic model (called the stochastic space-dependent kinetics model or SSKM) based on the FSM and the Its stochastic differential equations was newly developed for the analysis of monoenergetic spacetime nuclear reactor kinetics in one dimension. First, the FSM equations for determining the mean values of neutron and delayed-neutron precursor populations were considered as the deterministic ones without taking into account their variances and covariances. Second, the system of interest was randomized again in the light of the Its stochastic differential equations in order to derive the SSKM. The proposed model

The determination of neutron energy spectrum in reactor core C1 of reactor VR-1 Sparrow

Energy Technology Data Exchange (ETDEWEB)

Vins, M. [Department of Nuclear Reactors, Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University, V Holesovickach 2, 180 00 Prague 8 (Czech Republic)], E-mail: vinsmiro@seznam.cz

2008-07-15

This contribution overviews neutron spectrum measurement, which was done on training reactor VR-1 Sparrow with a new nuclear fuel. Former nuclear fuel IRT-3M was changed for current nuclear fuel IRT-4M with lower enrichment of 235U (enrichment was reduced from former 36% to 20%) in terms of Reduced Enrichment for Research and Test Reactors (RERTR) Program. Neutron spectrum measurement was obtained by irradiation of activation foils at the end of pipe of rabit system and consecutive deconvolution of obtained saturated activities. Deconvolution was performed by computer iterative code SAND-II with 620 groups' structure. All gamma measurements were performed on Canberra HPGe. Activation foils were chosen according physical and nuclear parameters from the set of certificated foils. The Resulting differential flux at the end of pipe of rabit system agreed well with typical spectrum of light water reactor. Measurement of neutron spectrum has brought better knowledge about new reactor core C1 and improved methodology of activation measurement. (author)

Neutron dosimeters and survey meters in accelerators, reactors and other neutron environments

International Nuclear Information System (INIS)

1989-03-01

Neutron fields in occupationally accessible areas around nuclear reactors, radioisotope sources and medical and high energy accelerators have been characterized using currently available information. Neutron, rem meters, such as the Leake detector, are the most suitable instruments available for conducting neutron dose rate surveys in the vicinity of radioisotope neutron sources, nuclear reactors and medical accelerators. However, these instruments have been shown to be deficient in that they overrespond by a factor of four to neutrons in the 0.1 to 1 MeV range and are insensitive to neutrons from about 1 eV up to about 10 keV. Also, they are insensitive to neutrons above 20 MeV and their use must be restricted near high energy accelerators where significant numbers of neutrons above 20 MeV are known to be present. The most suitable instrument of measure dose from neutrons above 20 MeV is the 12 C(n,2n) 11 C scintillation chamber. Commercially available rem meters frequently use BF 3 counters in the pulse mode to detect thermal neutrons. Therefore, measurements around pulsed accelerators must be made with caution to ensure that the detector is not saturated during each pulse and that the accelerator pulse period is greater than the response time of the detector. The personal neutron dosimeters currently available either are known to be insensitive to neutrons above 20 MeV or have not been tested. Also, all except the albedo dosimeter are insensitive to or have not been tested for neutron energies in the range 1 eV to 10 keV. Several dosimeter types respond reasonably well to neutrons in the energy range 0.1 to 15 MeV, for example, CR-39, bubble and superheated drop detectors. However, the first gas a lower limit of sensitivity of about 0.3 mSv. The bubble detector can be designed to measure doses as small as 1μSv and offers the additional benefit of direct-reading capability. The superheated drop detector is not suitable for use around pulsed accelerators because

Use of self powered neutron detectors in the IEA-R1 reactor

International Nuclear Information System (INIS)

Galo Rocha, F. del.

1989-01-01

A survey of self-powered neutron detectors, SPND, which are used as part of the in-core instrumentation of nuclear reactors is presented. Measurements with Co and Er SPND's were made in the IEA-R1 reactor for determining the neutron flux distribution and the integral reactor power. Due to the size of the available detectors, the neutron flux distribution could not be obtained with accuracy. The results obtained in the reactor power measurements demonstrate that the SPND have the linearity and the quick response necessary for a reactor power channel. This work also presents a proposed design of a SPND using Pt as wire emissor. This proposed design is based in the experience gained in building two prototypes. The greatest difficulties encountered include materials and technology to perform the delicate weldings. (author)

Measurements of neutron flux in the RA reactor; Merenje karakteristika neutronskog fluksa u reaktoru RA

Energy Technology Data Exchange (ETDEWEB)

Raisic, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1961-12-15

This report includes results of the following measurements performed at the RA reactor: thermal neutron flux in the experimental channels, epithermal and fast neutron flux, neutron flux in the biological shield, neutron flux distribution in the reactor cell.

Fusion Reactor Materials

International Nuclear Information System (INIS)

Moons, F.

1998-01-01

SCK-CEN's programme on fusion reactor materials includes studies (1) to investigate fracture mechanics of neutron-irradiated beryllium; (2) to describe the helium behaviour in irradiated beryllium at atomic scale; (3) to define the kinetics of beryllium reacting with air or steam; (3) to perform a feasibility study for the testing of integrated blanket modules under neutron irradiation. Progress and achievements in 1997 are reported

Absolute measurement of neutron fluxes inside the reactor core

International Nuclear Information System (INIS)

Ajdacic, S. V.

1964-10-01

The subject of this work is the development and study of two methods of neutron measurements in nuclear reactors, the new method of high neutron flux measurements and the Li 6 -semiconductor neutron spectrometer. This work is presented in four sections: Section I. The introduction explains the need for neutron measurements in reactors. A critical survey is given of the existing methods of high neutron flux measurement and methods of fast neutron spectrum determination. Section II. Theoretical basis of the work of semiconductor counters and their most important characteristics are given. Section III. The main point of this section is in presenting the basis of the new method which the author developed, i.e., the long-tube method, and the results obtained by it, with particular emphasis on absolute measurement of high neutron fluxes. Advantages and limitations of this method are discussed in details at the end of this section. Section IV. A comparison of the existing semiconductor neutron spectrometers is made and their advantages and shortcomings underlined. A critical analysis of the obtained results with the Li 6 -semiconductor spectrometer with plane geometry is given. A new type of Li 6 -semiconductor spectrometer is described, its characteristics experimentally determined, and a comparison of it with a classical Li 6 -spectrometer made (author)

A fast and flexible reactor physics model for simulating neutron spectra and depletion in fast reactors - 202

International Nuclear Information System (INIS)

Recktenwald, G.D.; Bronk, L.A.; Deinert, M.R.

2010-01-01

Determining the time dependent concentration of isotopes within a nuclear reactor core is central to the analysis of nuclear fuel cycles. We present a fast, flexible tool for determining the time dependent neutron spectrum within fast reactors. The code (VBUDS: visualization, burnup, depletion and spectra) uses a two region, multigroup collision probability model to simulate the energy dependent neutron flux and tracks the buildup and burnout of 24 actinides, as well as fission products. While originally developed for LWR simulations, the model is shown to produce fast reactor spectra that show high degree of fidelity to available fast reactor benchmarks. (authors)

Physical particularities of nuclear reactors using heavy moderators of neutrons

Energy Technology Data Exchange (ETDEWEB)

Kulikov, G. G., E-mail: ggkulikov@mephi.ru; Shmelev, A. N. [National Research Nuclear University MEPhI (Moscow Engineering Physics Institute) (Russian Federation)

2016-12-15

In nuclear reactors, thermal neutron spectra are formed using moderators with small atomic weights. For fast reactors, inserting such moderators in the core may create problems since they efficiently decelerate the neutrons. In order to form an intermediate neutron spectrum, it is preferable to employ neutron moderators with sufficiently large atomic weights, using {sup 233}U as a fissile nuclide and {sup 232}Th and {sup 231}Pa as fertile ones. The aim of the work is to investigate the properties of heavy neutron moderators and to assess their advantages. The analysis employs the JENDL-4.0 nuclear data library and the SCALE program package for simulating the variation of fuel composition caused by irradiation in the reactor. The following main results are obtained. By using heavy moderators with small neutron moderation steps, one is able to (1) increase the rate of resonance capture, so that the amount of fertile material in the fuel may be reduced while maintaining the breeding factor of the core; (2) use the vacant space for improving the fuel-element properties by adding inert, strong, and thermally conductive materials and by implementing dispersive fuel elements in which the fissile material is self-replenished and neutron multiplication remains stable during the process of fuel burnup; and (3) employ mixtures of different fertile materials with resonance capture cross sections in order to increase the resonance-lattice density and the probability of resonance neutron capture leading to formation of fissile material. The general conclusion is that, by forming an intermediate neutron spectrum with heavy neutron moderators, one can use the fuel more efficiently and improve nuclear safety.