Neutronics and thermal hydraulics coupling scheme for design improvement of liquid metal fast systems

International Nuclear Information System (INIS)

Sanchez-Espinoza, V.H.; Jaeger, W.; Travleev, A.; Monti, L.; Doern, R.

2009-01-01

Many advanced reactor concepts are nowadays under investigations within the Generation IV international initiative as well as in European research programs including subcritical and critical fast reactor systems cooled by liquid metal, gas and supercritical water. The Institute of Neutron Physics and Reactor Technology (INR) at the Forschungszentrum Karlsruhe GmbH is involved in different European projects like IP EUROTRANS, ELSY, ESFR. The main goal of these projects is, among others, to assess the technical feasibility of proposed concepts regarding safety, economics and transmutation requirements. In view of increased computer capabilities, improved computational schemes, where the neutronic and the thermal hydraulic solution is iteratively coupled, become practicable. The codes ERANOS2.1 and TRACE are being coupled to analyze fuel assembly or core designs of lead-cooled fast reactors (LFR). The neutronic solution obtained with the coupled system for a LFR fuel assembly was compared with the MCNP5 solution. It was shown that the coupled system is predicting physically sound results. The iterative coupling scheme was realized using Perlscripts and auxiliary Fortran programs to ensure that the mapping between the neutronic and the thermal hydraulic part is consistent. The coupled scheme is very flexible and appropriate for the neutron physical and thermal hydraulic investigation of fuel assemblies and of cores of lead cooled fast reactors. The developed methods and the obtained results will be presented and discussed. (author)

Determination of the Axial-Vector Weak Coupling Constant with Ultracold Neutrons

International Nuclear Information System (INIS)

Liu, J.; Mendenhall, M. P.; Carr, R.; Filippone, B. W.; Hickerson, K. P.; Perez Galvan, A.; Russell, R.; Holley, A. T.; Hoagland, J.; VornDick, B.; Back, H. O.; Pattie, R. W. Jr.; Young, A. R.; Bowles, T. J.; Clayton, S.; Currie, S.; Hogan, G. E.; Ito, T. M.; Makela, M.; Morris, C. L.

2010-01-01

A precise measurement of the neutron decay β asymmetry A 0 has been carried out using polarized ultracold neutrons from the pulsed spallation ultracold neutron source at the Los Alamos Neutron Science Center. Combining data obtained in 2008 and 2009, we report A 0 =-0.119 66±0.000 89 -0.00140 +0.00123 , from which we determine the ratio of the axial-vector to vector weak coupling of the nucleon g A /g V =-1.275 90 -0.00445 +0.00409 .

A tightly coupled scheme for neutronics and thermal–hydraulics using open-source software

International Nuclear Information System (INIS)

Wu, Hsingtzu; Rizwan-uddin

2016-01-01

Highlights: • A novel tight coupling scheme called “the Integrated Tight Coupling (ITC) method” is proposed. • The ITC method is implemented using DRAGON and OpenFOAM. • The ITC method is verified using a 1.5-D example. - Abstract: Coupling neutronic and thermal–hydraulic analyses of a nuclear reactor core is important because it helps identify the most relevant safety issues without conservative assumptions. Currently coupled computations solve the same governing equations using different coupling methods, which can be sorted into two categories: loose coupling and tight coupling. This paper proposes and verifies a third coupling approach called “the Integrated Tight Coupling (ITC) method”. The mathematical equations in the nuclear fuel are rearranged to be integrated via a novel concept of the energy-group function. In addition, the data from the neutron cross section library can be used directly. The ITC method is implemented using two open-source codes: the DRAGON code and OpenFOAM. Additionally, a coupled computation using these two codes is new and has not been done. The ITC method is verified using a 1.5-D (1-D neutronics and 2-D thermal–hydraulics) symmetric unit cell example. The mesh of the tightly integrated computation is 25% coarser than the loosely coupled one. Starting from a similar initial guess, the number of iterations for the ITC method is 24% fewer than that for the loosely coupled computation to reach the same accuracy. In addition, the ITC method is tested with different initial guesses. For all cases tested, the scheme converged to the same solution. With further improvement and additional testing, the ITC method has the potential to be incorporated with other neutronics and thermal–hydraulics codes.

Challenges in coupled thermal-hydraulics and neutronics simulations for LWR safety analysis

International Nuclear Information System (INIS)

Ivanov, Kostadin; Avramova, Maria

2007-01-01

The simulation of nuclear power plant accident conditions requires three-dimensional (3D) modeling of the reactor core to ensure a realistic description of physical phenomena. The operational flexibility of Light Water Reactor (LWR) plants can be improved by utilizing accurate 3D coupled neutronics/thermal-hydraulics calculations for safety margins evaluations. There are certain requirements to the coupling of thermal-hydraulic system codes and neutron-kinetics codes that ought to be considered. The objective of these requirements is to provide accurate solutions in a reasonable amount of CPU time in coupled simulations of detailed operational transient and accident scenarios. These requirements are met by the development and implementation of six basic components of the coupling methodologies: ways of coupling (internal or external coupling); coupling approach (integration algorithm or parallel processing); spatial mesh overlays; coupled time-step algorithms; coupling numerics (explicit, semi-implicit and implicit schemes); and coupled convergence schemes. These principles of the coupled simulations are discussed in details along with the scientific issues associated with the development of appropriate neutron cross-section libraries for coupled code transient modeling. The current trends in LWR nuclear power generation and regulation as well as the design of next generation LWR reactor concepts along with the continuing computer technology progress stimulate further development of these coupled code systems. These efforts have been focused towards extending the analysis capabilities as well as refining the scale and level of detail of the coupling. This article analyses the coupled phenomena and modeling challenges on both global (assembly-wise) and local (pin-wise) levels. The issues related to the consistent qualification of coupled code systems as well as their application to different types of LWR transients are presented. Finally, the advances in numerical

Alanine and TLD coupled detectors for fast neutron dose measurements in neutron capture therapy (NCT)

Energy Technology Data Exchange (ETDEWEB)

Cecilia, A.; Baccaro, S.; Cemmi, A. [ENEA-FIS-ION, Casaccia RC, Via Anguillarese 301, 00060 Santa Maria di Galeria, Rome (Italy); Colli, V.; Gambarini, G. [Dept. of Physics of the Univ., INFN, Via Celoria 16, 20133 Milan (Italy); Rosi, G. [ENEA-FIS-ION, Casaccia RC, Via Anguillarese 301, 00060 Santa Maria di Galeria, Rome (Italy); Scolari, L. [Dept. of Physics of the Univ., INFN, Via Celoria 16, 20133 Milan (Italy)

2004-07-01

A method was investigated to measure gamma and fast neutron doses in phantoms exposed to an epithermal neutron beam designed for neutron capture therapy (NCT). The gamma dose component was measured by TLD-300 [CaF{sub 2}:Tm] and the fast neutron dose, mainly due to elastic scattering with hydrogen nuclei, was measured by alanine dosemeters [CH{sub 3}CH(NH{sub 2})COOH]. The gamma and fast neutron doses deposited in alanine dosemeters are very near to those released in tissue, because of the alanine tissue equivalence. Couples of TLD-300 and alanine dosemeters were irradiated in phantoms positioned in the epithermal column of the Tapiro reactor (ENEA-Casaccia RC). The dosemeter response depends on the linear energy transfer (LET) of radiation, hence the precision and reliability of the fast neutron dose values obtained with the proposed method have been investigated. Results showed that the combination of alanine and TLD detectors is a promising method to separate gamma dose and fast neutron dose in NCT. (authors)

Continuous energy adjoint Monte Carlo for coupled neutron-photon transport

Energy Technology Data Exchange (ETDEWEB)

Hoogenboom, J.E. [Delft Univ. of Technology (Netherlands). Interfaculty Reactor Inst.

2001-07-01

Although the theory for adjoint Monte Carlo calculations with continuous energy treatment for neutrons as well as for photons is known, coupled neutron-photon transport problems present fundamental difficulties because of the discrete energies of the photons produced by neutron reactions. This problem was solved by forcing the energy of the adjoint photon to the required discrete value by an adjoint Compton scattering reaction or an adjoint pair production reaction. A mathematical derivation shows the exact procedures to follow for the generation of an adjoint neutron and its statistical weight. A numerical example demonstrates that correct detector responses are obtained compared to a standard forward Monte Carlo calculation. (orig.)

Coupled neutronics/thermal-hydraulics for analysis of molten salt reactor

International Nuclear Information System (INIS)

Guo, Zhangpeng; Zhou, Jianjun; Zhang, Dalin; Chaudri, Khurrum Saleem; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

2013-01-01

Highlights: ► A multiple-channel analysis code (MAC) is developed to be coupled with MCNP. ► 1/8 of core is simulated in MCNP and thermal-hydraulic code. ► The coupling calculation can achieve stable state after a few iterations. ► The coupling calculation results are in reasonable agreement with the analytic solutions of the ORNL. ► Parametric studies of MSR are performed to provide valuable information for future design MSR. -- Abstract: The Generation IV International Forum (GIF) selected molten salt reactor (MSR) among six advanced reactor types. It is characterized by a liquid circulating fuel that also serves as coolant. In this study, a multiple-channel analysis code (MAC) is developed and it is coupled with MCNP4c to analyze the neutronics/thermal-hydraulics behavior of molten salt reactor experiment (MSRE). The MAC calculates thermal-hydraulic parameters, such as temperature distribution, flow distribution and pressure drop. MCNP4c performs the analysis of effective multiplication factor, neutron flux and power distribution. A linkage code is developed to exchange data between MAC and MCNP to implement coupling iteration process until the power convergence is achieved. The coupling calculation can achieve converged solution after a few iterations. The results are in reasonable agreement with the analytic solutions from the ORNL. For further design analysis, parametric studies are performed to provide valuable information for new design of MSR. The effect of inlet temperature, graphite to molten salt volume ratio (G/Ms) from varying channel diameter and different power levels on the effective multiplication factor, neutron flux, graphite lifetime and temperature distribution are discussed in detail

Two-dimensional shielding benchmarks for iron at YAYOI, (1)

International Nuclear Information System (INIS)

Oka, Yoshiaki; An, Shigehiro; Kasai, Shigeru; Miyasaka, Shun-ichi; Koyama, Kinji.

The aim of this work is to assess the collapsed neutron and gamma multigroup cross sections for two dimensional discrete ordinate transport code. Two dimensional distributions of neutron flux and gamma ray dose through a 70cm thick and 94cm square iron shield were measured at the fast neutron source reactor ''YAYOI''. The iron shield was placed over the lead reflector in the vertical experimental column surrounded by heavy concrete wall. The detectors used in this experiment were threshold detectors In, Ni, Al, Mg, Fe and Zn, sandwitch resonance detectors Au, W and Co, activation foils Au for neutrons and thermoluminescence detectors for gamma ray dose. The experimental results were compared with the calculated ones by the discrete ordinate transport code ANISN and TWOTRAN. The region-wise, coupled neutron-gamma multigroup cross-sections (100n+20gamma, EURLIB structure) were generated from ENDF/B-IV library for neutrons and POPOP4 library for gamma-ray production cross-sections by using the code system RADHEAT. The effective microscopic neutron cross sections were obtained from the infinite dilution values applying ABBN type self-shielding factors. The gamma ray production multigroup cross-sections were calculated from these effective microscopic neutron cross-sections. For two-dimensional calculations the group constants were collapsed into 10 neutron groups and 3 gamma groups by using ANISN. (auth.)

Three dimensional neutronic/thermal-hydraulic coupled simulation of MSR in transient state condition

International Nuclear Information System (INIS)

Zhou, Jianjun; Zhang, Daling; Qiu, Suizheng; Su, Guanghui; Tian, Wenxi; Wu, Yingwei

2015-01-01

Highlights: • Developed a three dimensional neutronic/thermal-hydraulic coupled transient analysis code for MSR. • Investigated the neutron distribution and thermal-hydraulic characters of the core under transient condition. • Analyzed three different transient conditions of inlet temperature drop, reactivity jump and pump coastdown. - Abstract: MSR (molten salt reactor) use liquid molten salt as coolant and fuel solvent, which was the only one liquid reactor of six Generation IV reactor types. As a liquid reactor the physical property of reactor was significantly influenced by fuel salt flow and the conventional analysis methods applied in solid fuel reactors are not applicable for this type of reactors. The present work developed a three dimensional neutronic/thermal-hydraulic coupled code investigated the neutronics and thermo-hydraulics characteristics of the core in transient condition based on neutron diffusion theory and numerical heat transfer. The code consists of two group neutron diffusion equations for fast and thermal neutron fluxes and six group balance equations for delayed neutron precursors. The code was separately validated by neutron benchmark and flow and heat transfer benchmark. Three different transient conditions was analyzed with inlet temperature drop, reactivity jump and pump coastdown. The results provide some valuable information in design and research this kind of reactor

Three dimensional neutronic/thermal-hydraulic coupled simulation of MSR in transient state condition

Energy Technology Data Exchange (ETDEWEB)

Zhou, Jianjun [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China); College of Mechanical and Power Engineering, China Three Gorges University, No 8, Daxue road, Yichang, Hubei 443002 (China); Zhang, Daling, E-mail: dlzhang@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China); Qiu, Suizheng; Su, Guanghui; Tian, Wenxi; Wu, Yingwei [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China)

2015-02-15

Highlights: • Developed a three dimensional neutronic/thermal-hydraulic coupled transient analysis code for MSR. • Investigated the neutron distribution and thermal-hydraulic characters of the core under transient condition. • Analyzed three different transient conditions of inlet temperature drop, reactivity jump and pump coastdown. - Abstract: MSR (molten salt reactor) use liquid molten salt as coolant and fuel solvent, which was the only one liquid reactor of six Generation IV reactor types. As a liquid reactor the physical property of reactor was significantly influenced by fuel salt flow and the conventional analysis methods applied in solid fuel reactors are not applicable for this type of reactors. The present work developed a three dimensional neutronic/thermal-hydraulic coupled code investigated the neutronics and thermo-hydraulics characteristics of the core in transient condition based on neutron diffusion theory and numerical heat transfer. The code consists of two group neutron diffusion equations for fast and thermal neutron fluxes and six group balance equations for delayed neutron precursors. The code was separately validated by neutron benchmark and flow and heat transfer benchmark. Three different transient conditions was analyzed with inlet temperature drop, reactivity jump and pump coastdown. The results provide some valuable information in design and research this kind of reactor.

Some results on the neutron transport and the coupling of equations

International Nuclear Information System (INIS)

Bal, G.

1997-01-01

Neutron transport in nuclear reactors is well modeled by the linear Boltzmann transport equation. Its resolution is relatively easy but very expensive. To achieve whole core calculations, one has to consider simpler models, such as diffusion or homogeneous transport equations. However, the solutions may become inaccurate in particular situations (as accidents for instance). That is the reason why we wish to solve the equations on small area accurately and more coarsely on the remaining part of the core. It is than necessary to introduce some links between different discretizations or modelizations. In this note, we give some results on the coupling of different discretizations of all degrees of freedom of the integral-differential neutron transport equation (two degrees for the angular variable, on for the energy component, and two or three degrees for spatial position respectively in 2D (cylindrical symmetry) and 3D). Two chapters are devoted to the coupling of discrete ordinates methods (for angular discretization). The first one is theoretical and shows the well posing of the coupled problem, whereas the second one deals with numerical applications of practical interest (the results have been obtained from the neutron transport code developed at the R and D, which has been modified for introducing the coupling). Next, we present the nodal scheme RTN0, used for the spatial discretization. We show well posing results for the non-coupled and the coupled problems. At the end, we deal with the coupling of energy discretizations for the multigroup equations obtained by homogenization. Some theoretical results of the discretization of the velocity variable (well-posing of problems), which do not deal directly with the purposes of coupling, are presented in the annexes. (author)

Coupled-channels analyses for 9,11Li + 208Pb fusion reactions with multi-neutron transfer couplings

Science.gov (United States)

Choi, Ki-Seok; Cheoun, Myung-Ki; So, W. Y.; Hagino, K.; Kim, K. S.

2018-05-01

We discuss the role of two-neutron transfer processes in the fusion reaction of the 9,11Li + 208Pb systems. We first analyze the 9Li + 208Pb reaction by taking into account the coupling to the 7Li + 210Pb channel. To this end, we assume that two neutrons are directly transferred to a single effective channel in 210Pb and solve the coupled-channels equations with the two channels. By adjusting the coupling strength and the effective Q-value, we successfully reproduce the experimental fusion cross sections for this system. We then analyze the 11Li + 208Pb reaction in a similar manner, that is, by taking into account three effective channels with 11Li + 208Pb, 9Li + 210Pb, and 7Li + 212Pb partitions. In order to take into account the halo structure of the 11Li nucleus, we construct the potential between 11Li and 208Pb with a double folding procedure, while we employ a Woods-Saxon type potential with the global Akyüz-Winther parameters for the other channels. Our calculation indicates that the multiple two-neutron transfer process plays a crucial role in the 11Li + 208Pb fusion reaction at energies around the Coulomb barrier.

TORT/MCNP coupling method for the calculation of neutron flux around a core of BWR

International Nuclear Information System (INIS)

Kurosawa, M.

2005-01-01

For the analysis of BWR neutronics performance, accurate data are required for neutron flux distribution over the In-Reactor Pressure Vessel equipments taking into account the detailed geometrical arrangement. The TORT code can calculate neutron flux around a core of BWR in a three-dimensional geometry model, but has difficulties in fine geometrical modelling and lacks huge computer resource. On the other hand, the MCNP code enables the calculation of the neutron flux with a detailed geometry model, but requires very long sampling time to give enough number of particles. Therefore, a TORT/MCNP coupling method has been developed to eliminate the two problems mentioned above in each code. In this method, the TORT code calculates angular flux distribution on the core surface and the MCNP code calculates neutron spectrum at the points of interest using the flux distribution. The coupling method will be used as the DOT-DOMINO-MORSE code system. This TORT/MCNP coupling method was applied to calculate the neutron flux at points where induced radioactivity data were measured for 54 Mn and 60 Co and the radioactivity calculations based on the neutron flux obtained from the above method were compared with the measured data. (authors)

TORT/MCNP coupling method for the calculation of neutron flux around a core of BWR.

Science.gov (United States)

Kurosawa, Masahiko

2005-01-01

For the analysis of BWR neutronics performance, accurate data are required for neutron flux distribution over the In-Reactor Pressure Vessel equipments taking into account the detailed geometrical arrangement. The TORT code can calculate neutron flux around a core of BWR in a three-dimensional geometry model, but has difficulties in fine geometrical modelling and lacks huge computer resource. On the other hand, the MCNP code enables the calculation of the neutron flux with a detailed geometry model, but requires very long sampling time to give enough number of particles. Therefore, a TORT/MCNP coupling method has been developed to eliminate the two problems mentioned above in each code. In this method, the TORT code calculates angular flux distribution on the core surface and the MCNP code calculates neutron spectrum at the points of interest using the flux distribution. The coupling method will be used as the DOT-DOMINO-MORSE code system. This TORT/MCNP coupling method was applied to calculate the neutron flux at points where induced radioactivity data were measured for 54Mn and 60Co and the radioactivity calculations based on the neutron flux obtained from the above method were compared with the measured data.

Coupling of 3D neutronics models with the system code ATHLET

International Nuclear Information System (INIS)

Langenbuch, S.; Velkov, K.

1999-01-01

The system code ATHLET for plant transient and accident analysis has been coupled with 3D neutronics models, like QUABOX/CUBBOX, for the realistic evaluation of some specific safety problems under discussion. The considerations for the coupling approach and its realization are discussed. The specific features of the coupled code system established are explained and experience from first applications is presented. (author)

Intensity enhancement of cold neutrons from a coupled liquid-hydrogen moderator for pulsed cold neutron sources

International Nuclear Information System (INIS)

Ogawa, Y.; Kiyanagi, Y.; Kosugi, N.; Iwasa, H.; Furusaka, M.; Watanabe, N.

1999-01-01

In order to obtain higher cold neutron intensity from a coupled liquid-hydrogen moderator with a premoderator for pulsed cold neutron sources, we examined a partial enhancement method, namely, narrow beam extraction for both a flat liquid-hydrogen moderator and a single-groove one. Combined with the narrow beam extraction, which is especially suitable for small-angle scattering and neutron reflectometry experiments, a single-groove moderator provides higher intensity, by about 30%, than a flat-surface moderator at the region of interest on a viewed surface. The effect of double-side beam extraction from such moderators on the intensity gain factor is also discussed. (author)

Neutron Imaging with Timepix Coupled Lithium Indium Diselenide

Directory of Open Access Journals (Sweden)

Elan Herrera

2017-12-01

Full Text Available The material lithium indium diselenide, a single crystal neutron sensitive semiconductor, has demonstrated its capabilities as a high resolution imaging device. The sensor was prepared with a 55 μ m pitch array of gold contacts, designed to couple with the Timepix imaging ASIC. The resulting device was tested at the High Flux Isotope Reactor, demonstrating a response to cold neutrons when enriched in 95% 6 Li. The imaging system performed a series of experiments resulting in a <200 μ m resolution limit with the Paul Scherrer Institute (PSI Siemens star mask and a feature resolution of 34 μ m with a knife-edge test. Furthermore, the system was able to resolve the University of Tennessee logo inscribed into a 3D printed 1 cm 3 plastic block. This technology marks the application of high resolution neutron imaging using a direct readout semiconductor.

CRISSUE-S, Neutronics/Thermal-hydraulics Coupling in LWR Technology

International Nuclear Information System (INIS)

D'Auria, Francesco; Bousbia Salah, Anis; Galassi, G.M.; Vedovi, Juswald; Van Goethem, Georges; Hadek, Jan; Macek, Jiri; Rindelhardt, Udo; Rohde, Ulrich; Ahnert Iglesias, Carol; Aragones Beltran, Jose Maria; Reventos, Francesc; Cuadra, Arantxa; Gago, Jose Luis; Verdu, Gumersindo; Miro, Rafael; Ginestar, Damian; Sanchez, Ana Maria; Sjoberg, Anders; Yitbarek, M.; Sandervag, Oddbjoern; Garis, Ninos; Frid, Wiktor; Panayotov, Dobromir; Ivanov, Kostadin; Uddin, Rizwan; Sartori, Enrico

2004-01-01

Description: The CRISSUE-S project was created with the aim of re-evaluating fundamental technical issues in the technology of LWRs. Specifically, the project seeks to address the interactions between neutron kinetics and thermal-hydraulics that affect neutron moderation and influence the accident performance of the NPPs. This is undertaken in the light of the advanced computational tools that are readily available to the scientific community today. Specifically, the CRISSUE-S activity deals with the control of fission power and the use of high burn up fuel; these topics are part of the EC Work Programme as well as that of other international organisations such as the OECD/NEA and the IAEA. The problems of evaluating reactivity induced accident (RIA) consequences and eventually deciding the possibility of NPP prolongation must be addressed and resolved. RIA constitutes one of the most important of the ?less-resolved? safety issues, and treating this problem may have huge positive financial, social and environmental impacts. Public acceptance of nuclear technology implies that problems such as these be satisfactorily resolved. Cross-disciplinary (regulators, industry, utilities and research bodies) interaction and co operation within CRISSUE-S provides results which can directly and immediately be beneficial to EU industry. Co-operation at an international level: the participation of the EU, former Eastern European countries, the USA, and observers from Japan testify to the broad interest these problems engender. Competencies in broad areas such as thermal-hydraulics, neutronics and fuel, overall system design and reactor surveillance are needed to address the problems that are posed here. Excellent expertise is available in specific areas, while limited knowledge exists in the interface zones of those areas, e.g. in the coupling between thermal-hydraulics and neutronics. In general terms, the activities carried out and described here aim at exploiting available

Coupled 3D neutronics/thermal hydraulics modeling of the SAFARI-1 MTR

International Nuclear Information System (INIS)

Rosenkrantz, Adam; Avramova, Maria; Ivanov, Kostadin; Prinsloo, Rian; Botes, Danniëll; Elsakhawy, Khalid

2014-01-01

Highlights: • Development of 3D coupled neutronics/thermal–hydraulic model of SAFARI-1. • Verification of 3D steady-state NEM based neutronics model for SAFARI-1. • Verification of 3D COBRA-TF based thermal–hydraulic model of SAFARI-1. • Quantification of the effect of correct modeling of thermal–hydraulic feedback. - Abstract: The purpose of this study was to develop a coupled accurate multi-physics model of the SAFARI-1 Material Testing Reactor (MTR), a facility that is used for both research and the production of medical isotopes. The model was developed as part of the SAFARI-1 benchmarking project as a cooperative effort between the Pennsylvania State University (PSU) and the South African Nuclear Energy Corporation (Necsa). It was created using a multi-physics coupling of state of the art nuclear reactor simulation tools, consisting of a neutronics code and a thermal hydraulics code. The neutronics tool used was the PSU code NEM, and the results from this component were verified using the Necsa neutronics code OSCAR-4, which is utilized for SAFARI-1 core design and fuel management. On average, the multiplication factors of the neutronics models agreed to within 5 pcm and the radial assembly-averaged powers agreed to within 0.2%. The thermal hydraulics tool used was the PSU version of COBRA-TF (CTF) sub-channel code, and the results of this component were verified against another thermal hydraulics code, the RELAP5-3D system code, used at Necsa for thermal–hydraulics analysis of SAFARI-1. Although only assembly-averaged results from RELAP5-3D were available, they fell within the range of values for the corresponding assemblies in the comprehensive CTF solution. This comparison allows for the first time to perform a quantification of steady-state errors for a low-powered MTR with an advanced thermal–hydraulic code such as CTF on a per-channel basis as compared to simpler and coarser-mesh RELAP5-3D modeling. Additionally, a new cross section

3D neutronic codes coupled with thermal-hydraulic system codes for PWR, and BWR and VVER reactors

Energy Technology Data Exchange (ETDEWEB)

Langenbuch, S.; Velkov, K. [GRS, Garching (Germany); Lizorkin, M. [Kurchatov-Institute, Moscow (Russian Federation)] [and others

1997-07-01

This paper describes the objectives of code development for coupling 3D neutronics codes with thermal-hydraulic system codes. The present status of coupling ATHLET with three 3D neutronics codes for VVER- and LWR-reactors is presented. After describing the basic features of the 3D neutronic codes BIPR-8 from Kurchatov-Institute, DYN3D from Research Center Rossendorf and QUABOX/CUBBOX from GRS, first applications of coupled codes for different transient and accident scenarios are presented. The need of further investigations is discussed.

Methodology of Continuous-Energy Adjoint Monte Carlo for Neutron, Photon, and Coupled Neutron-Photon Transport

International Nuclear Information System (INIS)

Hoogenboom, J. Eduard

2003-01-01

Adjoint Monte Carlo may be a useful alternative to regular Monte Carlo calculations in cases where a small detector inhibits an efficient Monte Carlo calculation as only very few particle histories will cross the detector. However, in general purpose Monte Carlo codes, normally only the multigroup form of adjoint Monte Carlo is implemented. In this article the general methodology for continuous-energy adjoint Monte Carlo neutron transport is reviewed and extended for photon and coupled neutron-photon transport. In the latter cases the discrete photons generated by annihilation or by neutron capture or inelastic scattering prevent a direct application of the general methodology. Two successive reaction events must be combined in the selection process to accommodate the adjoint analog of a reaction resulting in a photon with a discrete energy. Numerical examples illustrate the application of the theory for some simplified problems

Light-water-reactor coupled neutronic and thermal-hydraulic codes

International Nuclear Information System (INIS)

Diamond, D.J.

1982-01-01

An overview is presented of computer codes that model light water reactor cores with coupled neutronics and thermal-hydraulics. This includes codes for transient analysis and codes for steady state analysis which include fuel depletion and fission product buildup. Applications in nuclear design, reactor operations and safety analysis are given and the major codes in use in the USA are identified. The neutronic and thermal-hydraulic methodologies and other code features are outlined for three steady state codes (PDQ7, NODE-P/B and SIMULATE) and four dynamic codes (BNL-TWIGL, MEKIN, RAMONA-3B, RETRAN-02). Speculation as to future trends with such codes is also presented

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.)

Sensitivity of neutron scattering properties to the coupling to giant resonances

International Nuclear Information System (INIS)

Delaroche, J.P.; Guss, P.P.; Floyd, C.E.; Walter, R.L.; Tornow, W.

1983-01-01

Extended coupled channels calculations have been performed for neutron elastic and inelastic cross sections and analyzing powers for vibrational nuclei with 40 208 Pb and 12 MeV results for 40 Ca are illustrated here

Experimental validation of GADRAS's coupled neutron-photon inverse radiation transport solver

International Nuclear Information System (INIS)

Mattingly, John K.; Mitchell, Dean James; Harding, Lee T.

2010-01-01

Sandia National Laboratories has developed an inverse radiation transport solver that applies nonlinear regression to coupled neutron-photon deterministic transport models. The inverse solver uses nonlinear regression to fit a radiation transport model to gamma spectrometry and neutron multiplicity counting measurements. The subject of this paper is the experimental validation of that solver. This paper describes a series of experiments conducted with a 4.5 kg sphere of α-phase, weapons-grade plutonium. The source was measured bare and reflected by high-density polyethylene (HDPE) spherical shells with total thicknesses between 1.27 and 15.24 cm. Neutron and photon emissions from the source were measured using three instruments: a gross neutron counter, a portable neutron multiplicity counter, and a high-resolution gamma spectrometer. These measurements were used as input to the inverse radiation transport solver to evaluate the solver's ability to correctly infer the configuration of the source from its measured radiation signatures.

Coupled full core neutron transport/CFD simulations of pressurized water reactors

International Nuclear Information System (INIS)

Kochunas, B.; Stimpson, S.; Collins, B.; Downar, T.; Brewster, R.; Baglietto, E.; Yan, J.

2012-01-01

Recently as part of the CASL project, a capability to perform 3D whole-core coupled neutron transport and computational fluid dynamics (CFD) calculations was demonstrated. This work uses the 2D/1D transport code DeCART and the commercial CFD code STAR-CCM+. It builds on previous CASL work demonstrating coupling for smaller spatial domains. The coupling methodology is described along with the problem simulated and results are presented for fresh hot full power conditions. An additional comparison is made to an equivalent model that uses lower order T/H feedback to assess the importance and cost of high fidelity feedback to the neutronics problem. A simulation of a quarter core Combustion Engineering (CE) PWR core was performed with the coupled codes using a Fixed Point Gauss-Seidel iteration technique. The total approximate calculation requirements are nearly 10,000 CPU hours and 1 TB of memory. The problem took 6 coupled iterations to converge. The CFD coupled model and low order T/H feedback model compared well for global solution parameters, with a difference in the critical boron concentration and average outlet temperature of 14 ppm B and 0.94 deg. C, respectively. Differences in the power distribution were more significant with maximum relative differences in the core-wide pin peaking factor (Fq) of 5.37% and average relative differences in flat flux region power of 11.54%. Future work will focus on analyzing problems more relevant to CASL using models with less approximations. (authors)

Convergence analysis of neutronic/thermohydraulic coupling behavior of SCWR

International Nuclear Information System (INIS)

Liu, Shichang; Cai, Jiejin

2013-01-01

The neutronic/thermohydraulic coupling (N–T coupling) calculations play an important role in core design and stability analysis. The traditional iterative method is not applicable for some new reactors (such as supercritical water-cooled reactor) which have intense N–T coupling behavior. In this paper, the mathematical model of N–T coupling based on fixed point theory is established firstly, with the convergent criterion, which can show the real-time convergence situation of iteration. Secondly, the self-adaptive relaxation factor and corresponding algorithm are proposed. Thirdly, the convergence analysis of the method of self-adaptive relaxation factor and common relaxation iteration has been performed, based on three calculation examples of SCWR fuel assembly. The results show that the proposed algorithm can efficiently reduce the calculation time and be adapted to different coupling cases and different initial distribution. It is easy to program, providing convenience for reactor design and analysis. This research also provides the theoretical basis for further study of N–T coupling behavior of new reactors such as SCWR

(Anti)-ferromagnetic coupling in Fe/Si multilayers from polarized neutron reflectomy

NARCIS (Netherlands)

Fredrikze, H.; Graaf, van der A; Valkier, M.; Kohlhepp, J.T.; Broeder, den F.J.A.

1997-01-01

Polarized neutron reflectometry data on Fe/Si multilayers are interpreted using strongly depth-dependent magnetization in the Fe layers. This behaviour is ascribed to a depth-dependent mixture of ferromagnetic and anti-ferromagnetic coupled regions in the sample.

A Newton-based Jacobian-free approach for neutronic-Monte Carlo/thermal-hydraulic static coupled analysis

International Nuclear Information System (INIS)

Mylonakis, Antonios G.; Varvayanni, M.; Catsaros, N.

2017-01-01

Highlights: •A Newton-based Jacobian-free Monte Carlo/thermal-hydraulic coupling approach is introduced. •OpenMC is coupled with COBRA-EN with a Newton-based approach. •The introduced coupling approach is tested in numerical experiments. •The performance of the new approach is compared with the traditional “serial” coupling approach. -- Abstract: In the field of nuclear reactor analysis, multi-physics calculations that account for the bonded nature of the neutronic and thermal-hydraulic phenomena are of major importance for both reactor safety and design. So far in the context of Monte-Carlo neutronic analysis a kind of “serial” algorithm has been mainly used for coupling with thermal-hydraulics. The main motivation of this work is the interest for an algorithm that could maintain the distinct treatment of the involved fields within a tight coupling context that could be translated into higher convergence rates and more stable behaviour. This work investigates the possibility of replacing the usually used “serial” iteration with an approximate Newton algorithm. The selected algorithm, called Approximate Block Newton, is actually a version of the Jacobian-free Newton Krylov method suitably modified for coupling mono-disciplinary solvers. Within this Newton scheme the linearised system is solved with a Krylov solver in order to avoid the creation of the Jacobian matrix. A coupling algorithm between Monte-Carlo neutronics and thermal-hydraulics based on the above-mentioned methodology is developed and its performance is analysed. More specifically, OpenMC, a Monte-Carlo neutronics code and COBRA-EN, a thermal-hydraulics code for sub-channel and core analysis, are merged in a coupling scheme using the Approximate Block Newton method aiming to examine the performance of this scheme and compare with that of the “traditional” serial iterative scheme. First results show a clear improvement of the convergence especially in problems where significant

Analysis of coupled neutron-gamma radiations, applied to shieldings in multigroup albedo method

International Nuclear Information System (INIS)

Dunley, Leonardo Souza

2002-01-01

The principal mathematical tools frequently available for calculations in Nuclear Engineering, including coupled neutron-gamma radiations shielding problems, involve the full Transport Theory or the Monte Carlo techniques. The Multigroup Albedo Method applied to shieldings is characterized by following the radiations through distinct layers of materials, allowing the determination of the neutron and gamma fractions reflected from, transmitted through and absorbed in the irradiated media when a neutronic stream hits the first layer of material, independently of flux calculations. Then, the method is a complementary tool of great didactic value due to its clarity and simplicity in solving neutron and/or gamma shielding problems. The outstanding results achieved in previous works motivated the elaboration and the development of this study that is presented in this dissertation. The radiation balance resulting from the incidence of a neutronic stream into a shielding composed by 'm' non-multiplying slab layers for neutrons was determined by the Albedo method, considering 'n' energy groups for neutrons and 'g' energy groups for gammas. It was taken into account there is no upscattering of neutrons and gammas. However, it was considered that neutrons from any energy groups are able to produce gammas of all energy groups. The ANISN code, for an angular quadrature order S 2 , was used as a standard for comparison of the results obtained by the Albedo method. So, it was necessary to choose an identical system configuration, both for ANISN and Albedo methods. This configuration was six neutron energy groups and eight gamma energy groups, using three slab layers (iron aluminum - manganese). The excellent results expressed in comparative tables show great agreement between the values determined by the deterministic code adopted as standard and, the values determined by the computational program created using the Albedo method and the algorithm developed for coupled neutron

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

An immersed body method for coupled neutron transport and thermal hydraulic simulations of PWR assemblies

International Nuclear Information System (INIS)

Jewer, S.; Buchan, A.G.; Pain, C.C.; Cacuci, D.G.

2014-01-01

Highlights: • A new method of coupled radiation transport, heat and momentum exchanges on fluids, and heat transfer simulations. • Simulation of the thermal hydraulics and radiative properties within whole PWR assemblies. • An immersed body method for modelling complex solid domains on practical computational meshes. - Abstract: A recently developed immersed body method is adapted and used to model a typical pressurised water reactor (PWR) fuel assembly. The approach is implemented with the numerical framework of the finite element, transient criticality code, FETCH which is composed of the neutron transport code, EVENT, and the CFD code, FLUIDITY. Within this framework the neutron transport equation, Navier–Stokes equations and a fluid energy conservation equation are solved in a coupled manner on a coincident structured or unstructured mesh. The immersed body method has been used to model the solid fuel pins. The key feature of this method is that the fluid/neutronic domain and the solid domain are represented by overlapping and non-conforming meshes. The main difficulty of this approach, for which a solution is proposed in this work, is the conservative mapping of the energy and momentum exchange between the fluid/neutronic mesh and the solid fuel pin mesh. Three numerical examples are presented which include a validation of the fuel pin submodel against an analytical solution; an uncoupled (no neutron transport solution) PWR fuel assembly model with a specified power distribution which was validated against the COBRA-EN subchannel analysis code; and finally a coupled model of a PWR fuel assembly with reflective neutron boundary conditions. Coupling between the fluid and neutron transport solutions is through the nuclear cross sections dependence on Doppler fuel temperature, coolant density and temperature, which was taken into account by using pre-calculated cross-section lookup tables generated using WIMS9a. The method was found to show good agreement

Strongly coupled chameleon fields: Possible test with a neutron Lloyd's mirror interferometer

International Nuclear Information System (INIS)

Pokotilovski, Yu.N.

2013-01-01

The consideration of possible neutron Lloyd's mirror interferometer experiment to search for strongly coupled chameleon fields is presented. The chameleon scalar fields were proposed to explain the acceleration of expansion of the Universe. The presence of a chameleon field results in a change of a particle's potential energy in vicinity of a massive body. This interaction causes a phase shift of neutron waves in the interferometer. The sensitivity of the method is estimated

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)

Thermal neutron imaging through XRQA2 GAFCHROMIC films coupled with a cadmium radiator

Energy Technology Data Exchange (ETDEWEB)

Sacco, D. [INFN – LNF, Via E. Fermi n.40, Frascati, 00044 Roma (Italy); INAIL – DIT, Via di Fontana Candida n.1, 00040 Monteporzio Catone (Italy); Bedogni, R., E-mail: roberto.bedogni@lnf.infn.it [INFN – LNF, Via E. Fermi n.40, Frascati, 00044 Roma (Italy); Bortot, D. [Politecnico di Milano, Dipartimento di Energia, Via La Masa 34, 20156 Milano (Italy); INFN – Milano, Via Celoria16, 20133 Milano (Italy); Palomba, M. [ENEA Casaccia, Via Anguillarese, 301, S. Maria di Galeria, 00123 Roma (Italy); Pola, A. [Politecnico di Milano, Dipartimento di Energia, Via La Masa 34, 20156 Milano (Italy); INFN – Milano, Via Celoria16, 20133 Milano (Italy); Introini, M.V.; Lorenzoli, M. [Politecnico di Milano, Dipartimento di Energia, Via La Masa 34, 20156 Milano (Italy); Gentile, A. [INFN – LNF, Via E. Fermi n.40, Frascati, 00044 Roma (Italy); Strigari, L. [Laboratory of Medical Physics, Regina Elena National Cancer Institute, Via E. Chianesi 53, 00144 Roma (Italy); Pressello, C. [Department of Medical Physics, Azienda Ospedaliera San Camillo Forlanini, Circonvallazione Gianicolense 87, 00152 Roma (Italy); Soriani, A. [Laboratory of Medical Physics, Regina Elena National Cancer Institute, Via E. Chianesi 53, 00144 Roma (Italy); Gómez-Ros, J.M. [INFN – LNF, Via E. Fermi n.40, Frascati, 00044 Roma (Italy); CIEMAT, Av. Complutense 40, 28040 Madrid (Spain)

2015-10-21

A simple and inexpensive method to perform passive thermal neutron imaging on large areas was developed on the basis of XRQA2 GAFCHROMIC films, commonly employed for quality assurance in radiology. To enhance their thermal neutron response, the sensitive face of film was coupled with a 1 mm thick cadmium radiator, forming a sandwich. By exchanging the order of Cd filter and sensitive film with respect to the incident neutron beam direction, two different configurations (beam-Cd-film and beam-film-Cd) were identified. These configurations were tested at thermal neutrons fluence values in the range 10{sup 9}–10{sup 10} cm{sup −2}, using the ex-core radial thermal neutron column of the ENEA Casaccia – TRIGA reactor. The results are presented in this work.

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)

The generation, validation and testing of a coupled 219-group neutron 36-group gamma ray AMPX-II library

International Nuclear Information System (INIS)

Panini, G.C.; Siciliano, F.; Lioi, A.

1987-01-01

The main characteristics of a P 3 coupled 219-group neutron 36-group gamma-ray library in the AMPX-II Master Interface Format obtained processing ENDF/B-IV data by means of various AMPX-II System modules are presented in this note both for the more reprocessing aspects and features of the generated component files-neutrons, photon and secondary gamma-ray production cross sections. As far as the neutron data are concerned there is the avaibility of 186 data sets regarding most significant fission products. Results of the additional validation of the neutron data pertaining to eighteen benchmark experiments are also given. Some calculational tests on both neutron and coupled data emphasize the important role of the secondary gamma-ray data in nuclear criticality safety calculations

AMPX: a modular code system for generating coupled multigroup neutron-gamma libraries from ENDF/B

Energy Technology Data Exchange (ETDEWEB)

Greene, N.M.; Lucius, J.L.; Petrie, L.M.; Ford, W.E. III; White, J.E.; Wright, R.Q.

1976-03-01

AMPX is a modular system for producing coupled multigroup neutron-gamma cross section sets. Basic neutron and gamma cross-section data for AMPX are obtained from ENDF/B libraries. Most commonly used operations required to generate and collapse multigroup cross-section sets are provided in the system. AMPX is flexibly dimensioned; neutron group structures, and gamma group structures, and expansion orders to represent anisotropic processes are all arbitrary and limited only by available computer core and budget. The basic processes provided will (1) generate multigroup neutron cross sections; (2) generate multigroup gamma cross sections; (3) generate gamma yields for gamma-producing neutron interactions; (4) combine neutron cross sections, gamma cross sections, and gamma yields into final ''coupled sets''; (5) perform one-dimensional discrete ordinates transport or diffusion theory calculations for neutrons and gammas and, on option, collapse the cross sections to a broad-group structure, using the one-dimensional results as weighting functions; (6) plot cross sections, on option, to facilitate the ''evaluation'' of a particular multigroup set of data; (7) update and maintain multigroup cross section libraries in such a manner as to make it not only easy to combine new data with previously processed data but also to do it in a single pass on the computer; and (8) output multigroup cross sections in convenient formats for other codes. (auth)

Coupled neutronic/thermal-hydraulic analysis of the HPLWR three pass core

International Nuclear Information System (INIS)

Monti, Lanfranco; Starflinger, Joerg; Schulenberg, Thomas

2008-01-01

The High Performance Light Water Reactor is an innovative Gen-IV reactor cooled and moderated with water at supercritical pressure. The three pass core concept has been proposed to reduce peaking factors, i.e. hot-channel effects, and it further increases the core heterogeneity, which is mainly due to pronounced water density reduction. For this kind of nuclear reactor, the significant feedbacks - which exist between the properties of the components and the power generation rate - can not be neglected and require a coupled Neutronic/Thermal-Hydraulic analysis even for steady state conditions. The main goal of this paper is to present the developed tool for coupled analyses of the HPLWR. Two state-of-the-art codes have been chosen for Thermal-Hydraulic and Neutronic core analyses, namely TRACE and ERANOS, and they have been coupled with in an iterative procedure in which they are run in series until a steady state condition has been reached. In the simplifying assumptions of uniform enrichment distribution, zero burn-up and ignoring the effect of the control rods, the obtained steady state condition will be discussed and a core power map, flow rate redistribution as well as water and fuel temperature variations will be presented. (author)

Coupled multi-group neutron photon transport for the simulation of high-resolution gamma-ray spectroscopy applications

Energy Technology Data Exchange (ETDEWEB)

Burns, Kimberly A. [Georgia Inst. of Technology, Atlanta, GA (United States)

2009-08-01

The accurate and efficient simulation of coupled neutron-photon problems is necessary for several important radiation detection applications. Examples include the detection of nuclear threats concealed in cargo containers and prompt gamma neutron activation analysis for nondestructive determination of elemental composition of unknown samples.

LO-phonon and plasmon coupling in neutron-transmutation-doped GaAs

International Nuclear Information System (INIS)

Kuriyama, K.; Sakai, K.; Okada, M.

1996-01-01

Coupling between the longitudinal-optic (LO) phonon mode and the longitudinal plasma mode in neutron-transmutation-doped (NTD) semi-insulating GaAs was studied using Raman-scattering spectroscopy and a Fourier-transform infrared spectrometer. When the electron concentration due to the activation of NTD impurities (Ge Ga and Se As ) approaches ∼8x10 16 cm -3 , the LO-phonon endash plasmon coupling is observed. This behavior is consistent with the free-electron absorption due to the activation of NTD impurities in samples annealed above 600 degree C. copyright 1996 The American Physical Society

Neutronic characteristics of coupled moderator proposed in integrated model

International Nuclear Information System (INIS)

Teshigawara, Makoto; Meigo, Shin-ichiro; Sakata, Hideaki; Kai, Tetsuya; Harada, Masahide; Ikeda, Yujiro; Watanabe, Noboru

2001-05-01

A pulsed spallation source for the materials science and the life science is currently developing for its construction in the High Intensity Proton Accelerator Project proposed jointly by the Japan Atomic Energy Research Institute (JAERI) and the High Energy Accelerator Research Organization (KEK). This report presents the analytical results of the neutronic characteristics of the coupled moderator based on the analytical results obtained by using an integrated model which has established on the extensive neutronic and technical study. Total heat deposition in a hydrogen (H 2 ) moderator working as the main moderator was about 420 W/MW. Maximum nuclear heat density in the H 2 moderator was about 1 W/cm 3 /MW. Also total heat deposition in a premoderator was about 9.2 kW/MW. The heat density of the premoderator was comparable to that of the moderator vessel made of aluminum alloy. The heat density of the premoderator and the moderator vessel is about 1.2-2 times higher than that of the hydrogen moderator. The temperature from 300 K to 400 K of the premoderator did not affect on neutron intensity of the H 2 moderator. This suggested an engineering advantage on the thermal and hydraulic design. 6000 or 7000 type of a aluminum alloy was considered from the viewpoint of the neutron beam transmission. The proton beams scattered by the proton beam window did not affect on the nuclear heating in the H 2 moderator. The heat deposition in the H 2 moderator and the neutron intensity of the H 2 moderator did not depend on the proton beam profile but it did on the distance between the proton beam and the moderator. (author)

Multigroup and coupled forward-adjoint Monte Carlo calculation efficiencies for secondary neutron doses from proton beams

International Nuclear Information System (INIS)

Kelsey IV, Charles T.; Prinja, Anil K.

2011-01-01

We evaluate the Monte Carlo calculation efficiency for multigroup transport relative to continuous energy transport using the MCNPX code system to evaluate secondary neutron doses from a proton beam. We consider both fully forward simulation and application of a midway forward adjoint coupling method to the problem. Previously we developed tools for building coupled multigroup proton/neutron cross section libraries and showed consistent results for continuous energy and multigroup proton/neutron transport calculations. We observed that forward multigroup transport could be more efficient than continuous energy. Here we quantify solution efficiency differences for a secondary radiation dose problem characteristic of proton beam therapy problems. We begin by comparing figures of merit for forward multigroup and continuous energy MCNPX transport and find that multigroup is 30 times more efficient. Next we evaluate efficiency gains for coupling out-of-beam adjoint solutions with forward in-beam solutions. We use a variation of a midway forward-adjoint coupling method developed by others for neutral particle transport. Our implementation makes use of the surface source feature in MCNPX and we use spherical harmonic expansions for coupling in angle rather than solid angle binning. The adjoint out-of-beam transport for organs of concern in a phantom or patient can be coupled with numerous forward, continuous energy or multigroup, in-beam perturbations of a therapy beam line configuration. Out-of-beam dose solutions are provided without repeating out-of-beam transport. (author)

Accuracy and Efficiency of a Coupled Neutronics and Thermal Hydraulics Model

International Nuclear Information System (INIS)

Pope, Michael A.; Mousseau, Vincent A.

2009-01-01

The accuracy requirements for modern nuclear reactor simulation are steadily increasing due to the cost and regulation of relevant experimental facilities. Because of the increase in the cost of experiments and the decrease in the cost of simulation, simulation will play a much larger role in the design and licensing of new nuclear reactors. Fortunately as the work load of simulation increases, there are better physics models, new numerical techniques, and more powerful computer hardware that will enable modern simulation codes to handle this larger workload. This manuscript will discuss a numerical method where the six equations of two-phase flow, the solid conduction equations, and the two equations that describe neutron diffusion and precursor concentration are solved together in a tightly coupled, nonlinear fashion for a simplified model of a nuclear reactor core. This approach has two important advantages. The first advantage is a higher level of accuracy. Because the equations are solved together in a single nonlinear system, the solution is more accurate than the traditional 'operator split' approach where the two-phase flow equations are solved first, the heat conduction is solved second and the neutron diffusion is solved third, limiting the temporal accuracy to 1st order because the nonlinear coupling between the physics is handled explicitly. The second advantage of the method described in this manuscript is that the time step control in the fully implicit system can be based on the timescale of the solution rather than a stability-based time step restriction like the material Courant. Results are presented from a simulated control rod movement and a rod ejection that address temporal accuracy for the fully coupled solution and demonstrate how the fastest timescale of the problem can change between the state variables of neutronics, conduction and two-phase flow during the course of a transient.

Accuracy and Efficiency of a Coupled Neutronics and Thermal Hydraulics Model

International Nuclear Information System (INIS)

Vincent A. Mousseau; Michael A. Pope

2007-01-01

The accuracy requirements for modern nuclear reactor simulation are steadily increasing due to the cost and regulation of relevant experimental facilities. Because of the increase in the cost of experiments and the decrease in the cost of simulation, simulation will play a much larger role in the design and licensing of new nuclear reactors. Fortunately as the work load of simulation increases, there are better physics models, new numerical techniques, and more powerful computer hardware that will enable modern simulation codes to handle the larger workload. This manuscript will discuss a numerical method where the six equations of two-phase flow, the solid conduction equations, and the two equations that describe neutron diffusion and precursor concentration are solved together in a tightly coupled, nonlinear fashion for a simplified model of a nuclear reactor core. This approach has two important advantages. The first advantage is a higher level of accuracy. Because the equations are solved together in a single nonlinear system, the solution is more accurate than the traditional 'operator split' approach where the two-phase flow equations are solved first, the heat conduction is solved second and the neutron diffusion is solved third, limiting the temporal accuracy to 1st order because the nonlinear coupling between the physics is handled explicitly. The second advantage of the method described in this manuscript is that the time step control in the fully implicit system can be based on the timescale of the solution rather than a stability-based time step restriction like the material Courant. Results are presented from a simulated control rod movement and a rod ejection that address temporal accuracy for the fully coupled solution and demonstrate how the fastest timescale of the problem can change between the state variables of neutronics, conduction and two-phase flow during the course of a transient

Comparison of multimesh hp-FEM to interpolation and projection methods for spatial coupling of thermal and neutron diffusion calculations

International Nuclear Information System (INIS)

Dubcova, Lenka; Solin, Pavel; Hansen, Glen; Park, HyeongKae

2011-01-01

Multiphysics solution challenges are legion within the field of nuclear reactor design and analysis. One major issue concerns the coupling between heat and neutron flow (neutronics) within the reactor assembly. These phenomena are usually very tightly interdependent, as large amounts of heat are quickly produced with an increase in fission events within the fuel, which raises the temperature that affects the neutron cross section of the fuel. Furthermore, there typically is a large diversity of time and spatial scales between mathematical models of heat and neutronics. Indeed, the different spatial resolution requirements often lead to the use of very different meshes for the two phenomena. As the equations are coupled, one must take care in exchanging solution data between them, or significant error can be introduced into the coupled problem. We propose a novel approach to the discretization of the coupled problem on different meshes based on an adaptive multimesh higher-order finite element method (hp-FEM), and compare it to popular interpolation and projection methods. We show that the multimesh hp-FEM method is significantly more accurate than the interpolation and projection approaches considered in this study.

Pin level neutronic - thermal hydraulic two-way-coupling using DYN3D-SP3 and SUBCHANFLOW

International Nuclear Information System (INIS)

Torres, Armando Gomez; Espinoza, Victor Sanchez; Imke, Uwe; Juan, Rafael Macian

2011-01-01

Nowadays several Reactor Dynamic Codes, (RDC) are able to solve the diffusion equation or even the transport equation (SP3 approximation) considering feedback parameters coming from the thermalhydraulic (TH) core behavior. These kinds of codes (DYN3D, PARCS, among others) usually contain a 1D two phase flow thermalhydraulic model capable to pass them assembly averaged feedback parameters. At fuel assembly base this nodal coupling is completely a two way coupling. The Neutronic part calculates the mean power of the whole assembly and passes it to the TH part in order to actualize the heat source. In turn, the TH model passes the assembly-based feedback parameters to the neutronic code for actualizing the nodal cross sections. The process will be repeated until convergence. At pin level, the current situation is somehow different. Although the neutronic solver can pass the pin power distribution in every sub - node (pin distribution), the 1-D TH model will average the pin power distribution to assembly-based scale and will give back assembly averaged feedbacks to the neutronic part for cross sections up-date (one and a half way coupling), leading to information loss in the calculation. A new coupled program system DYNSUB was developed by coupling DYN3D-SP3 and SUBCHANFLOW at pin level. DYNSUB was used to analyze stationary PWR minicore problems at pin-level. The comparison of the Keff predicted by DYNSUB with the one calculated by DYN3D-SP3 (coarse TH solution) shows small differences of up to 26 pcm. Differences up to 4.5% were found in the radial distribution of the pin power. The local safety parameters such as cladding and fuel temperature predicted with DYNSUB shows larger deviations compared with the ones obtained with DYN3D-SP3. These differences may increase when analyzing transients. (author)

Coupled neutronics and thermal-hydraulics numerical simulations of a Molten Salt Fast Reactor (MSFR)

International Nuclear Information System (INIS)

Laureau, A.; Rubiolo, P.R.; Heuer, D.; Merle-Lucotte, E.; Brovchenko, M.

2013-01-01

Coupled neutronics and thermalhydraulic numerical analyses of a molten salt fast reactor (MSFR) are presented. These preliminary numerical simulations are carried-out using the Monte Carlo code MCNP and the Computation Fluid Dynamic code OpenFOAM. The main objectives of this analysis performed at steady-reactor conditions are to confirm the acceptability of the current neutronic and thermalhydraulic designs of the reactor, to study the effects of the reactor operating conditions on some of the key MSFR design parameters such as the temperature peaking factor. The effects of the precursor's motion on the reactor safety parameters such as the effective fraction of delayed neutrons have been evaluated. (authors)

Thermohydraulic stability coupled to the neutronic in a BWR; Estabilidad termohidraulica acoplada a la neutronica en un BWR

Energy Technology Data Exchange (ETDEWEB)

Calleros M, G.; Zapata Y, M.; Gomez H, R.A.; Mendez M, A. [Comision Federal de Electricidad, Central Nucleoelectrica de Laguna Verde, Carretera Cardel-Nautla Km. 42.5, Mpio. Alto Lucero, Veracruz (Mexico); Castlllo D, R. [ININ, Carretera Mexico-Toluca Km 36.5, La Marquesa, Estado de Mexico (Mexico)]. e-mail: gcm9acpp@cfe.gob.mx

2006-07-01

In a BWR type reactor the phenomenon of the nuclear fission is presented, in which are liberated in stochastic form neutrons, originating that the population of the same ones varies in statistic form around a mean value. This variation will cause that when the neutron flow impacts on the neutron detectors, its are had as a result neutron flow signals with fluctuations around an average value. In this article it is shown that it conforms it lapses the time, this variations in the neutron flow (and therefore, in the flow signal due only to the fission), they presented oscillations inside a stable range, which won't be divergent. Considering that the BWR is characterized because boiling phenomena are presented, which affect the moderation of the neutrons, additional variations will be had in the signal coming from the neutron detectors, with relationship to the fission itself, which will be influenced by the feedback of the moderator's reactivity and of the temperature of the fuel pellet. Also, as the BWR it has coupled control systems to maintain the coolant level one and of the thermal power of the reactor, for each control action it was affected the neutron population. This means that the reactor could end up straying of a stable state condition. By it previously described, the study of the thermohydraulic stability coupled to the neutronic is complex. In this work it is shown the phenomenology, the mathematical models and the theoretical behavior associated to the stability of the BWR type reactor; the variables that affect it are identified, the models that reproduce the behavior of the thermohydraulic stability coupled to the neutronic, the way to maintain stable the reactor and the instrumentation that can settle to detect and to suppress uncertainties is described. In particular, is make reference to the evolution of the methods to maintain the stability of the reactor and the detection system and suppression of uncertainties implemented in the

A time-dependent neutron transport model and its coupling to thermal-hydraulics

International Nuclear Information System (INIS)

Pautz, A.

2001-01-01

A new neutron transport code for time-dependent analyses of nuclear systems has been developed. The code system is based on the well-known Discrete Ordinates code DORT, which solves the steady-state neutron/photon transport equation in two dimensions for an arbitrary number of energy groups and the most common regular geometries. For the implementation of time-dependence a fully implicit first-order scheme was employed to minimize errors due to temporal discretization. This requires various modifications to the transport equation as well as the extensive use of elaborated acceleration mechanisms. The convergence criteria for fluxes, fission rates etc. had to be strongly tightened to ensure the reliability of results. To perform coupled analyses, an interface to the GRS system code ATHLET has been developed. The nodal power densities from the neutron transport code are passed to ATHLET to calculate thermal-hydraulic system parameters, e.g. fuel and coolant temperatures. These are in turn used to generate appropriate nuclear cross sections by interpolation of pre-calculated data sets for each time step. Finally, to demonstrate the transient capabilities of the coupled code system, the research reactor FRM-II has been analysed. Several design basis accidents were modelled, like the loss of off site power, loss of secondary heat sink and unintended control rod withdrawal. (author)

Strongly coupled chameleon fields: Possible test with a neutron Lloyd's mirror interferometer

Energy Technology Data Exchange (ETDEWEB)

Pokotilovski, Yu.N., E-mail: pokot@nf.jinr.ru [Joint Institute for Nuclear Research, 141980 Dubna, Moscow Region (Russian Federation)

2013-02-26

The consideration of possible neutron Lloyd's mirror interferometer experiment to search for strongly coupled chameleon fields is presented. The chameleon scalar fields were proposed to explain the acceleration of expansion of the Universe. The presence of a chameleon field results in a change of a particle's potential energy in vicinity of a massive body. This interaction causes a phase shift of neutron waves in the interferometer. The sensitivity of the method is estimated.

Investigation of the possibility to use a fine-mesh solver for resolving coupled neutronics and thermal-hydraulics

International Nuclear Information System (INIS)

Jareteg, K.; Vinai, P.; Demaziere, C.

2013-01-01

The development of a fine-mesh coupled neutronic/thermal-hydraulic solver is touched upon in this paper. The reported work investigates the feasibility of using finite volume techniques to discretize a set of conservation equations modeling neutron transport, fluid dynamics, and heat transfer within a single numerical tool. With the long-term objective of developing fine-mesh computing capabilities for a few selected fuel assemblies in a nuclear core, this preliminary study considers an infinite array of a single fuel assembly having a finite height. Thermal-hydraulic conditions close to the ones existing in PWRs are taken as a first test case. The neutronic modeling relies on the diffusion approximation in a multi-energy group formalism, with cross-sections pre-calculated and tabulated at the sub-pin level using a Monte Carlo technique. The thermal-hydraulics is based on the Navier-Stokes equations, complemented by an energy conservation equation. The non-linear coupling terms between the different conservation equations are fully resolved using classical iteration techniques. Early tests demonstrate that the numerical tool provides an unprecedented level of details of the coupled solution estimated within the same numerical tool and thus avoiding any external data transfer, using fully consistent models between the neutronics and the thermal-hydraulics. (authors)

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)

Whole core pin-by-pin coupled neutronic-thermal-hydraulic steady state and transient calculations using COBAYA3 code

International Nuclear Information System (INIS)

Jimenez, J.; Herrero, J. J.; Cuervo, D.; Aragones, J. M.

2010-10-01

Nowadays coupled 3-dimensional neutron kinetics and thermal-hydraulic core calculations are performed by applying a radial average channel approach using a meshing of one quarter of assembly in the best case. This approach does not take into account the subchannels effects due to the averaging of the physical fields and the loose of heterogeneity in the thermal-hydraulic model. Therefore the models do not have enough resolution to predict those subchannels effects which are important for the fuel design safety margins, because it is in the local scale, where we can search the hottest pellet or the maximum heat flux. The Polytechnic University of Madrid advanced multi-scale neutron-kinetics and thermal-hydraulics methodologies being implemented in COBAYA3 include domain decomposition by alternate core dissections for the local 3-dimensional fine-mesh scale problems (pin cells/subchannels) and an analytical nodal diffusion solver for the coarse mesh scale coupled with the thermal-hydraulic using a model of one channel per assembly or per quarter of assembly. In this work, we address the domain decomposition by the alternate core dissections methodology applied to solve coupled 3-dimensional neutronic-thermal-hydraulic problems at the fine-mesh scale. The neutronic-thermal-hydraulic coupling at the cell-subchannel scale allows the treatment of the effects of the detailed thermal-hydraulic feedbacks on cross-sections, thus resulting in better estimates of the local safety margins at the pin level. (Author)

Analysis of neutron cross sections using the coupled-channel theory

International Nuclear Information System (INIS)

Tanaka, Shigeya

1975-01-01

Fast neutron total and scattering cross sections calculated with the coupled-channel theory and the spherical optical model are compared with experimental data. The optical-potential parameters used in both the calculations were obtained from comparison of calculations with scattering data for 209 Bi. The calculations for total cross sections were made for thirty-five nuclides from 23 Na to 239 Pu in the energy range of 0.25 to 15 MeV, and good results were obtained with the coupled-channel calculations. The comparisons of the calculations with the elastic data for about twenty nuclides were made at incident energies of 8 and 14 MeV. In general, the coupled-channel calculations at 8 MeV have given better agreements with the experimental data than the spherical optical-model calculations. At 14 MeV, differences between both the calculations were small. The analysis was also made for the elastic and inelastic scattering by several nuclei such as Fe, Ni, 120 Sn, Pu in the low energy region, and good results have been given by the coupled-channel calculations. Thus, it is demonstrated that the coupled-channel calculations with one set of the optical parameters well reproduce the total and scattering cross sections over a wide energy and mass region. (auth.)

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

Neutron-proton analyzing power at 12 MeV and charged πNN coupling constant

International Nuclear Information System (INIS)

Braun, R.T.; Tornow, W.; Gonzalez Trotter, D.E.; Howell, C.R.; Machleidt, R.; Roper, C.D.; Salinas, F.; Setze, H.R.; Walter, R.L.

1995-01-01

Recent reanalysis of scattering data by the Nijmegen group has led to new values for the πNN coupling constants, g 2 πdegree /4π and g 2 π± /4π, about 6% smaller than the previously accepted values. The impact of this finding is far reaching. Since the neutron-proton A y (θ) is dominated at low energies by the one-pion-exchange mechanism, accurate np data should provide unique information as to the magnitude of g 2 π± /4π. Using a new experimental setup consisting of a shielded neutron source, a five-pair neutron detector array, a n- 4 He polarimeter, and an intense polarized source with fast spin-flipping capability, we have measured a 15 point angular distribution of the neutron-proton A y (θ) at and incident neutron energy of 12 MeV to a statistical accuracy of 5x10 -4 . We will discuss the data taking procedures, the analysis, and the corrections applied to the data. Preliminary results will be presented

RADHEAT-V3, a code system for generating coupled neutron and gamma-ray group constants and analyzing radiation transport

International Nuclear Information System (INIS)

Koyama, Kinji; Taji, Yukichi; Miyasaka, Shun-ichi; Minami, Kazuyoshi.

1977-07-01

The modular code system RADHEAT is for producing coupled multigroup neutron and gamma-ray cross section sets, analyzing the neutron and gamma-ray transport, and calculating the energy deposition and atomic displacements due to these radiations in a nuclear reactor or shield. The basic neutron cross sections and secondary gamma-ray production data are taken from ENDF/B and POPOP4 libraries respectively. The system (1) generates multigroup neutron cross sections, energy deposition coefficients and atomic displacement factors due to neutron reactions, (2) generates multigroup gamma-ray cross sections and energy transfer coefficients, (3) generates secondary gamma-ray production cross sections, (4) combines these cross sections into the coupled set, (5) outputs and updates the multigroup cross section libraries in convenient formats for other transport codes, (6) analyzes the neutron and gamma-ray transport and calculates the energy deposition and the number density of atomic displacements in a medium, (7) collapses the cross sections to a broad-group structure, by option, using the weighting functions obtained by one-dimensional transport calculation, and (8) plots, by option, multigroup cross sections, and neutron and gamma-ray distributions. Definitions of the input data required in various options of the code system are also given. (auth.)

A 2D/1D coupling neutron transport method based on the matrix MOC and NEM methods

International Nuclear Information System (INIS)

Zhang, H.; Zheng, Y.; Wu, H.; Cao, L.

2013-01-01

A new 2D/1D coupling method based on the matrix MOC method (MMOC) and nodal expansion method (NEM) is proposed for solving the three-dimensional heterogeneous neutron transport problem. The MMOC method, used for radial two-dimensional calculation, constructs a response matrix between source and flux with only one sweep and then solves the linear system by using the restarted GMRES algorithm instead of the traditional trajectory sweeping process during within-group iteration for angular flux update. Long characteristics are generated by using the customization of commercial software AutoCAD. A one-dimensional diffusion calculation is carried out in the axial direction by employing the NEM method. The 2D and ID solutions are coupled through the transverse leakage items. The 3D CMFD method is used to ensure the global neutron balance and adjust the different convergence properties of the radial and axial solvers. A computational code is developed based on these theories. Two benchmarks are calculated to verify the coupling method and the code. It is observed that the corresponding numerical results agree well with references, which indicates that the new method is capable of solving the 3D heterogeneous neutron transport problem directly. (authors)

A 2D/1D coupling neutron transport method based on the matrix MOC and NEM methods

Energy Technology Data Exchange (ETDEWEB)

Zhang, H.; Zheng, Y.; Wu, H.; Cao, L. [School of Nuclear Science and Technology, Xi' an Jiaotong University, No. 28, Xianning West Road, Xi' an, Shaanxi 710049 (China)

2013-07-01

A new 2D/1D coupling method based on the matrix MOC method (MMOC) and nodal expansion method (NEM) is proposed for solving the three-dimensional heterogeneous neutron transport problem. The MMOC method, used for radial two-dimensional calculation, constructs a response matrix between source and flux with only one sweep and then solves the linear system by using the restarted GMRES algorithm instead of the traditional trajectory sweeping process during within-group iteration for angular flux update. Long characteristics are generated by using the customization of commercial software AutoCAD. A one-dimensional diffusion calculation is carried out in the axial direction by employing the NEM method. The 2D and ID solutions are coupled through the transverse leakage items. The 3D CMFD method is used to ensure the global neutron balance and adjust the different convergence properties of the radial and axial solvers. A computational code is developed based on these theories. Two benchmarks are calculated to verify the coupling method and the code. It is observed that the corresponding numerical results agree well with references, which indicates that the new method is capable of solving the 3D heterogeneous neutron transport problem directly. (authors)

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)

BARC 75 - A 75 group neutron-photon coupled cross-section library with P5- anisotropic scattering matrices

International Nuclear Information System (INIS)

Garg, S.B.

1990-01-01

A 75 group neutron-photon coupled cross-section library has been developed for 42 reactor nuclides utilizing the basic cross-section files - ENDF/B-IV for neutrons and DLC-7F for photons. 50 neutron energy groups and gamma energy groups are included in this library which should be well suited to carry out safety, shielding and core physics studies of nuclear reactors based on fission or fusion processes. This library is also adequate for oil logging and mineral exploration investigations. (author). 11 refs., 3 tabs

Numerical research on the neutronic/thermal-hydraulic/mechanical coupling characteristics of the optimized helium cooled solid breeder blanket for CFETR

International Nuclear Information System (INIS)

Cui, Shijie; Zhang, Dalin; Cheng, Jie; Tian, Wenxi; Su, G.H.

2017-01-01

As one of the candidate tritium breeding blankets for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of the helium cooled solid breeder blanket has recently been proposed. The neutronic, thermal-hydraulic and mechanical characteristics of the blanket directly affect its tritium breeding and safety performance. Therefore, neutronic/thermal-hydraulic/mechanical coupling analyses are of vital importance for a reliable blanket design. In this work, first, three-dimensional neutronics analysis and optimization of the typical outboard equatorial blanket module (No. 12) were performed for the comprehensive optimal scheme. Then, thermal and fluid dynamic analyses of the scheme under both normal and critical conditions were performed and coupled with the previous neutronic calculation results. With thermal-hydraulic boundaries, thermo-mechanical analyses of the structure materials under normal, critical and blanket over-pressurization conditions were carried out. In addition, several parametric sensitivity studies were also conducted to investigate the influences of the main parameters on the blanket temperature distributions. In this paper, the coupled analyses verify the reasonability of the optimized conceptual design preliminarily and can provide an important reference for the further analysis and optimization design of the CFETR helium cooled solid breeder blanket.

Numerical research on the neutronic/thermal-hydraulic/mechanical coupling characteristics of the optimized helium cooled solid breeder blanket for CFETR

Energy Technology Data Exchange (ETDEWEB)

Cui, Shijie; Zhang, Dalin, E-mail: dlzhang@mail.xjtu.edu.cn; Cheng, Jie; Tian, Wenxi; Su, G.H.

2017-01-15

As one of the candidate tritium breeding blankets for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of the helium cooled solid breeder blanket has recently been proposed. The neutronic, thermal-hydraulic and mechanical characteristics of the blanket directly affect its tritium breeding and safety performance. Therefore, neutronic/thermal-hydraulic/mechanical coupling analyses are of vital importance for a reliable blanket design. In this work, first, three-dimensional neutronics analysis and optimization of the typical outboard equatorial blanket module (No. 12) were performed for the comprehensive optimal scheme. Then, thermal and fluid dynamic analyses of the scheme under both normal and critical conditions were performed and coupled with the previous neutronic calculation results. With thermal-hydraulic boundaries, thermo-mechanical analyses of the structure materials under normal, critical and blanket over-pressurization conditions were carried out. In addition, several parametric sensitivity studies were also conducted to investigate the influences of the main parameters on the blanket temperature distributions. In this paper, the coupled analyses verify the reasonability of the optimized conceptual design preliminarily and can provide an important reference for the further analysis and optimization design of the CFETR helium cooled solid breeder blanket.

Reaction channel coupling effects for nucleons on 16O: Induced undularity and proton-neutron potential differences

Science.gov (United States)

Keeley, N.; Mackintosh, R. S.

2018-01-01

Background: Precise fitting of scattering observables suggests that the nucleon-nucleus interaction is l dependent. Such l dependence has been shown to be S -matrix equivalent to an undulatory l -independent potential. The undulations include radial regions where the imaginary term is emissive. Purpose: To study the dynamical polarization potential (DPP) generated in proton-16O and neutron-16O interaction potentials by coupling to pickup channels. Undulatory features occurring in these DPPs can be compared with corresponding features of empirical optical model potentials (OMPs). Furthermore, the additional inclusion of coupling to vibrational states of the target will provide evidence for dynamically generated nonlocality. Methods: The fresco code provides the elastic channel S -matrix Sl j for chosen channel couplings. Inversion, Sl j→V (r ) +l .s VSO(r ) , followed by subtraction of the bare potential, yields an l -independent and local representation of the DPP due to the chosen couplings. Results: The DPPs have strongly undulatory features, including radial regions of emissivity. Certain features of empirical DPPs appear, e.g., the full inverted potential has emissive regions. The DPPs for different collective states are additive except near the nuclear center, whereas the collective and reaction channel DPPs are distinctly nonadditive over a considerable radial range, indicating dynamical nonlocality. Substantial differences between the DPPs due to pickup coupling for protons and neutrons occur; these imply a greater difference between proton and neutron OMPs than the standard phenomenological prescription. Conclusions: The onus is on those who object to undularity in the local and l -independent representation of nucleon elastic scattering to show why such undulations do not occur. This work suggests that it is not legitimate to halt model-independent fits to high-quality data at the appearance of undularity.

Coupled Model of channels in parallel and neutron kinetics in two dimensions

International Nuclear Information System (INIS)

Cecenas F, M.; Campos G, R.M.; Valle G, E. del

2004-01-01

In this work an arrangement of thermohydraulic channels is presented that represent those four quadrants of a nucleus of reactor type BWR. The channels are coupled to a model of neutronic in two dimensions that allow to generate the radial profile of power of the reactor. Nevertheless that the neutronic pattern is of two dimensions, it is supplemented with axial additional information when considering the axial profiles of power for each thermo hydraulic channel. The stationary state is obtained the one it imposes as frontier condition the same pressure drop for all the channels. This condition is satisfied to iterating on the flow of coolant in each channel to equal the pressure drop in all the channels. This stationary state is perturbed later on when modifying the values for the effective sections corresponding to an it assembles. The calculation in parallel of the neutronic and the thermo hydraulic is carried out with Vpm (Virtual parallel machine) by means of an outline teacher-slave in a local net of computers. (Author)

Uncertainty and sensitivity analysis in the neutronic parameters generation for BWR and PWR coupled thermal-hydraulic–neutronic simulations

International Nuclear Information System (INIS)

Ánchel, F.; Barrachina, T.; Miró, R.; Verdú, G.; Juanas, J.; Macián-Juan, R.

2012-01-01

Highlights: ► Best-estimate codes are affected by the uncertainty in the methods and the models. ► Influence of the uncertainty in the macroscopic cross-sections in a BWR and PWR RIA accidents analysis. ► The fast diffusion coefficient, the scattering cross section and both fission cross sections are the most influential factors. ► The absorption cross sections very little influence. ► Using a normal pdf the results are more “conservative” comparing the power peak reached with uncertainty quantified with a uniform pdf. - Abstract: The Best Estimate analysis consists of a coupled thermal-hydraulic and neutronic description of the nuclear system's behavior; uncertainties from both aspects should be included and jointly propagated. This paper presents a study of the influence of the uncertainty in the macroscopic neutronic information that describes a three-dimensional core model on the most relevant results of the simulation of a Reactivity Induced Accident (RIA). The analyses of a BWR-RIA and a PWR-RIA have been carried out with a three-dimensional thermal-hydraulic and neutronic model for the coupled system TRACE-PARCS and RELAP-PARCS. The cross section information has been generated by the SIMTAB methodology based on the joint use of CASMO-SIMULATE. The statistically based methodology performs a Monte-Carlo kind of sampling of the uncertainty in the macroscopic cross sections. The size of the sampling is determined by the characteristics of the tolerance intervals by applying the Noether–Wilks formulas. A number of simulations equal to the sample size have been carried out in which the cross sections used by PARCS are directly modified with uncertainty, and non-parametric statistical methods are applied to the resulting sample of the values of the output variables to determine their intervals of tolerance.

Coupled neutronic core and subchannel analysis of nanofluids in VVER-1000 type reactor

Energy Technology Data Exchange (ETDEWEB)

Zarifi, Ehsan; Sepanloo, Kamran [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of). Reactor and Nuclear Safety School; Jahanfarnia, Golamreza [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Nuclear Engineering, Science and Research Branch

2017-05-15

This study is aimed to perform the coupled thermal-hydraulic/neutronic analysis of nanofluids as the coolant in the hot fuel assembly of VVER-1000 reactor core. Water-based nanofluid containing various volume fractions of Al{sub 2}O{sub 3} nanoparticle is analyzed. WIMS and CITATION codes are used for neutronic simulation of the reactor core, calculating neutron flux and thermal power distribution. In the thermal-hydraulic modeling, the porous media approach is used to analyze the thermal behavior of the reactor core and the subchannel analysis is used to calculate the hottest fuel assembly thermal-hydraulic parameters. The derived conservation equations for coolant and conduction heat transfer equation for fuel and clad are discretized by Finite volume method and solved numerically using visual FORTRAN program. Finally the analysis results for nanofluids and pure water are compared together. The achieved results show that at low concentration (0.1 percent volume fraction) alumina is the optimum nanoparticles for normal reactor operation.

Optimization of spring exchange coupled ferrites, studied by in situ neutron diffraction

DEFF Research Database (Denmark)

Ahlburg, Jakob; Christensen, Mogens; Granados-Miralles, Cecilia

Strong permanent magnets with a high energy-product are vital for a great number of electronic devices, these can be found in transformers, loudspeakers, windmills etc. Normally the preferred type of magnets are Rare Earth Metals (REM) containing magnets. REM excels since the magnetic contributio...... with varying temperature (fixed flow) or varying flow (fixed temperature) has been performed. To optimize the exchange-coupling several experiments with fixed temperature and flow, have been performed where the conversion from spinel to metal has been varied....... reduced. These metal oxides are antiferromagnetically ordered an is therefore considered a parasitic phase. However by fine-tuning the reaction temperature and hydrogen flow rate the occurrence of the phase can be minimized. In order to distinguish between Co and Fe Neutrons are chosen. Since neutrons...

Neutron transport

International Nuclear Information System (INIS)

Berthoud, Georges; Ducros, Gerard; Feron, Damien; Guerin, Yannick; Latge, Christian; Limoge, Yves; Santarini, Gerard; Seiler, Jean-Marie; Vernaz, Etienne; Coste-Delclaux, Mireille; M'Backe Diop, Cheikh; Nicolas, Anne; Andrieux, Catherine; Archier, Pascal; Baudron, Anne-Marie; Bernard, David; Biaise, Patrick; Blanc-Tranchant, Patrick; Bonin, Bernard; Bouland, Olivier; Bourganel, Stephane; Calvin, Christophe; Chiron, Maurice; Damian, Frederic; Dumonteil, Eric; Fausser, Clement; Fougeras, Philippe; Gabriel, Franck; Gagnier, Emmanuel; Gallo, Daniele; Hudelot, Jean-Pascal; Hugot, Francois-Xavier; Dat Huynh, Tan; Jouanne, Cedric; Lautard, Jean-Jacques; Laye, Frederic; Lee, Yi-Kang; Lenain, Richard; Leray, Sylvie; Litaize, Olivier; Magnaud, Christine; Malvagi, Fausto; Mijuin, Dominique; Mounier, Claude; Naury, Sylvie; Nicolas, Anne; Noguere, Gilles; Palau, Jean-Marc; Le Pallec, Jean-Charles; Peneliau, Yannick; Petit, Odile; Poinot-Salanon, Christine; Raepsaet, Xavier; Reuss, Paul; Richebois, Edwige; Roque, Benedicte; Royer, Eric; Saint-Jean, Cyrille de; Santamarina, Alain; Serot, Olivier; Soldevila, Michel; Tommasi, Jean; Trama, Jean-Christophe; Tsilanizara, Aime; Behar, Christophe; Provitina, Olivier; Lecomte, Michael; Forestier, Alain; Bender, Alexandra; Parisot, Jean-Francois; Finot, Pierre

2013-10-01

This bibliographical note presents a reference book which addresses the study of neutron transport in matter, the study of conditions for a chain reaction and the study of modifications of matter composition due to nuclear reactions. This book presents the main nuclear data, their measurement, assessment and processing, and the spallation. It proposes an overview of methods applied for the study of neutron transport: basic equations and their derived forms, deterministic methods and Monte Carlo method of resolution of the Boltzmann equation, methods of resolution of generalized Bateman equations, methods of time resolution of space kinetics coupled equations. It presents the main calculation codes, discusses the qualification and experimental aspects, and gives an overview of neutron transport applications: neutron transport calculation of reactors, neutron transport coupled with other disciplines, physics of fuel cycle, criticality

Description of a neutron field perturbed by a probe using coupled Monte Carlo and discrete ordinates radiation transport calculations

International Nuclear Information System (INIS)

Zazula, J.M.

1984-01-01

This work concerns calculation of a neutron response, caused by a neutron field perturbed by materials surrounding the source or the detector. Solution of a problem is obtained using coupling of the Monte Carlo radiation transport computation for the perturbed region and the discrete ordinates transport computation for the unperturbed system. (author). 62 refs

Three Mile Island Unit 1 Main Steam Line Break Three-Dimensional Neutronics/Thermal-Hydraulics Analysis: Application of Different Coupled Codes

International Nuclear Information System (INIS)

D'Auria, Francesco; Moreno, Jose Luis Gago; Galassi, Giorgio Maria; Grgic, Davor; Spadoni, Antonino

2003-01-01

A comprehensive analysis of the double ended main steam line break (MSLB) accident assumed to occur in the Babcock and Wilcox Three Mile Island Unit 1 (TMI-1) has been carried out at the Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione of the University of Pisa, Italy, in cooperation with the University of Zagreb, Croatia. The overall activity has been completed within the framework of the participation in the Organization for Economic Cooperation and Development-Committee on the Safety of Nuclear Installations-Nuclear Science Committee pressurized water reactor MSLB benchmark.Thermal-hydraulic system codes (various versions of Relap5), three-dimensional (3-D) neutronics codes (Parcs, Quabbox, and Nestle), and one subchannel code (Cobra) have been adopted for the analysis. Results from the following codes (or code versions) are assumed as reference:1. Relap5/mod3.2.2, beta version, coupled with the 3-D neutron kinetics Parcs code parallel virtual machine (PVM) coupling2. Relap5/mod3.2.2, gamma version, coupled with the 3-D neutron kinetics Quabbox code (direct coupling)3. Relap5/3D code coupled with the 3-D neutron kinetics Nestle code.The influence of PVM and of direct coupling is also discussed.Boundary and initial conditions of the system, including those relevant to the fuel status, have been supplied by Pennsylvania State University in cooperation with GPU Nuclear Corporation (the utility, owner of TMI) and the U.S. Nuclear Regulatory Commission. The comparison among the results obtained by adopting the same thermal-hydraulic nodalization and the coupled code version is discussed in this paper.The capability of the control rods to recover the accident has been demonstrated in all the cases as well as the capability of all the codes to predict the time evolution of the assigned transient. However, one stuck control rod caused some 'recriticality' or 'return to power' whose magnitude is largely affected by boundary and initial conditions

Validation of coupled neutronic / thermal-hydraulic codes for VVER reactors. Final report

Energy Technology Data Exchange (ETDEWEB)

Mittag, S.; Grundmann, U.; Kliem, S.; Kozmenkov, Y.; Rindelhardt, U.; Rohde, U.; Weiss, F.-P.; Langenbuch, S.; Krzykacz-Hausmann, B.; Schmidt, K.-D.; Vanttola, T.; Haemaelaeinen, A.; Kaloinen, E.; Kereszturi, A.; Hegyi, G.; Panka, I.; Hadek, J.; Strmensky, C.; Darilek, P.; Petkov, P.; Stefanova, S.; Kuchin, A.; Khalimonchuk, V.; Hlbocky, P.; Sico, D.; Danilin, S.; Ionov, V.; Nikonov, S.; Powney, D.

2004-08-01

In recent years, the simulation methods for the safety analysis of nuclear power plants have been continuously improved to perform realistic calculations. Therefore in VALCO work package 2 (WP 2), the usual application of coupled neutron-kinetic / thermal-hydraulic codes to VVER has been supplemented by systematic uncertainty and sensitivity analyses. A comprehensive uncertainty analysis has been carried out. The GRS uncertainty and sensitivity method based on the statistical code package SUSA was applied to the two transients studied earlier in SRR-1/95: A load drop of one turbo-generator in Loviisa-1 (VVER-440), and a switch-off of one feed water pump in Balakovo-4 (VVER-1000). The main steps of these analyses and the results obtained by applying different coupled code systems (SMABRE - HEXTRAN, ATHLET - DYN3D, ATHLET - KIKO3D, ATHLET - BIPR-8) are described in this report. The application of this method is only based on variations of input parameter values. No internal code adjustments are needed. An essential result of the analysis using the GRS SUSA methodology is the identification of the input parameters, such as the secondary-circuit pressure, the control-assembly position (as a function of time), and the control-assembly efficiency, that most sensitively affect safety-relevant output parameters, like reactor power, coolant heat-up, and primary pressure. Uncertainty bands for these output parameters have been derived. The variation of potentially uncertain input parameter values as a consequence of uncertain knowledge can activate system actions causing quite different transient evolutions. This gives indications about possible plant conditions that might be reached from the initiating event assuming only small disturbances. In this way, the uncertainty and sensitivity analysis reveals the spectrum of possible transient evolutions. Deviations of SRR-1/95 coupled code calculations from measurements also led to the objective to separate neutron kinetics from

Coupling of neutron transport equations. First results; Couplage d`equations en transport neutronique. premiere approche 1D monocinetique

Energy Technology Data Exchange (ETDEWEB)

Bal, G.

1995-07-01

To achieve whole core calculations of the neutron transport equation, we have to follow this 2 step method: space and energy homogenization of the assemblies; resolution of the homogenized equation on the whole core. However, this is no more valid when accidents occur (for instance depressurization causing locally strong heterogeneous media). One solution consists then in coupling two kinds of resolutions: a fine computation on the damaged cell (fine mesh, high number of energy groups) coupled with a coarse one everywhere else. We only deal here with steady state solutions (which already live in 6D spaces). We present here two such methods: The coupling by transmission of homogenized sections and the coupling by transmission of boundary conditions. To understand what this coupling is, we first restrict ourselves to 1D with respect to space in one energy group. The first two chapters deal with a recall of basic properties of the neutron transport equation. We give at chapter 3 some indications of the behaviour of the flux with respect to the cross sections. We present at chapter 4 some couplings and give some properties. Chapter 5 is devoted to a presentation of some numerical applications. (author). 9 refs., 7 figs.

Neutronics of pulsed spallation neutron sources

CERN Document Server

Watanabe, N

2003-01-01

Various topics and issues on the neutronics of pulsed spallation neutron sources, mainly for neutron scattering experiments, are reviewed to give a wide circle of readers a better understanding of these sources in order to achieve a high neutronic performance. Starting from what neutrons are needed, what the spallation reaction is and how to produce slow-neutrons more efficiently, the outline of the target and moderator neutronics are explained. Various efforts with some new concepts or ideas have already been devoted to obtaining the highest possible slow-neutron intensity with desired pulse characteristics. This paper also reviews the recent progress of such efforts, mainly focused on moderator neutronics, since moderators are the final devices of a neutron source, which determine the source performance. Various governing parameters for neutron-pulse characteristics such as material issues, geometrical parameters (shape and dimensions), the target-moderator coupling scheme, the ortho-para-hydrogen ratio, po...

Development of a coupled neutronic/thermal-hydraulic tool with multi-scale capabilities and applications to HPLWR core analysis

International Nuclear Information System (INIS)

Monti, Lanfranco; Starflinger, Joerg; Schulenberg, Thomas

2011-01-01

Highlights: → Advanced analysis and design techniques for innovative reactors are addressed. → Detailed investigation of a 3 pass core design with a multi-physics-scales tool. → Coupled 40-group neutron transport/equivalent channels TH core analyses methods. → Multi-scale capabilities: from equivalent channels to sub-channel pin-by-pin study. → High fidelity approach: reduction of conservatism involved in core simulations. - Abstract: The High Performance Light Water Reactor (HPLWR) is a thermal spectrum nuclear reactor cooled and moderated with light water operated at supercritical pressure. It is an innovative reactor concept, which requires developing and applying advanced analysis tools as described in the paper. The relevant water density reduction associated with the heat-up, together with the multi-pass core design, results in a pronounced coupling between neutronic and thermal-hydraulic analyses, which takes into account the strong natural influence of the in-core distribution of power generation and water properties. The neutron flux gradients within the multi-pass core, together with the pronounced dependence of water properties on the temperature, require to consider a fine spatial resolution in which the individual fuel pins are resolved to provide precise evaluation of the clad temperature, currently considered as one of the crucial design criteria. These goals have been achieved considering an advanced analysis method based on the usage of existing codes which have been coupled with developed interfaces. Initially neutronic and thermal-hydraulic full core calculations have been iterated until a consistent solution is found to determine the steady state full power condition of the HPLWR core. Results of few group neutronic analyses might be less reliable in case of HPLWR 3-pass core than for conventional LWRs because of considerable changes of the neutron spectrum within the core, hence 40 groups transport theory has been preferred to the

TISKTH-3: a couple neutronics/thermal-hydraulics code for the transient analysis of light water reactors

International Nuclear Information System (INIS)

Peng Muzhang; Zhang Quan; Wang Guoli; Zhang Yuman

1988-01-01

TISKTH-3 is a coupled neutronics/thermal-hydraulics code for the transient analysis. A 3-dimensional neutron kinetics equation solved by the Nodal Green's Function Method is used for the neutronics model of the code. A homogeneous equilibrium model with a complete boiling curve and two numerical solutions of the implicit and explicit scheme is used for the thermal-hydraulics model of the code. A 2-dimensional heat conduction equation with variable conductivity solved by the method of weighted residuals is used for the fuel rod heat transfer model of the code. TISKTH-3 is able to analyze the fast transient process and complicate accident situations in the core. The initative applications have shown that the stability and convergency in the calculations with the code are satisfactory

TISKTH-3: a couple neutronics/thermal-hydraulics code for the transient analysis of light water reactors

Energy Technology Data Exchange (ETDEWEB)

Muzhang, Peng; Quan, Zhang; Guoli, Wang; Yuman, Zhang

1988-03-01

TISKTH-3 is a coupled neutronics/thermal-hydraulics code for the transient analysis. A 3-dimensional neutron kinetics equation solved by the Nodal Green's Function Method is used for the neutronics model of the code. A homogeneous equilibrium model with a complete boiling curve and two numerical solutions of the implicit and explicit scheme is used for the thermal-hydraulics model of the code. A 2-dimensional heat conduction equation with variable conductivity solved by the method of weighted residuals is used for the fuel rod heat transfer model of the code. TISKTH-3 is able to analyze the fast transient process and complicate accident situations in the core. The initative applications have shown that the stability and convergency in the calculations with the code are satisfactory.

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

Neutronics of pulsed spallation neutron sources

International Nuclear Information System (INIS)

Watanabe, Noboru

2003-01-01

Various topics and issues on the neutronics of pulsed spallation neutron sources, mainly for neutron scattering experiments, are reviewed to give a wide circle of readers a better understanding of these sources in order to achieve a high neutronic performance. Starting from what neutrons are needed, what the spallation reaction is and how to produce slow-neutrons more efficiently, the outline of the target and moderator neutronics are explained. Various efforts with some new concepts or ideas have already been devoted to obtaining the highest possible slow-neutron intensity with desired pulse characteristics. This paper also reviews the recent progress of such efforts, mainly focused on moderator neutronics, since moderators are the final devices of a neutron source, which determine the source performance. Various governing parameters for neutron-pulse characteristics such as material issues, geometrical parameters (shape and dimensions), the target-moderator coupling scheme, the ortho-para-hydrogen ratio, poisoning, etc are discussed, aiming at a high performance pulsed spallation source

Coupled neutronics/thermal-hydraulics and safety characteristics of liquid-fueled molten salt reactors

International Nuclear Information System (INIS)

Qiu, Suizheng; Zhang, Dalin; Liu, Minghao; Liu, Limin; Xu, Rongshuan; Gong, Cheng; Su, Guanghui

2016-01-01

Molten salt reactor (MSR) as one candidate of the Generation IV advanced nuclear power systems is attracted more attention in China due to its top ranked fuel cycle and thorium utilization. The MSRs are characterized by using liquid-fuel, which offers complicated coupling problem of neutronics and thermal hydraulics. In this paper, the fundamental model and numerical method are established to calculate and analyze the safety characteristics for liquid-fuel MSRs. The theories and methodologies are applied to the MOSART concept. The liquid-fuel flow effects on neutronics, reactivity coefficients and three operation parameters' influences at steady state are obtained, which provide the basic information for safety analysis. The unprotected loss of flow transient is calculated, the results of which shows the inherent safety characteristics of MOSART due to its strong negative reactivity feedbacks.

Coupled neutronics/thermal-hydraulics and safety characteristics of liquid-fueled molten salt reactors

Energy Technology Data Exchange (ETDEWEB)

Qiu, Suizheng; Zhang, Dalin; Liu, Minghao; Liu, Limin; Xu, Rongshuan; Gong, Cheng; Su, Guanghui [Xi' an Jiaotong Univ. (China). State Key Laboratory of Multiphase Flow in Power Engineering

2016-05-15

Molten salt reactor (MSR) as one candidate of the Generation IV advanced nuclear power systems is attracted more attention in China due to its top ranked fuel cycle and thorium utilization. The MSRs are characterized by using liquid-fuel, which offers complicated coupling problem of neutronics and thermal hydraulics. In this paper, the fundamental model and numerical method are established to calculate and analyze the safety characteristics for liquid-fuel MSRs. The theories and methodologies are applied to the MOSART concept. The liquid-fuel flow effects on neutronics, reactivity coefficients and three operation parameters' influences at steady state are obtained, which provide the basic information for safety analysis. The unprotected loss of flow transient is calculated, the results of which shows the inherent safety characteristics of MOSART due to its strong negative reactivity feedbacks.

The coupled code system DORT-TD/THERMIX and its application to the OECD/NEA/NSC PBMR400 MW coupled neutronics thermal hydraulics transient benchmark

International Nuclear Information System (INIS)

Pautz, A.; Tyobeka, B.; Ivanov, K.

2009-01-01

In new reactor designs that are still under review such as the Pebble Bed Modular Reactor (PBMR), not much experimental data exists to benchmark newly developed computer codes against. Such a situation requires that nuclear engineers and designers of this novel reactor design must resort to the validation of a newly developed code through a code-to-code benchmarking exercise because there are validated codes that are currently in use to analyze this reactor design, albeit very few of them. There are numerous HTR core physics benchmarks that are currently being pursued by different organizations, for different purposes. One such benchmark exercise is the PBMR-400MW OECD/NEA coupled neutronics/thermal hydraulics transient benchmark. In this paper, a newly developed coupled neutronics thermal hydraulics code system, DORT-TD/THERMIX with both transport and diffusion theory options, is used to simulate both the steady-state as well as several transient scenarios in this benchmark problem. (orig.)

The coupling of the Star-Cd software to a whole-core neutron transport code Decart for PWR applications

International Nuclear Information System (INIS)

Thomas, J.W.; Lee, H.C.; Downar, T.J.; Sofu, T.; Weber, D.P.; Joo, H.G.; Cho, J.Y.

2003-01-01

As part of a U.S.- Korea collaborative U.S. Department of Energy INERI project, a comprehensive high-fidelity reactor-core modeling capability is being developed for detailed analysis of existing and advanced PWR reactor designs. An essential element of the project has been the development of an interface between the computational fluid dynamics (CFD) module, STAR-CD, and the neutronics module, DeCART. Since the computational mesh for CFD and neutronics calculations are generally different, the capability to average and decompose data on these different meshes has been an important part of code coupling activities. An averaging process has been developed to extract neutronics zone temperatures in the fuel and coolant and to generate appropriate multi group cross sections and densities. Similar procedures have also been established to map the power distribution from the neutronics zones to the mesh structure used in the CFD module. Since MPI is used as the parallel model in STAR-CD and conflicts arise during initiation of a second level of MPI, the interface developed here is based on using TCP/IP protocol sockets to establish communication between the CFD and neutronics modules. Preliminary coupled calculations have been performed for PWR fuel assembly size problems and converged solutions have been achieved for a series of steady-state problems ranging from a single pin to a 1/8 model of a 17 x 17 PWR fuel assembly. (authors)

TRACE/VALKIN: a neutronics-thermohydraulics coupled code to analyze strong 3D transients

Energy Technology Data Exchange (ETDEWEB)

Rafael Miro; Gumersindo Verdu; Ana Maria Sanchez [Chemical and Nuclear Engineering Department. Polytechnic University of Valencia. Cami de Vera s/n. 46022 Valencia (Spain); Damian Ginestar [Applied Mathematics Department. Polytechnic University of Valencia. Cami de Vera s/n. 46022 Valencia (Spain)

2005-07-01

Full text of publication follows: A nuclear reactor simulator consists mainly of two different blocks, which solve the models used for the basic physical phenomena taking place in the reactor. In this way, there is a neutronic module which simulates the neutron balance in the reactor core, and a thermal-hydraulics module, which simulates the heat transfer in the fuel, the heat transfer from the fuel to the water, and the different condensation and evaporation processes taking place in the reactor core and in the condenser systems. TRACE is a two-phase, two-fluid thermal-hydraulic reactor systems analysis code. The TRACE acronym stands for TRAC/RELAP Advanced Computational Engine, reflecting its ability to run both RELAP5 and TRAC legacy input models. It includes a three-dimensional kinetics module called PARCS for performing advanced analysis of coupled core thermal-hydraulic/kinetics problems. TRACE-VALKIN code is a new time domain analysis code to study transients in LWR reactors. This code uses the best estimate code TRACE to give account of the heat transfer and thermal-hydraulic processes, and a 3D neutronics module. This module has two options, the MODKIN option that makes use of a modal method based on the assumption that the neutronic flux can be approximately expanded in terms of the dominant lambda modes associated with a static configuration of the reactor core, and the NOKIN option that uses a one-step backward discretization of the neutron diffusion equation. The lambda modes are obtained using the Implicit Restarted Arnoldi approach or the Jacob-Davidson algorithm. To check the performance of the coupled code TRACE-VALKIN against complex 3D neutronic transients, using the cross-sections tables generated with the translator SIMTAB from SIMULATE to TRACE/VALKIN, the Cofrentes NPP SCRAM-61 transient is simulated. Cofrentes NPP is a General Electric BWR-6 design located in Valencia-land (Spain). It is in operation since 1985 and currently in its fifteenth

Program package for calculation of cross sections of neutron scattering on deformed nuclei by the coupled-channel method

International Nuclear Information System (INIS)

Kloss, Yu.Yu.

1985-01-01

Program package and numerical solution of the problem for a system of coupled equations used in optical model to solve a problem on low and mean energy neutron scattering on deformed nuclei, is considered. With these programs differnet scattering cross sections depending on the incident neutron energy on even-even and even-odd nuclei were obtained. The programm permits to obtain different scattering cross sections (elastic, inelastic), excitation cross sections of the first three energy levels of rotational band depending on the energy, angular distributions and neutron polarizations including excited channels. In the program there is possibility for accounting even-even nuclei octupole deformation

A review on the relativistic effective field theory with parameterized couplings for nuclear matter and neutron stars

Energy Technology Data Exchange (ETDEWEB)

Vasconcellos, C. A. Zen, E-mail: cesarzen@cesarzen.com [Instituto de Física, Universidade Federal do Rio Grande do Sul (UFRGS), Av. Bento Gonçalves 9500, 91501-970, Porto Alegre (Brazil); International Center for Relativistic Astrophysics Network (ICRANet), Piazza della Repubblica 10, 65122 Pescara (Italy)

2015-12-17

Nuclear science has developed many excellent theoretical models for many-body systems in the domain of the baryon-meson strong interaction for the nucleus and nuclear matter at low, medium and high densities. However, a full microscopic understanding of nuclear systems in the extreme density domain of compact stars is still lacking. The aim of this contribution is to shed some light on open questions facing the nuclear many-body problem at the very high density domain. Here we focus our attention on the conceptual issue of naturalness and its role in shaping the baryon-meson phase space dynamics in the description of the equation of state (EoS) of nuclear matter and neutrons stars. In particular, in order to stimulate possible new directions of research, we discuss relevant aspects of a recently developed relativistic effective theory for nuclear matter within Quantum Hadrodynamics (QHD) with genuine many-body forces and derivative natural parametric couplings. Among other topics we discuss in this work the connection of this theory with other known effective QHD models of the literature and its potentiality in describing a new physics for dense matter. The model with parameterized couplings exhausts the whole fundamental baryon octet (n, p, Σ{sup −}, Σ{sup 0}, Σ{sup +}, Λ, Ξ{sup −}, Ξ{sup 0}) and simulates n-order corrections to the minimal Yukawa baryon couplings by considering nonlinear self-couplings of meson fields and meson-meson interaction terms coupled to the baryon fields involving scalar-isoscalar (σ, σ∗), vector-isoscalar (ω, Φ), vector-isovector (ϱ) and scalar-isovector (δ) virtual sectors. Following recent experimental results, we consider in our calculations the extreme case where the Σ{sup −} experiences such a strong repulsion that its influence in the nuclear structure of a neutron star is excluded at all. A few examples of calculations of properties of neutron stars are shown and prospects for the future are discussed.

Analysis of the OECD/NRC BWR Turbine Trip Transient Benchmark with the Coupled Thermal-Hydraulics and Neutronics Code TRAC-M/PARCS

International Nuclear Information System (INIS)

Lee, Deokjung; Downar, Thomas J.; Ulses, Anthony; Akdeniz, Bedirhan; Ivanov, Kostadin N.

2004-01-01

An analysis of the Peach Bottom Unit 2 Turbine Trip 2 (TT2) experiment has been performed using the U.S. Nuclear Regulatory Commission coupled thermal-hydraulics and neutronics code TRAC-M/PARCS. The objective of the analysis was to assess the performance of TRAC-M/PARCS on a BWR transient with significance in two-phase flow and spatial variations of the neutron flux. TRAC-M/PARCS results are found to be in good agreement with measured plant data for both steady-state and transient phases of the benchmark. Additional analyses of four fictitious extreme scenarios are performed to provide a basis for code-to-code comparisons and comprehensive testing of the thermal-hydraulics/neutronics coupling. The obtained results of sensitivity studies on the effect of direct moderator heating on transient simulation indicate the importance of this modeling aspect

Concatenating algorithms for parallel numerical simulations coupling radiation hydrodynamics with neutron transport

International Nuclear Information System (INIS)

Mo Zeyao

2004-11-01

Multiphysics parallel numerical simulations are usually essential to simplify researches on complex physical phenomena in which several physics are tightly coupled. It is very important on how to concatenate those coupled physics for fully scalable parallel simulation. Meanwhile, three objectives should be balanced, the first is efficient data transfer among simulations, the second and the third are efficient parallel executions and simultaneously developments of those simulation codes. Two concatenating algorithms for multiphysics parallel numerical simulations coupling radiation hydrodynamics with neutron transport on unstructured grid are presented. The first algorithm, Fully Loosely Concatenation (FLC), focuses on the independence of code development and the independence running with optimal performance of code. The second algorithm. Two Level Tightly Concatenation (TLTC), focuses on the optimal tradeoffs among above three objectives. Theoretical analyses for communicational complexity and parallel numerical experiments on hundreds of processors on two parallel machines have showed that these two algorithms are efficient and can be generalized to other multiphysics parallel numerical simulations. In especial, algorithm TLTC is linearly scalable and has achieved the optimal parallel performance. (authors)

Coupled neutronics/thermal-hydraulics analysis of a high-performance light-water reactor fuel assembly

International Nuclear Information System (INIS)

Waata, C.L.

2006-07-01

The use of water at supercritical pressure as coolant and moderator introduces a challenge in the design of a High-Performance Light-Water Reactor (HPLWR) fuel assembly. At supercritical pressure condition (P=25 MPa), the thermal-hydraulics behaviour of water differs strongly from that at sub-critical pressure due to a rapid variation of the thermal-physical properties across the pseudo-critical line. Due of the strong link between the water (moderation) and the neutron spectrum and subsequently the power distribution, a coupling of neutronics and thermal-hydraulics has become a necessity for reactor concepts operating at supercritical pressure condition. The effect of neutron moderation on the local parameters of thermal-hydraulics and vice-verse in a fuel assembly has to be considered for an accurate design analysis. In this study, the Monte Carlo N-Particle code (MCNP) and the sub-channel code STAFAS (Sub-channel Thermal-hydraulics Analysis of a Fuel Assembly under Supercritical conditions) have been coupled for the design analysis of a fuel assembly with supercritical water as coolant and moderator. Both codes are well known for complex geometry modelling. The MCNP code is used for neutronics analyses and for the prediction of power profiles of individual fuel rods. The sub-channel code STAFAS for the thermal-hydraulics analyses takes into account the coolant properties beyond the critical point as well as separate moderator channels. The coupling procedure is realized automatically. MCNP calculates the power distribution in each fuel rod, which is then transferred into STAFAS to obtain the corresponding thermal-hydraulic conditions in each sub-channel. The new thermal-hydraulic conditions are used to generate a new input deck for the next MCNP calculation. This procedure is repeated until a converged state is achieved. The coupled code system was tested on a proposed fuel assembly design of a HPLWR. An under-relaxation was introduced to achieve convergence

The OECD/NEA/NSC PBMR400 MW coupled neutronics thermal hydraulics transient benchmark - Steady-state results and status

International Nuclear Information System (INIS)

Reitsma, F.; Han, J.; Ivanov, K.; Sartori, E.

2008-01-01

The PBMR is a High-Temperature Gas-cooled Reactor (HTGR) concept developed to be built in South Africa. The analysis tools used for core neutronic design and core safety analysis need to be verified and validated. Since only a few pebble-bed HTR experimental facilities or plant data are available the use of code-to-code comparisons are an essential part of the V and V plans. As part of this plan the PBMR 400 MW design and a representative set of transient cases is defined as an OECD benchmark. The scope of the benchmark is to establish a series of well-defined multi-dimensional computational benchmark problems with a common given set of cross-sections, to compare methods and tools in coupled neutronics and thermal hydraulics analysis with a specific focus on transient events. The OECD benchmark includes steady-state and transients cases. Although the focus of the benchmark is on the modelling of the transient behaviour of the PBMR core, it was also necessary to define some steady-state cases to ensure consistency between the different approaches before results of transient cases could be compared. This paper describes the status of the benchmark project and shows the results for the three steady state exercises defined as a standalone neutronics calculation, a standalone thermal-hydraulic core calculation, and a coupled neutronics/thermal-hydraulic simulation. (authors)

SAID analysis of meson photoproduction: Determination of neutron and proton EM couplings

Directory of Open Access Journals (Sweden)

Strakovsky Igor

2014-06-01

Full Text Available We present an overview of the GW SAID group effort to analyze on new pion photoproduction on both proton- and neutron-targets. The main database contribution came from the recent CLAS and MAMI unpolarized and polarized measurements. The differential cross section for the processes γn → π−p was extracted from new measurements accounting for Fermi motion effects in the impulse approximation (IA as well as NN- and πN effects beyond the IA. The electromagnetic coupling results are compared to other recent studies.

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)

CFD - neutronic coupled calculation of a quarter of a simplified PWR fuel assembly including spacer pressure drop and turbulence enhancement

Energy Technology Data Exchange (ETDEWEB)

Pena, C.; Pellacani, F.; Macian Juan, R., E-mail: carlos.pena@ntech.mw.tum.de, E-mail: pellacani@ntech.mw.tum.de, E-mail: macian@ntech.mw.tum.de [Technische Universitaet Muenchen, Garching (Germany). Ntech Lehrstuhl fuer Nukleartechnik; Chiva, S., E-mail: schiva@emc.uji.es [Universitat Jaume I, Castellon de la Plana (Spain). Dept. de Ingenieria Mecanica y Construccion; Barrachina, T.; Miro, R., E-mail: rmiro@iqn.upv.es, E-mail: tbarrachina@iqn.upv.es [Universitat Politecnica de Valencia (ISIRYM/UPV) (Spain). Institute for Industrial, Radiophysical and Environmental Safety

2011-07-01

A computational code system based on coupling the 3D neutron diffusion code PARCS v2.7 and the Ansys CFX 13.0 Computational Fluid Dynamics (CFD) code has been developed as a tool for nuclear reactor systems simulations. This paper presents the coupling methodology between the CFD and the neutronic code. The methodology to simulate a 3D-neutronic problem coupled with 1D thermal hydraulics is already a mature technology, being part of the regular calculations performed to analyze different kinds of Reactivity Insertion Accidents (RIA) and asymmetric transients in Nuclear Power Plants, with state-of-the-art coupled codes like TRAC-B/NEM, RELAP5/PARCS, TRACE/PARCS, RELAP3D, RETRAN3D, etc. This work represents one of the first attempts to couple the multiphysics of a nuclear reactor core with a 3D spatial resolution in a computer code. This will open new possibilities regarding the analysis of fuel elements, contributing to a better understanding and design of the heat transfer process and specific fluid dynamics phenomena such as cross flow among fuel elements. The transient simulation of control rod insertion, boron dilution and cold water injection will be made possible with a degree of accuracy not achievable with current methodologies based on the use of system and/or subchannel codes. The transport of neutrons depends on several parameters, like fuel temperature, moderator temperature and density, boron concentration and fuel rod insertion. These data are calculated by the CFD code with high local resolution and used as input to the neutronic code to calculate a 3D nodal power distribution that will be returned and remapped to the CFD code control volumes (cells). Since two different nodalizations are used to discretized the same system, an averaging and interpolating procedure is needed to realize an effective data exchange. These procedures have been developed by means of the Ansys CFX 'User Fortran' interface; a library with several subroutines has

CFD - neutronic coupled calculation of a quarter of a simplified PWR fuel assembly including spacer pressure drop and turbulence enhancement

International Nuclear Information System (INIS)

Pena, C.; Pellacani, F.; Macian Juan, R.; Chiva, S.; Barrachina, T.; Miro, R.

2011-01-01

A computational code system based on coupling the 3D neutron diffusion code PARCS v2.7 and the Ansys CFX 13.0 Computational Fluid Dynamics (CFD) code has been developed as a tool for nuclear reactor systems simulations. This paper presents the coupling methodology between the CFD and the neutronic code. The methodology to simulate a 3D-neutronic problem coupled with 1D thermal hydraulics is already a mature technology, being part of the regular calculations performed to analyze different kinds of Reactivity Insertion Accidents (RIA) and asymmetric transients in Nuclear Power Plants, with state-of-the-art coupled codes like TRAC-B/NEM, RELAP5/PARCS, TRACE/PARCS, RELAP3D, RETRAN3D, etc. This work represents one of the first attempts to couple the multiphysics of a nuclear reactor core with a 3D spatial resolution in a computer code. This will open new possibilities regarding the analysis of fuel elements, contributing to a better understanding and design of the heat transfer process and specific fluid dynamics phenomena such as cross flow among fuel elements. The transient simulation of control rod insertion, boron dilution and cold water injection will be made possible with a degree of accuracy not achievable with current methodologies based on the use of system and/or subchannel codes. The transport of neutrons depends on several parameters, like fuel temperature, moderator temperature and density, boron concentration and fuel rod insertion. These data are calculated by the CFD code with high local resolution and used as input to the neutronic code to calculate a 3D nodal power distribution that will be returned and remapped to the CFD code control volumes (cells). Since two different nodalizations are used to discretized the same system, an averaging and interpolating procedure is needed to realize an effective data exchange. These procedures have been developed by means of the Ansys CFX 'User Fortran' interface; a library with several subroutines has been

Inelastic Neutron Scattering and Magnetisation Investigation of an Exchange-Coupled Dy2 SMM

Science.gov (United States)

Baker, Michael L.; Zhang, Qing; Sarachik, Myriam P.; Kent, Andrew D.; Chen, Yizhang; Butch, Nicholas; Pineda, Eufemio M.; McInnes, Eric

The strong spin orbit coupling and weak crystal field energies of simple exchange-coupled rare earth SMMs makes the precise evaluation of their magnetic properties nontrivial. Here we report a detailed investigation of the single molecule magnet hqH2Dy2(hq)4(NO3)3MeOH. Inelastic neutron scattering is used to obtain direct access to several low energy crystal field excitations. The INS results display several features that are not found in earlier FIR absorption experiments, while other features found in the latter are absent. Based on the effective point charge model, numerical calculations are currently underway to resolve these apparent discrepancies using complementary magnetisation measurements to resolve the exchange between Dy ions. Work supported by ARO W911NF-13-1-1025 (CCNY) and NSF-DMR-1309202 (NYU).

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

Development of a coupled code system based on system transient code, RETRAN, and 3-D neutronics code, MASTER

International Nuclear Information System (INIS)

Kim, K. D.; Jung, J. J.; Lee, S. W.; Cho, B. O.; Ji, S. K.; Kim, Y. H.; Seong, C. K.

2002-01-01

A coupled code system of RETRAN/MASTER has been developed for best-estimate simulations of interactions between reactor core neutron kinetics and plant thermal-hydraulics by incorporation of a 3-D reactor core kinetics analysis code, MASTER into system transient code, RETRAN. The soundness of the consolidated code system is confirmed by simulating the MSLB benchmark problem developed to verify the performance of a coupled kinetics and system transient codes by OECD/NEA

Some results on the neutron transport and the coupling of equations; Quelques resultats sur le transport neutronique et le couplage d`equations

Energy Technology Data Exchange (ETDEWEB)

Bal, G. [Electricite de France (EDF), Direction des Etudes et Recherches, 92 - Clamart (France)

1997-12-31

Neutron transport in nuclear reactors is well modeled by the linear Boltzmann transport equation. Its resolution is relatively easy but very expensive. To achieve whole core calculations, one has to consider simpler models, such as diffusion or homogeneous transport equations. However, the solutions may become inaccurate in particular situations (as accidents for instance). That is the reason why we wish to solve the equations on small area accurately and more coarsely on the remaining part of the core. It is than necessary to introduce some links between different discretizations or modelizations. In this note, we give some results on the coupling of different discretizations of all degrees of freedom of the integral-differential neutron transport equation (two degrees for the angular variable, on for the energy component, and two or three degrees for spatial position respectively in 2D (cylindrical symmetry) and 3D). Two chapters are devoted to the coupling of discrete ordinates methods (for angular discretization). The first one is theoretical and shows the well posing of the coupled problem, whereas the second one deals with numerical applications of practical interest (the results have been obtained from the neutron transport code developed at the R and D, which has been modified for introducing the coupling). Next, we present the nodal scheme RTN0, used for the spatial discretization. We show well posing results for the non-coupled and the coupled problems. At the end, we deal with the coupling of energy discretizations for the multigroup equations obtained by homogenization. Some theoretical results of the discretization of the velocity variable (well-posing of problems), which do not deal directly with the purposes of coupling, are presented in the annexes. (author). 34 refs.

Assessment of the impact of neutronic/thermal-hydraulic coupling on the design and performance of nuclear reactors for space propulsion

International Nuclear Information System (INIS)

Aithal, S.M.; Aldemir, T.; Vafai, K.

1994-01-01

A series of studies has been performed to investigate the potential impact of the coupling between neutronics and thermal hydraulics on the design and performance assessment of solid core reactors for nuclear thermal space propulsion, using the particle bed reactor (PBR) concept as an example system. For a given temperature distribution in the reactor, the k eff and steady-state core power distribution are obtained from three-dimensional, continuous energy Monte Carlo simulations using the MCNP code. For a given core power distribution, determination of the temperature distribution in the core and hydrogen-filled annulus between the reflector and pressure vessel is based on a nonthermal equilibrium analysis. The results show that a realistic estimation of fuel, core size, and control requirements for PBRs using hydrogenous moderators, as well as optimization of the overall engine design, may require coupled neutronic/thermal-hydraulic studies. However, it may be possible to estimate the thermal safety margins and propellant exit temperatures based on power distributions obtained from neutronic calculations at room temperature. The results also show that, while variation of the hydrogen flow rate in the annulus has been proposed as a partial control mechanism for PBRs, such control mechanism may not be feasible for PBRs with high moderator-to-fuel ratios and hence soft core neutron spectra

Fast neutron scattering from soft nuclei: coupled-channel formalism and illustrations

International Nuclear Information System (INIS)

Delaroche, J.P.

1986-01-01

Spectra of most of the even-even nuclei have a character which is neither that of a pure vibrator nor that of a pure rotor. Instead, the nuclear spectra display very often both characters. Therefore, improvements in the analysises of nucleon scattering and reaction cross sections require that appropriate collective models of nuclear structure be used. A selection of these models is reviewed, and suggestions are given as to how to extend the familiar coupled-channel formalism to incorporate these enriched collective pictures. These extensions are primarily intended to describe inelastic scattering from levels belonging to β - , γ - and octupole bands. Illustrations are given for neutron and proton scattering off various nuclei [fr

Reactor physics experiments in PURNIMA sub critical facility coupled with 14 MeV neutron source

International Nuclear Information System (INIS)

Kumar, Rajeev; Degweker, S.B.; Patel, Tarun; Bishnoi, Saroj; Adhikari, P.S.

2011-01-01

Accelerator Driven Sub-critical Systems (ADSS) are attracting increasing worldwide attention due to their superior safety characteristics and their potential for burning actinide and fission product waste and energy production. A number of countries around the world have drawn up roadmaps/programs for development of ADSS. Indian interest in ADSS has an additional dimension, which is related to the planned utilization of our large thorium reserves for future nuclear energy generation. A programme for development of ADSS is taken up at the Bhabha Atomic Research Centre (BARC) in India. This includes R and D activities for high current proton accelerator development, target development and Reactor Physics studies. As part of the ADSS Reactor Physics research programme, a sub-critical facility is coming up in BARC which will be coupled with an existing D-D/D-T neutron generator. Two types of cores are planned. In one of these, the sub-critical reactor assembly consists of natural uranium moderated by high density polyethylene (HDP) and reflected by BeO. The other consists of natural uranium moderated by light water. The maximum neutron yield of the neutron source with tritium target is around 10 10 neutron per sec. Various reactor physics experiments like measurement of the source strength, neutron flux distribution, buckling estimation and sub-critical source multiplication are planned. Apart from this, measurement of the total fission power and neutron spectrum will also be carried out. Mainly activation detectors will be used in all in-core neutron flux measurement. Measurement of the degree of sub-criticality by various deterministic and noise methods is planned. Helium detectors with advanced data acquisition card will be used for the neutron noise experiments. Noise characteristics of ADSS are expected to be different from that of traditional reactors due to the non-Poisson statistical features of the source. A new theory incorporating these features has been

BWR transient analysis using neutronic / thermal hydraulic coupled codes including uncertainty quantification

International Nuclear Information System (INIS)

Hartmann, C.; Sanchez, V.; Tietsch, W.; Stieglitz, R.

2012-01-01

The KIT is involved in the development and qualification of best estimate methodologies for BWR transient analysis in cooperation with industrial partners. The goal is to establish the most advanced thermal hydraulic system codes coupled with 3D reactor dynamic codes to be able to perform a more realistic evaluation of the BWR behavior under accidental conditions. For this purpose a computational chain based on the lattice code (SCALE6/GenPMAXS), the coupled neutronic/thermal hydraulic code (TRACE/PARCS) as well as a Monte Carlo based uncertainty and sensitivity package (SUSA) has been established and applied to different kind of transients of a Boiling Water Reactor (BWR). This paper will describe the multidimensional models of the plant elaborated for TRACE and PARCS to perform the investigations mentioned before. For the uncertainty quantification of the coupled code TRACE/PARCS and specifically to take into account the influence of the kinetics parameters in such studies, the PARCS code has been extended to facilitate the change of model parameters in such a way that the SUSA package can be used in connection with TRACE/PARCS for the U and S studies. This approach will be presented in detail. The results obtained for a rod drop transient with TRACE/PARCS using the SUSA-methodology showed clearly the importance of some kinetic parameters on the transient progression demonstrating that the coupling of a best-estimate coupled codes with uncertainty and sensitivity tools is very promising and of great importance for the safety assessment of nuclear reactors. (authors)

Fast-neutron and gamma-ray imaging with a capillary liquid xenon converter coupled to a gaseous photomultiplier

Science.gov (United States)

Israelashvili, I.; Coimbra, A. E. C.; Vartsky, D.; Arazi, L.; Shchemelinin, S.; Caspi, E. N.; Breskin, A.

2017-09-01

Gamma-ray and fast-neutron imaging was performed with a novel liquid xenon (LXe) scintillation detector read out by a Gaseous Photomultiplier (GPM). The 100 mm diameter detector prototype comprised a capillary-filled LXe converter/scintillator, coupled to a triple-THGEM imaging-GPM, with its first electrode coated by a CsI UV-photocathode, operated in Ne/5%CH4 at cryogenic temperatures. Radiation localization in 2D was derived from scintillation-induced photoelectron avalanches, measured on the GPM's segmented anode. The localization properties of 60Co gamma-rays and a mixed fast-neutron/gamma-ray field from an AmBe neutron source were derived from irradiation of a Pb edge absorber. Spatial resolutions of 12± 2 mm and 10± 2 mm (FWHM) were reached with 60Co and AmBe sources, respectively. The experimental results are in good agreement with GEANT4 simulations. The calculated ultimate expected resolutions for our application-relevant 4.4 and 15.1 MeV gamma-rays and 1-15 MeV neutrons are 2-4 mm and ~ 2 mm (FWHM), respectively. These results indicate the potential applicability of the new detector concept to Fast-Neutron Resonance Radiography (FNRR) and Dual-Discrete-Energy Gamma Radiography (DDEGR) of large objects.

Applications of a lead pile coupled with fast reactor core of Yayoi as an intermediate energy neutron standard field

International Nuclear Information System (INIS)

Kosako, Toshiso; Nakazawa, Masaharu; Sekiguchi, Akira; Wakabayashi, Hiroaki.

1976-10-01

Intermediate neutron column of YAYOI reactor is here evaluated as an intermediate energy neutron standard field which provides a base of the measurements of various reaction rates in that energy region, including detector calibration and Doppler coefficient determination. The experiments were performed using YAYOI's core as a fast neutron source by coupling with the large lead pile, which is a 160 ton's octagon of 2.5 m high and with a thickness of about 2.5 m face to face distance. Spatial variation of the neutron flux in the lead pile was estimated by gold activation foils, and the neutron spectrum by sandwich foils, a helium-3 proportional counter and a proton recoil counter. The calculated results were obtained using one and two- dimensional discrete ordinate code, ANISN and TWOTRAN II. Through comparison of experiment with calculation, it became clear that the neutron field at the central block has simple energy spectrum and stable spatial distribution of the neutron flux, the absolute of which was 5.0 x 10 4 (n/cm 2 /sec/Watt) at the representative energy of 1 KeV. The energy spectrum of the position and the spatial dependent neutron flux in the lead pile are both represented by the semiempirical formula, which must be useful both for evaluation of experimental data and for future applications. (auth.)

Coupled thermal-hydraulic and neutronic simulations of Phenix control rod withdrawal tests with SIMMER-IV

International Nuclear Information System (INIS)

Kriventsev, Vladimir; Gabrielli, Fabrizio; Rineiski, Andrei

2014-01-01

The “end-of-life” tests performed in the Phenix reactor before its final shutdown in 2009, in particular the Control Rod (CR) withdrawal experiments provide an excellent opportunity for the validation and verification of the reactor physics computer codes and modeling approaches. SIMMER-IV, a modern three-dimensional reactor safety code, has been recently employed at Karlsruhe Institute of Technology (KIT) for simulating Phenix experiments in the framework of a benchmark exercise organized under the IAEA project. In this paper, we report and discuss main results obtained with SIMMER-IV at KIT. Particular attention is devoted to the coupling features of thermal-hydraulics and neutronics and their mutual influences. The reactor reactivity, power and neutron flux distributions calculated with SIMMER-IV are in good agreement both with experimental results and with calculations with advanced neutronics codes, such as ERANOS, while the CR reactivity worth is overestimated due to neglecting heterogeneity effects. Because of its multi-physics capabilities SIMMER also calculates the temperature distributions which are in a good agreement with the experimental test results. In this work we describe the improvements in SIMMER neutronics model by employing a correction that is based on the results of cell calculations performed with ERANOS. The study confirms that the 3D SIMMER-IV code can accurately predict major fast reactor neutronics and thermal hydraulic parameters, provided that a special treatment is employed for CR modeling. The results of calculations are analyzed in frames of SIMMER-IV validation and verification assessment. (author)

MSLB coupled 3D neutronics-thermalhydraulic analysis of a large PWR using RELAP5-3D

International Nuclear Information System (INIS)

Lo Nigro, A.; Spadoni, A.; D'Auria, F.; Saiu, G.

2001-01-01

A RELAP5-3D model of the Westinghouse AP1000 NSSS has been set up and it has been used to analyze the MSLB accident. Main results (both spatial distributions and time trends) have been represented with 3D plots and graphical movies. The method applied allows accounting for the coupled 3D neutronics and thermalyhdraulics effects, suggesting to consider its applicability in Safety Analysis.(author)

Uppsala neutron-proton scattering measurements and the πNN coupling constant

International Nuclear Information System (INIS)

Olsson, N.; Blomgren, J.; Conde, H.; Dangtip, S.; Elmgren, K.; Rahm, J.; Roennqvist, T.; Zorro, R.; Loiseau, B.

2000-01-01

The differential np scattering cross section has been measured at 96 MeV and 162 MeV at backward angles at the neutron beam facility of the The Svedberg Laboratory in Uppsala. The angular distributions have been normalized to the experimental total np cross section. Between 150 and 180 , the angular distributions are steeper than for most previous measurements and nucleon-nucleon potential predictions, but for all the angular range covered, the data agree very well in shape with the recent PSI data. At 180 , the difference versus older data amounts to about 10%, implying serious consequences because of the fundamental importance of this cross section. Values of the charged πNN coupling constant have been extracted from the data. (orig.)

Study on development of virtual reactor core laboratory (1). Development of prototype coupled neutronic, thermal-hydraulic and structural analysis system

International Nuclear Information System (INIS)

Uto, Nariaki; Sugaya, Toshio; Tsukimori, Kazuyuki; Negishi, Hitoshi; Enuma, Yasuhiro; Sakai, Takaaki

1999-09-01

A study on development of virtual reactor core laboratory, which is to conduct numerical experiments representative of complicated physical phenomena in practical reactor core systems on a computational environment, has progressed at Japan Nuclear Cycle Development Institute (JNC). The study aims at systematic evaluation of these phenomena into which nuclear reactions, thermal-hydraulic characteristics, structural responses and fuel behaviors combine, and effective utilization of the obtained comprehension for core design. This report presents a production of a prototype computational system which is required to construct the virtual reactor core laboratory. This system is to evaluate reactor core performance under the coupled neutronic, thermal-hydraulic and structural phenomena, and is composed of two analysis tools connected by a newly developed interface program; 1) an existing space-dependent coupled neutronic and thermal-hydraulic analysis system arranged at JNC and 2) a core deformation analysis code. It acts on a cluster of several DEC/Alpha workstations. A specific library called MPI1 (Message Passing Interface 1) is incorporated as a tool for communicating among the analysis modules consisting of the system. A series of calculations for simulating a sequence of Unprotected Loss Of Heat Sink (ULOHS) coupled with rapid drop of some neutron absorber devices in a prototype fast reactor is tried to investigate how the system works. The obtained results show the core deformation behavior followed by the reactivity change that can be properly evaluated. The results of this report show that the system is expected to be useful for analyzing sensitivity of reactor core performance with respect to uncertainties of various design parameters and establishing a concept of passive safety reactor system, taking into account space distortion of neutron flux distribution during abnormal events as well as reactivity feedback from core deformation. (author)

A methodology for the coupling of RAMONA-3B neutron kinetics and TRAC-BF1 thermal-hydraulics

International Nuclear Information System (INIS)

Lopez, Arsenio Procopio; Morales Sandoval, Jaime B.

2005-01-01

The initial objective of this project was to directly couple the RAMONA and TRAC codes running on different PCs. The idea was to use the best part of each one and eliminate some of their limitations and widen the applicability of these codes to simulate different BWR and system components. It was required to try to minimize the amount of changes to present code subroutines and calculation procedures. If possible, just substitute values obtained in the parallel code. Preliminary results indicated that using a CHAN component of TRAC to model thermal-hydraulic phenomena for each neutronic channel modeled in RAMONA is rather difficult. Large amounts of CPU time consumption are obtained and lots of PCs would make this solution difficult, besides considerable large transients are introduced by the differences in thermal-hydraulic results of these codes. The substitution of the thermal-hydraulics of RAMONA, by the TRAC channel calculations, is possible but simulation of a null transient on both codes must be planed and a gradual change must be controlled by an additional supervisory subroutine. An indirect coupling of these codes, it is therefore proposed, in order to eliminate most of these limitations. In this indirect coupling, a thermal-hydraulic model of the average tube in a bundle and the global channel cooling fluid dynamics is programmed for each neutronic channel while the global reactor vessel and core is modeled by TRAC with just four channels and four rings. Results are more reliable, coupling is simpler and faster simulations are possible

The coupling of the neutron transport application RATTLESNAKE to the nuclear fuels performance application BISON under the MOOSE framework

Energy Technology Data Exchange (ETDEWEB)

Gleicher, Frederick N.; Williamson, Richard L.; Ortensi, Javier; Wang, Yaqi; Spencer, Benjamin W.; Novascone, Stephen R.; Hales, Jason D.; Martineau, Richard C.

2014-10-01

The MOOSE neutron transport application RATTLESNAKE was coupled to the fuels performance application BISON to provide a higher fidelity tool for fuel performance simulation. This project is motivated by the desire to couple a high fidelity core analysis program (based on the self-adjoint angular flux equations) to a high fidelity fuel performance program, both of which can simulate on unstructured meshes. RATTLESNAKE solves self-adjoint angular flux transport equation and provides a sub-pin level resolution of the multigroup neutron flux with resonance treatment during burnup or a fast transient. BISON solves the coupled thermomechanical equations for the fuel on a sub-millimeter scale. Both applications are able to solve their respective systems on aligned and unaligned unstructured finite element meshes. The power density and local burnup was transferred from RATTLESNAKE to BISON with the MOOSE Multiapp transfer system. Multiple depletion cases were run with one-way data transfer from RATTLESNAKE to BISON. The eigenvalues are shown to agree well with values obtained from the lattice physics code DRAGON. The one-way data transfer of power density is shown to agree with the power density obtained from an internal Lassman-style model in BISON.

A New Coupled CFD/Neutron Kinetics System for High Fidelity Simulations of LWR Core Phenomena: Proof of Concept

Directory of Open Access Journals (Sweden)

Jorge Pérez Mañes

2014-01-01

Full Text Available The Institute for Neutron Physics and Reactor Technology (INR at the Karlsruhe Institute of Technology (KIT is investigating the application of the meso- and microscale analysis for the prediction of local safety parameters for light water reactors (LWR. By applying codes like CFD (computational fluid dynamics and SP3 (simplified transport reactor dynamics it is possible to describe the underlying phenomena in a more accurate manner than by the nodal/coarse 1D thermal hydraulic coupled codes. By coupling the transport (SP3 based neutron kinetics (NK code DYN3D with NEPTUNE-CFD, within a parallel MPI-environment, the NHESDYN platform is created. The newly developed system will allow high fidelity simulations of LWR fuel assemblies and cores. In NHESDYN, a heat conduction solver, SYRTHES, is coupled to NEPTUNE-CFD. The driver module of NHESDYN controls the sequence of execution of the solvers as well as the communication between the solvers based on MPI. In this paper, the main features of NHESDYN are discussed and the proof of the concept is done by solving a single pin problem. The prediction capability of NHESDYN is demonstrated by a code-to-code comparison with the DYNSUB code. Finally, the future developments and validation efforts are highlighted.

Neutron producing reactions in PuBe neutron sources

Energy Technology Data Exchange (ETDEWEB)

Bagi, János [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU) (Germany); Lakosi, László; Nguyen, Cong Tam [Centre for Energy Research, Hungarian Academy of Sciences, Budapest (Hungary)

2016-01-01

There are a plenty of out-of-use plutonium–beryllium neutron sources in Eastern Europe presenting both nuclear safeguards and security issues. Typically, their actual Pu content is not known. In the last couple of years different non-destructive methods were developed for their characterization. For such methods detailed knowledge of the nuclear reactions taking place within the source is necessary. In this paper we investigate the role of the neutron producing reactions, their contribution to the neutron yield and their dependence on the properties of the source.

Approach to the calculation of energy deposition in a container of fuel irradiated by the neutronic codes coupling fluid-dynamics

International Nuclear Information System (INIS)

Hueso, C.; Aleman, A.; Colomer, C.; Fabbri, M.; Martin, M.; Saellas, J.

2013-01-01

In this work identifies a possible area of improvement through the creation of a code of coupling between deposition energy codes which calculate neutron (MCNP), and data from heading into fluid dynamics (ANSYS-Fluent) or codes thermomechanical, called MAFACS (Monte Carlo ANSYS Fluent Automatic Coupling Software), being possible to so summarize the process by shortening the needs of computing time, increasing the precision of the results and therefore improving the design of the components.

PHISICS multi-group transport neutronic capabilities for RELAP5

Energy Technology Data Exchange (ETDEWEB)

Epiney, A.; Rabiti, C.; Alfonsi, A.; Wang, Y.; Cogliati, J.; Strydom, G. [Idaho National Laboratory (INL), 2525 N. Fremont Ave., Idaho Falls, ID 83402 (United States)

2012-07-01

PHISICS is a neutronic code system currently under development at INL. Its goal is to provide state of the art simulation capability to reactor designers. This paper reports on the effort of coupling this package to the thermal hydraulic system code RELAP5. This will enable full prismatic core and system modeling and the possibility to model coupled (thermal-hydraulics and neutronics) problems with more options for 3D neutron kinetics, compared to the existing diffusion theory neutron kinetics module in RELAP5 (NESTLE). The paper describes the capabilities of the coupling and illustrates them with a set of sample problems. (authors)

Neutron CSI: Integrated platform for non-destructive composition and stress imaging with neutrons

International Nuclear Information System (INIS)

Materna, T.; Pirling, T.

2011-01-01

We propose to build an interdisciplinary platform for non-destructive analysis and imaging with neutrons. The project regroups an instrument already available at ILL (Laue-Langevin Institute), SALSA, with a new one for Neutron Tomography coupled to Prompt-Gamma Neutron Activation (PGNA) as well as partial usage of another proposed instrument, FIPPS. The focus of the proposition is the versatility of high spatial resolution and energy-selective neutron tomography to provide a rapid and precise 3D morphological map of an object as well as indirect information on its 3D elemental and structural composition through the scan of Bragg-edges in transmission. Coupled to PGNA imaging and the strain analysis power of SALSA, the aim of the platform is to answer key questions occurring in geological, metallurgical, engineering and medical fields, material research and cultural heritage. (authors)

Test case specifications for coupled neutronics-thermal hydraulics calculation of Gas-cooled Fast Reactor

Science.gov (United States)

Osuský, F.; Bahdanovich, R.; Farkas, G.; Haščík, J.; Tikhomirov, G. V.

2017-01-01

The paper is focused on development of the coupled neutronics-thermal hydraulics model for the Gas-cooled Fast Reactor. It is necessary to carefully investigate coupled calculations of new concepts to avoid recriticality scenarios, as it is not possible to ensure sub-critical state for a fast reactor core under core disruptive accident conditions. Above mentioned calculations are also very suitable for development of new passive or inherent safety systems that can mitigate the occurrence of the recriticality scenarios. In the paper, the most promising fuel material compositions together with a geometry model are described for the Gas-cooled fast reactor. Seven fuel pin and fuel assembly geometry is proposed as a test case for coupled calculation with three different enrichments of fissile material in the form of Pu-UC. The reflective boundary condition is used in radial directions of the test case and vacuum boundary condition is used in axial directions. During these condition, the nuclear system is in super-critical state and to achieve a stable state (which is numerical representation of operational conditions) it is necessary to decrease the reactivity of the system. The iteration scheme is proposed, where SCALE code system is used for collapsing of a macroscopic cross-section into few group representation as input for coupled code NESTLE.

Coupled neutronics and thermal hydraulics of high density cores for FRM II

Energy Technology Data Exchange (ETDEWEB)

Breitkreutz, Harald

2011-03-04

According to the 'Verwaltungsvereinbarung zwischen Bund und Land vom 30.5.2003' and its updating on 13.11.2010, the Forschungs-Neutronenquelle Heinz Maier-Leibnitz, Frm II, has to convert its fuel element to an uranium enrichment which is significantly lower than the current 93%, in case this is economically reasonable and doesn't impact the reactor performance immoderate. In the framework of this conversion, new calculations regarding neutronics and thermal hydraulics for the anticipated core configurations have to be made. The computational power available nowadays allows for detailed 3D calculations, on the neutronic as well as on the thermal hydraulic side. In this context, a new program system, 'X{sup 2}', was developed. It couples the Monte Carlo code McnpX, the computational fluid dynamics code Cfx and the burn-up code sequence MonteBurns. The codes were modified and extended to meet the requirements of the coupled calculation concept. To verify the new program system, highly detailed calculations for the current fuel element were made and compared to simulations and measurements that were performed in the past. The results strengthen the works performed so far and show that the original, conservative approach overestimates all critical thermal hydraulic values. Using the CFD software, effects like the impact of the combs that fix the fuel plates and the pressure drop at the edges of the fuel plates were studied in great detail for the first time. Afterwards, a number of possible new fuel elements with lower enrichment, based on disperse and monolithic UMo (uranium with 8 wt.-% Mo) were analysed. A number of straight-forward conversion scenarios was discussed, showing that a further compaction of the fuel element, an extended cycle length or an increased reactor power is needed to compensate the flux loss, which is caused by the lower enrichment. This flux loss is in excess of 7%. The discussed new fuel elements include a 50

Coupled neutronics and thermal hydraulics of high density cores for FRM II

International Nuclear Information System (INIS)

Breitkreutz, Harald

2011-01-01

According to the 'Verwaltungsvereinbarung zwischen Bund und Land vom 30.5.2003' and its updating on 13.11.2010, the Forschungs-Neutronenquelle Heinz Maier-Leibnitz, Frm II, has to convert its fuel element to an uranium enrichment which is significantly lower than the current 93%, in case this is economically reasonable and doesn't impact the reactor performance immoderate. In the framework of this conversion, new calculations regarding neutronics and thermal hydraulics for the anticipated core configurations have to be made. The computational power available nowadays allows for detailed 3D calculations, on the neutronic as well as on the thermal hydraulic side. In this context, a new program system, 'X 2 ', was developed. It couples the Monte Carlo code McnpX, the computational fluid dynamics code Cfx and the burn-up code sequence MonteBurns. The codes were modified and extended to meet the requirements of the coupled calculation concept. To verify the new program system, highly detailed calculations for the current fuel element were made and compared to simulations and measurements that were performed in the past. The results strengthen the works performed so far and show that the original, conservative approach overestimates all critical thermal hydraulic values. Using the CFD software, effects like the impact of the combs that fix the fuel plates and the pressure drop at the edges of the fuel plates were studied in great detail for the first time. Afterwards, a number of possible new fuel elements with lower enrichment, based on disperse and monolithic UMo (uranium with 8 wt.-% Mo) were analysed. A number of straight-forward conversion scenarios was discussed, showing that a further compaction of the fuel element, an extended cycle length or an increased reactor power is needed to compensate the flux loss, which is caused by the lower enrichment. This flux loss is in excess of 7%. The discussed new fuel elements include a 50% enriched disperse UMo core with

Heat Source Characterization In A TREAT Fuel Particle Using Coupled Neutronics Binary Collision Monte-Carlo Calculations

Energy Technology Data Exchange (ETDEWEB)

Schunert, Sebastian; Schwen, Daniel; Ghassemi, Pedram; Baker, Benjamin; Zabriskie, Adam; Ortensi, Javier; Wang, Yaqi; Gleicher, Frederick; DeHart, Mark; Martineau, Richard

2017-04-01

This work presents a multi-physics, multi-scale approach to modeling the Transient Test Reactor (TREAT) currently prepared for restart at the Idaho National Laboratory. TREAT fuel is made up of microscopic fuel grains (r ˜ 20µm) dispersed in a graphite matrix. The novelty of this work is in coupling a binary collision Monte-Carlo (BCMC) model to the Finite Element based code Moose for solving a microsopic heat-conduction problem whose driving source is provided by the BCMC model tracking fission fragment energy deposition. This microscopic model is driven by a transient, engineering scale neutronics model coupled to an adiabatic heating model. The macroscopic model provides local power densities and neutron energy spectra to the microscpic model. Currently, no feedback from the microscopic to the macroscopic model is considered. TREAT transient 15 is used to exemplify the capabilities of the multi-physics, multi-scale model, and it is found that the average fuel grain temperature differs from the average graphite temperature by 80 K despite the low-power transient. The large temperature difference has strong implications on the Doppler feedback a potential LEU TREAT core would see, and it underpins the need for multi-physics, multi-scale modeling of a TREAT LEU core.

Development of a discrete-ordinate approximation of the neutron transport equation for coupled xy-R-geometry

International Nuclear Information System (INIS)

Maertens, H.D.

1982-01-01

The inhomogenious structure of modern heavy water reactor fuel elements result in a strong spacial dependence of the neutron flux. The flux distribution can be calculated in detail by numerical methods, which describe exactly the geometrical heterogeniety and take into account the neutron flux anisotropy by higher transport theoretical approximations. Starting from the discrete ordinate method an approximation of the neutron transport equation has been developed, allowing for a cylindrical representation of the fuel-elements in a rectangular array of rods. The discretisation of the space variables, is based on the finite-difference approximation, defining a rectangular lattice in a two-dimensional cartesian coordinate system, which can be cut and replaced by circular mesh elements of a partially one-dimensional cylindrical coordinate system at arbitrary space points. To couple the two spacial regions the outer circle line of a cylindrical geometry is approximated in the cartesian system by a polygon with n >= 8. A cylindrical geometry is approximated in the cartesian system by a polygon with n>=8. A cylindrical geometry is thus enclosed by a system of two-dimensional rectangular, triangular and trapezoid mesh elements. The directional distribution of the neutron flux is conserved when switching from the xy-system to the cylindrical coordinate system. The angle discretisation by balanced sets of squares (EQsub(n)) allows a simple definition of transfer-coefficients for the redistribution of the neutron flux due to coordinate transformations. The procedure is verified and tested by selected problems. Possible applications and limits are discussed. (orig.) [de

Fusion neutronics

CERN Document Server

Wu, Yican

2017-01-01

This book provides a systematic and comprehensive introduction to fusion neutronics, covering all key topics from the fundamental theories and methodologies, as well as a wide range of fusion system designs and experiments. It is the first-ever book focusing on the subject of fusion neutronics research. Compared with other nuclear devices such as fission reactors and accelerators, fusion systems are normally characterized by their complex geometry and nuclear physics, which entail new challenges for neutronics such as complicated modeling, deep penetration, low simulation efficiency, multi-physics coupling, etc. The book focuses on the neutronics characteristics of fusion systems and introduces a series of theories and methodologies that were developed to address the challenges of fusion neutronics, and which have since been widely applied all over the world. Further, it introduces readers to neutronics design’s unique principles and procedures, experimental methodologies and technologies for fusion systems...

Constraints on hyperon couplings from neutron star equations of state

CERN Document Server

Miyazaki, K

2005-01-01

Based on the constituent quark picture of baryons and taking into account the contributions of isovector and strange mesons, we have developed the extended Zimanyi-Moszkowski model of dense baryon matter for studying neutron star (NS) equations of state (EOSs). Four sets of meson-hyperons coupling constants are investigated. The first is characterized by strong attractive N\\Sigma interaction while the others have repulsive N\\Sigma interactions. The second is characterized by strong attractive \\Lambda\\Lambda interaction. The third has weak \\Lambda\\Lambda but strong attractive \\Sigma\\Sigma interactions. The last one has much weaker \\Sigma\\Sigma interaction than the third one. By systematic analyses of the EOSs and mass sequences of NSs, it has been found that the strong attractive N\\Sigma, \\Lambda\\Lambda and \\Sigma\\Sigma interactions are ruled out. The result is consistent to the most recent information on hyperon interactions from the experimental and theoretical i! nvestigations of hypernuclei.

Naturalness of Nonlinear Scalar Self-Couplings in a Relativistic Mean Field Theory for Neutron Stars

International Nuclear Information System (INIS)

Maekawa, Claudio; Razeira, Moises; Vasconcellos, Cesar A. Z.; Dillig, Manfred; Bodmann, Bardo E. J.

2004-01-01

We investigate the role of naturalness in effective field theory. We focus on dense hadronic matter using a generalized relativistic multi-baryon lagrangian density mean field approach which contains nonlinear self-couplings of the σ, δ meson fields and the fundamental baryon octet. We adjust the model parameters to describe bulk static properties of ordinary nuclear matter. Then, we show that our approach represents a natural modelling of nuclear matter under the extreme conditions of density as the ones found in the interior of neutron stars

Thermal hydraulic and neutron kinetic coupled simulation of the IPR-R1 Triga reactor

Energy Technology Data Exchange (ETDEWEB)

Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia; Silva, Clarysson A.M. da; Veloso, Maria Auxiliadora F.; Soares, Humbero V., E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: clarysson@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br, E-mail: betovitor@ig.com.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 Rede), Rio de Janeiro, RJ (Brazil)

2013-07-01

The nuclear industry and the scientific community have turned the attention for the development of coupled 3D neutron kinetics (NK) and thermal-hydraulic (TH) system codes to investigate specific nuclear reactor transients. Improving in theoretical investigations of complex phenomena in nuclear reactor technology have been increased thanks to numerical methods and computational resources incorporated in nuclear codes. This paper presents a model for the IPR-R1 TRIGA research reactor using the RELAP5-3D 3.0 code. The development and the assessment of the thermal-hydraulic RELAP5 code model for the IPR-R1 have been validated for steady state and transient situations and the results were published in preceding works. Results of RELAP5-3D steady state and a transient case presented in this paper show good agreement with experimental data, validating then this model for point kinetic calculations. To supply adequate cross sections to the NK code, the WIMSD5 is being used. First results of steady state calculation using the 3D neutron modeling are being presented in this paper. (author)

Coherence features of the spin-aligned neutron-proton pair coupling scheme

International Nuclear Information System (INIS)

Qi, C; Blomqvist, J; Bäck, T; Cederwall, B; Johnson, A; Liotta, R J; Wyss, R

2012-01-01

The seniority scheme has been shown to be extremely useful for the classification of nuclear states in semi-magic nuclei. The neutron-proton (np) correlation breaks the seniority symmetry in a major way. As a result, the corresponding wave function is a mixture of many components with different seniority quantum numbers. In this paper, we show that the np interaction may favor a new kind of coupling in N=Z nuclei, i.e. the so-called isoscalar spin-aligned np pair mode. Shell model calculations reveal that the ground and low-lying yrast states of the N=Z nuclei 92 Pd and 96 Cd may be mainly built upon such spin-aligned np pairs, each carrying the maximum angular momentum J=9 allowed by the shell 0 g 9/2 which is dominant in this nuclear region.

Pellet by pellet neutron flux calculations coupled with nodal expansion method

International Nuclear Information System (INIS)

Aldo, Dall'Osso

2003-01-01

We present a technique whose aim is to replace 2-dimensional pin by pin de-homogenization, currently done in core reactor calculations with the nodal expansion method (NEM), by a 3-dimensional finite difference diffusion calculation. This fine calculation is performed as a zoom in each node taking as boundary conditions the results of the NEM calculations. The size of fine mesh is of the order of a fuel pellet. The coupling between fine and NEM calculations is realised by an albedo like boundary condition. Some examples are presented showing fine neutron flux shape near control rods or assembly grids. Other fine flux behaviour as the thermal flux rise in the fuel near the reflector is emphasised. In general the results show the interest of the method in conditions where the separability of radial and axial directions is not granted. (author)

Differential cross section for neutron scattering from 209Bi at 37 MeV and the weak particle-core coupling

International Nuclear Information System (INIS)

Zhou Zuying; Ruan Xichao; Du Yanfeng; Qi Bujia; Tang Hongqing; Xia Haihong; Walter, R. L.; Braun, R. T.; Howell, C. R.; Tornow, W.; Weisel, G. J.; Dupuis, M.; Delaroche, J. P.; Chen Zemin; Chen Zhenpeng; Chen Yingtang

2010-01-01

Differential scattering cross-section data have been measured at 43 angles from 11 deg. to 160 deg. for 37-MeV neutrons incident on 209 Bi. The primary motivation for the measurements is to address the scarcity of neutron scattering data above 30 MeV and to improve the accuracy of optical-model predictions at medium neutron energies. The high-statistics measurements were conducted at the China Institute of Atomic Energy using the 3 H(d,n) 4 He reaction as the neutron source, a pulsed deuteron beam, and time-of-flight (TOF) techniques. Within the resolution of the TOF spectrometer, the measurements included inelastic scattering components. The sum of elastic and inelastic scattering cross sections was computed in joint optical-model and distorted-wave Born approximation calculations under the assumption of the weak particle-core coupling. The results challenge predictions from well-established spherical optical potentials. Good agreement between data and calculations is achieved at 37 MeV provided that the balance between surface and volume absorption in a recent successful model [A. J. Koning and J. P. Delaroche, Nucl. Phys. A 713, 231 (2003)] is modified, thus suggesting the need for global optical-model improvements at medium neutron energies.

Coupled neutronic and thermal-hydraulic code benchmark activities at the International Nuclear Safety Center

International Nuclear Information System (INIS)

Podlazov, L. N.

1998-01-01

Two realistic benchmark problems are defined and used to assess the performance of coupled thermal-hydraulic and neutronic codes used in simulating dynamic processes in VVER-1000 and RBMK reactor systems. One of the problems simulates a design basis accident involving the ejection of three control and protection system rods from a VVER-1000 reactor. The other is based on a postulated rod withdrawal from an operating RBMK reactor. Preliminary results calculated by various codes are compared. While these results show significant differences, the intercomparisons performed so far provide a basis for further evaluation of code limitations and modeling assumptions

Elemental analysis using instrumental neutron activation analysis and inductively coupled plasma atomic emission spectrometry: a comparative study

International Nuclear Information System (INIS)

Chung, Yong Sam; Choi, Kwang Soon; Moon, Jong Hwa; Kim, Sun Ha; Lim, Jong Myoung; Kim, Young Jin; Quraishi, Shamshad Begum

2003-05-01

Elemental analyses for certified reference materials were carried out using instrumental neutron activation analysis and inductively coupled plasma-atomic emission spectrometry. Five Certified Reference Materials (CRM) were selected for the study on comparative analysis of environmental samples. The CRM are Soil (NIST SRM 2709), Coal fly ash (NIST SRM 1633a), urban dust (NIST SRM 1649a) and air particulate on filter media (NIST SRM 2783 and human hair (GBW 09101)

Neutronic and thermal-hydraulic coupling using Milonga and OpenFOAM codes: an approach using free software

International Nuclear Information System (INIS)

Silva, Vitor Vasconcelos Araújo

2016-01-01

The development of a fine mesh coupled neutronics/thermal-hydraulics framework mainly using open source software is presented. The contributions proposed go in two different directions: one, is the focus on the open software development, a concept widely spread in many fields of knowledge but rarely explored in the nuclear engineering field; the second, is the use of operating system shared memory as a fast and reliable storage area to couple the computational fluid dynamics (CFD) software OpenFOAM to the free and flexible reactor core analysis code Milonga. This concept was applied to simulate the behavior of the TRIGA Mark 1 IPR-R1 reactor fuel pin in steady-state mode. The macroscopic cross-sections for the model, a set of two-group cross-sections data, were generated using WIMSD-5B code. The results show that this innovative coupled system gives consistent results, encouraging system further development and its use for complex nuclear systems. (author)

Neutronic / thermal-hydraulic coupling with the code system Trace / Parcs

International Nuclear Information System (INIS)

Mejia S, D. M.; Del Valle G, E.

2015-09-01

The developed models for Parcs and Trace codes corresponding for the cycle 15 of the Unit 1 of the Laguna Verde nuclear power plant are described. The first focused to the neutronic simulation and the second to thermal hydraulics. The model developed for Parcs consists of a core of 444 fuel assemblies wrapped in a radial reflective layer and two layers, a superior and another inferior, of axial reflector. The core consists of 27 total axial planes. The model for Trace includes the vessel and its internal components as well as various safety systems. The coupling between the two codes is through two maps that allow its intercommunication. Both codes are used in coupled form performing a dynamic simulation that allows obtaining acceptably a stable state from which is carried out the closure of all the main steam isolation valves (MSIVs) followed by the performance of safety relief valves (SRVs) and ECCS. The results for the power and reactivities introduced by the moderator density, the fuel temperature and total temperature are shown. Data are also provided like: the behavior of the pressure in the steam dome, the water level in the downcomer, the flow through the MSIVs and SRVs. The results are explained for the power, the pressure in the steam dome and the water level in the downcomer which show agreement with the actions of the MSIVs, SRVs and ECCS. (Author)

Combined neutron imaging techniques for cultural heritage purpose

International Nuclear Information System (INIS)

Materna, T.

2009-01-01

This article presents the different new neutron techniques developed by the Ancient Charm collaboration to image objects of cultural heritage importance: Prompt-gamma-ray activation imaging (PGAI) coupled to cold/thermal neutron transmission tomography, Neutron Resonance Capture Imaging (NRCI) and Neutron Resonance Tomography.

Manufacturing of thermal neutron sensor using pMOS

International Nuclear Information System (INIS)

Lee, Nam Ho; Kim, Seung Ho

2005-05-01

A pMOSFET sensor having a Gadolinium converter has been invented successfully as a slow neutron sensor that is sensitive to neutron energy down to 0.025 eV. The Gd layer converts low energy neutrons to ionizing radiation of which the amount is proportional to neutron dose. Ionising radiation from neutron reactions changes the charge state of the gate oxide of the pMOSFET. The Gd-pMOSFETs were tested at a neutron beam port of HANARO research reactor and a 60 CO irradiation facility to investigate slow neutron response and gamma response, respectively. The voltage change was proportional to the accumulated slow neutron dose. The results from Gd coupled MOSFET neutron dosemeters shows an excellent sensitivity (15 - 16mV/cGy) and linearity to thermal neutrons with negligible background contamination. The results demonstrate the outstanding performance of the Gd coupled MOSFET neutron dosemeters clearly. The Gd-pMOSFET can also be used in a mixed radiation field by subtracting the voltage change of a pMOSFET without Gd from that of the Gd-pMOSFET

An instability due to the nonlinear coupling of p-modes to g-modes: Implications for coalescing neutron star binaries

International Nuclear Information System (INIS)

Weinberg, Nevin N.; Arras, Phil; Burkart, Joshua

2013-01-01

A weakly nonlinear fluid wave propagating within a star can be unstable to three-wave interactions. The resonant parametric instability is a well-known form of three-wave interaction in which a primary wave of frequency ω a excites a pair of secondary waves of frequency ω b + ω c ≅ ω a . Here we consider a nonresonant form of three-wave interaction in which a low-frequency primary wave excites a high-frequency p-mode and a low-frequency g-mode such that ω b + ω c >> ω a . We show that a p-mode can couple so strongly to a g-mode of similar radial wavelength that this type of nonresonant interaction is unstable even if the primary wave amplitude is small. As an application, we analyze the stability of the tide in coalescing neutron star binaries to p-g mode coupling. We find that the equilibrium tide and dynamical tide are both p-g unstable at gravitational wave frequencies f gw ≳ 20 Hz and drive short wavelength p-g mode pairs to significant energies on very short timescales (much less than the orbital decay time due to gravitational radiation). Resonant parametric coupling to the tide is, by contrast, either stable or drives modes at a much smaller rate. We do not solve for the saturation of the p-g instability and therefore we cannot say precisely how it influences the evolution of neutron star binaries. However, we show that if even a single daughter mode saturates near its wave breaking amplitude, the p-g instability of the equilibrium tide will (1) induce significant orbital phase errors (Δφ ≳ 1 radian) that accumulate primarily at low frequencies (f gw ≲ 50 Hz) and (2) heat the neutron star core to a temperature of T ∼ 10 10 K. Since there are at least ∼100 unstable p-g daughter pairs, Δφ and T are potentially much larger than these values. Tides might therefore significantly influence the gravitational wave signal and electromagnetic emission from coalescing neutron star binaries at much larger orbital separations than previously

Coupling of unidimensional neutron kinetics to thermal hydraulics in parallel channels; Acoplamiento de cinetica neutronica unidimensional a canales termohidraulicos en paralelo

Energy Technology Data Exchange (ETDEWEB)

Cecenas F, M.; Campos G, R.M. [IIE, Av. Reforma 113, Col. Palmira, Cuernavaca, Morelos (Mexico)]. e-mail: mcf@iie.org.mx

2003-07-01

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)

Neutron emission study after muon capture by nuclei

International Nuclear Information System (INIS)

Bouyssy, Alain.

1974-01-01

Muon capture by nuclei, used in the beginning for checking the weak interaction, is now a method of investigation of nuclear structure. Study of spectrum, asymmetry and polarization of emitted neutrons after polarized muon capture has been done in three directions: weak coupling constants, final state interaction, nuclear wave functions. The neutron intensity and helicity are very dependent of the neutron - residual nucleus interaction, while the asymmetry is sensitive to the wave functions used for the proton. Moreover if the induced tensor coupling constant is different from zero the asymmetry is increased. Longitudinal polarization experiments, with those for neutron intensity, would be of great interest to give informations on neutron asymmetry [fr

Development and Applications of a General Coupled Thermal-hydraulic/Neutronic Model for the Ringhals-3 Pressurized Water Reactor

International Nuclear Information System (INIS)

Staalek, Mathias

2008-03-01

Coupled calculations are important for the simulation of nuclear power plants when there is a strong feedback between the neutron kinetics and the thermal-hydraulics. A general coupled model of the Ringhals-3 Pressurized Water Reactor has been developed for this purpose. The development is outlined in the thesis with details given in the appended papers. A PARCS model was developed for the core calculations and a RELAP5 model for the thermal-hydraulic calculations. The RELAP5 model has 157 channels for modelling the flow in the fuel assemblies. This means that there is a one-one correspondence radially between the neutronic and thermal-hydraulic nodalization. This detailed mapping between the neutron kinetics and the thermal-hydraulics makes it possible to use the model for all kinds of transient. To provide realistic material data to the PARCS model, a cross-section interface was developed. With this interface one can import material data from a binary CASMO-4 library file into PARCS. Due to the one-to-one mapping, any any core loading can easily be considered. The PARCS model was benchmarked against measurements of the steady-state power distribution of Ringhals-3. The power shape was well reproduced by the model. Validational work for steady-state conditions of the thermal-hydraulic was also successfully performed. The most challenging part of the validation of a coupled model is for transients. This is much more difficult since the dynamics of the system becomes very important. Two transients that occurred at Ringhals-3 were chosen for the validational work. The first transient was a Load Rejection Transient. In general the model gave good results but some problems were experienced, e.g. the pressurizer pressure turned out to be more difficult to be correctly simulated. The second transient was a Loss of Feed Water transient. A malfunctioning feed water control valve closed, and therefore shut down the feed water supply to the steam generator in one of the

Development and Applications of a General Coupled Thermal-hydraulic/Neutronic Model for the Ringhals-3 Pressurized Water Reactor

Energy Technology Data Exchange (ETDEWEB)

Staalek, Mathias

2008-03-15

Coupled calculations are important for the simulation of nuclear power plants when there is a strong feedback between the neutron kinetics and the thermal-hydraulics. A general coupled model of the Ringhals-3 Pressurized Water Reactor has been developed for this purpose. The development is outlined in the thesis with details given in the appended papers. A PARCS model was developed for the core calculations and a RELAP5 model for the thermal-hydraulic calculations. The RELAP5 model has 157 channels for modelling the flow in the fuel assemblies. This means that there is a one-one correspondence radially between the neutronic and thermal-hydraulic nodalization. This detailed mapping between the neutron kinetics and the thermal-hydraulics makes it possible to use the model for all kinds of transient. To provide realistic material data to the PARCS model, a cross-section interface was developed. With this interface one can import material data from a binary CASMO-4 library file into PARCS. Due to the one-to-one mapping, any any core loading can easily be considered. The PARCS model was benchmarked against measurements of the steady-state power distribution of Ringhals-3. The power shape was well reproduced by the model. Validational work for steady-state conditions of the thermal-hydraulic was also successfully performed. The most challenging part of the validation of a coupled model is for transients. This is much more difficult since the dynamics of the system becomes very important. Two transients that occurred at Ringhals-3 were chosen for the validational work. The first transient was a Load Rejection Transient. In general the model gave good results but some problems were experienced, e.g. the pressurizer pressure turned out to be more difficult to be correctly simulated. The second transient was a Loss of Feed Water transient. A malfunctioning feed water control valve closed, and therefore shut down the feed water supply to the steam generator in one of the

Dosimetric analysis of BNCT - Boron Neutron Capture Therapy - coupled to 252Cf brachytherapy

International Nuclear Information System (INIS)

Brandao, Samia F.; Campos, Tarcisio P.R.

2009-01-01

The incidence of brain tumors is increasing in world population; however, the treatments employed in this type of tumor have a high rate of failure and in some cases have been considered palliative, depending on histology and staging of tumor. Its necessary to achieve the control tumor dose without the spread irradiation cause damage in the brain, affecting patient neurological function. Stereotactic radiosurgery is a technique that achieves this; nevertheless, other techniques that can be used on the brain tumor control must be developed, in order to guarantee lower dose on health surroundings tissues other techniques must be developing. The 252 Cf brachytherapy applied to brain tumors has already been suggested, showing promising results in comparison to photon source, since the active source is placed into the tumor, providing greater dose deposition, while more distant regions are spared. BNCT - Boron Neutron Capture Therapy - is another technique that is in developing to brain tumors control, showing theoretical superiority on the rules of conventional treatments, due to a selective irradiation of neoplasics cells, after the patient receives a borate compound infusion and be subjected to a epithermal neutrons beam. This work presents dosimetric studies of the coupling techniques: BNCT with 252 Cf brachytherapy, conducted through computer simulation in MCNP5 code, using a precise and well discretized voxel model of human head, which was incorporated a representative Glioblastoma Multiform tumor. The dosimetric results from MCNP5 code were exported to SISCODES program, which generated isodose curves representing absorbed dose rate in the brain. Isodose curves, neutron fluency, and dose components from BNCT and 252 Cf brachytherapy are presented in this paper. (author)

The Verification of Coupled Neutronics Thermal-Hydraulics Code NODAL3 in the PWR Rod Ejection Benchmark

Directory of Open Access Journals (Sweden)

Surian Pinem

2014-01-01

Full Text Available A coupled neutronics thermal-hydraulics code NODAL3 has been developed based on the few-group neutron diffusion equation in 3-dimensional geometry for typical PWR static and transient analyses. The spatial variables are treated by using a polynomial nodal method while for the neutron dynamic solver the adiabatic and improved quasistatic methods are adopted. In this paper we report the benchmark calculation results of the code against the OECD/NEA CRP PWR rod ejection cases. The objective of this work is to determine the accuracy of NODAL3 code in analysing the reactivity initiated accident due to the control rod ejection. The NEACRP PWR rod ejection cases are chosen since many organizations participated in the NEA project using various methods as well as approximations, so that, in addition to the reference solutions, the calculation results of NODAL3 code can also be compared to other codes’ results. The transient parameters to be verified are time of power peak, power peak, final power, final average Doppler temperature, maximum fuel temperature, and final coolant temperature. The results of NODAL3 code agree well with the PHANTHER reference solutions in 1993 and 1997 (revised. Comparison with other validated codes, DYN3D/R and ANCK, shows also a satisfactory agreement.

Quasiparticle Interactions in Neutron Matter for Applications in Neutron Stars

Science.gov (United States)

Wambach, J.; Anisworth, T. L.; Pines, D.

1993-01-01

A microscopic model for the quaisiparticle interaction in neutron matter is presented. Both particle-particle (pp) and particle-hole (ph) correlation are are included. The pp correlations are treated in semi-empirical way, while ph correlations are incorporated by solving coupled two-body equations for the particle hole interaction and the scattering amplitude on the Fermi sphere. The resulting integral equations self-consistently sum the ph reducible diagrams. Antisymmetry is kept at all stages and hence the forward-scattering sum rules are obeyed. Results for Landau parameters and transport coefficients in a density regime representing the crust of a neutron star are presented. We also estimate the S-1 gap parameter for neutron superfluidity and comment briefly on neutron-star implications.

Coupled neutronic-thermal-hydraulics analysis in a coolant subchannel of a PWR using CFD techniques

Energy Technology Data Exchange (ETDEWEB)

Ribeiro, Felipe P.; Su, Jian, E-mail: sujian@nuclear.ufrj.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

2017-07-01

The high capacity of Computational Fluid Dynamics code to predict multi-dimensional thermal-hydraulics behaviour and the increased availability of capable computer systems are making that method a good tool to simulate phenomena of thermal-hydraulics nature in nuclear reactors. However, since there are no neutron kinetics models available in commercial CFD codes to the present day, the application of CFD in the nuclear reactor safety analyses is still limited. The present work proposes the implementation of the point kinetics model (PKM) in ANSYS - Fluent to predict the neutronic behaviour in a Westinghouse Sequoyah nuclear reactor, coupling with the phenomena of heat conduction in the rod and thermal-hydraulics in the cooling fluid, via the reactivity feedback. Firstly, a mesh convergence and turbulence model study was performed, using the Reynolds-Average Navier-Stokes method, with square arrayed rod bundle featuring pitch to diameter ratio of 1:32. Secondly, simulations using the k-! SST turbulence model were performed with an axial distribution of the power generation in the fuel to analyse the heat transfer through the gap and cladding, and its in fluence on the thermal-hydraulics behaviour of the cooling fluid. The wall shear stress distribution for the centre-line rods and the dimensionless velocity were evaluated to validate the model, as well as the in fluence of the mass flow rate variation on the friction factor. The coupled model enabled to perform a dynamic analysis of the nuclear reactor during events of insertion of reactivity and shutdown of primary coolant pumps. (author)

LANSA: A large neutron scintillator array for neutron spectroscopy at Nova

International Nuclear Information System (INIS)

Nelson, M.B.; Cable, M.D.; Bennett, C.K.; Mant, G.

1992-01-01

A very sensitive neutron time-of-flight spectrometer is now in use at Nova. LANSA consists of 960 channels of a neutron sensitive liquid scintillator (10 x lO x lO cm) coupled to a photomultiplier tube followed by a discriminator, TDC and ADC to allow the measurement of neutron arrival time as well as pulse size. LANSA is capable of measuring yields as low as 2.3 x 10 5 DT neutrons (100 detected hits) with resolution of 2.3 ns (170 key for 14 MeV neutrons with 20 m flight path). Shielding and collimation provide background levels low enough to allow measurement of secondary and tertiary reaction neutrons. Details of design, testing, calibration and experimental results will be presented. This work was performed under the auspices of the US Department of Energy by Lawrence Livermore National Laboratory under contract No. W-7405-ENG-48

Applications for coupled core neutronics and thermal-hydraulic models

International Nuclear Information System (INIS)

Eller, J.

1996-01-01

The unprecedented increases in computing capacity that have occurred during the last decade have affected our sciences, and thus our lives, to an extent that is difficult to overstate. All indications are that this trend will continue for years to come. Nuclear reactor systems analysis is one of many areas of engineering that has changed dramatically as a result of this evolution. Our ability to model the various mechanical and physical systems in greater and greater detail has allowed significant improvements in operational efficiency in spite of increasing regulatory requirements. Many of these efficiencies result from the use of more complex and geometrically detailed computer modeling, which is used to justify a reduction or elimination of some of the conservatisms required by earlier, less sophisticated analyses. And more recently, as our industries open-quotes downsize,close quotes efforts are being made to find ways to use the ever-increasing computing capacity to design systems that accomplish more work, in less time, and with fewer people. The balance of this paper discusses some of the visions that Duke Power Company feels would most benefit their particular methodologies. One of the concepts receiving a lot of attention involves an automated coupling of a thermal-hydraulic plant systems analysis model to a three-dimensional core neutronics program. The thermal-hydraulic analysis of several postulated system transients incorporates large conservatisms because of limited ability to model complex time-dependent asymmetric heat sources in adequate geometric detail. For these transients, the core behavior is closely coupled with the thermal-hydraulic behavior of the total plant system and vice versa. Steam-line break, uncontrolled rod withdrawal, and rod drop anayses are likely to benefit most from this type of linked process

Project and construction of counting system for neutron probe

International Nuclear Information System (INIS)

Monteiro, W.P.

1985-01-01

A counting system was developed for coupling neutron probe aiming to register pulses produced by slow neutron interaction in the detector. The neutron probe consists of fast neutron source, thermal neutron detector, amplifier circuit and pulse counting circuit. The counting system is composed by counting circuit, timer and signal circuit. (M.C.K.)

Evaluation of the fluence to dose conversion coefficients for high energy neutrons using a voxel phantom coupled with the GEANT4 code

CERN Document Server

Paganini, S

2005-01-01

Crews working on present-day jet aircraft are a large occupationally exposed group with a relatively high average effective dose from Galactic cosmic radiation. Crews of future high-speed commercial flying at higher altitudes would be even more exposed. To help reduce the significant uncertainties in calculations of such exposures, the male adult voxels phantom MAX, developed in the Nuclear Energy Department of Pernambuco Federal University in Brazil, has been coupled with the Monte Carlo simulation code GEANT4. This toolkit, distributed and upgraded from the international scientific community of CERN/Switzerland, simulates thermal to ultrahigh energy neutrons transport and interactions in the matter. The high energy neutrons are pointed as the component that contribute about 70% of the neutron effective dose that represent the 35% to 60% total dose at aircraft altitude. In this research calculations of conversion coefficients from fluence to effective dose are performed for neutrons of energies from 100 MeV ...

Proton energy dependence of slow neutron intensity

International Nuclear Information System (INIS)

Teshigawara, Makoto; Harada, Masahide; Watanabe, Noboru; Kai, Tetsuya; Sakata, Hideaki; Ikeda, Yujiro

2001-01-01

The choice of the proton energy is an important issue for the design of an intense-pulsed-spallation source. The optimal proton beam energy is rather unique from a viewpoint of the leakage neutron intensity but no yet clear from the slow-neutron intensity view point. It also depends on an accelerator type. Since it is also important to know the proton energy dependence of slow-neutrons from the moderators in a realistic target-moderator-reflector assembly (TMRA). We studied on the TMRA proposed for Japan Spallation Neutron Source. The slow-neutron intensities from the moderators per unit proton beam power (MW) exhibit the maximum at about 1-2 GeV. At higher proton energies the intensity per MW goes down; at 3 and 50 GeV about 0.91 and 0.47 times as low as that at 1 GeV. The proton energy dependence of slow-neutron intensities was found to be almost the same as that of total neutron yield (leakage neutrons) from the same bare target. It was also found that proton energy dependence was almost the same for the coupled and decoupled moderators, regardless the different moderator type, geometry and coupling scheme. (author)

Pulsed neutron source and instruments at neutron facility

Energy Technology Data Exchange (ETDEWEB)

Teshigawara, Makoto; Aizawa, Kazuya; Suzuki, Jun-ichi; Morii, Yukio; Watanabe, Noboru [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1997-11-01

We report the results of design studies on the optimal target shape, target - moderator coupling, optimal layout of moderators, and neutron instruments for a next generation pulsed spallation source in JAERI. The source utilizes a projected high-intensity proton accelerator (linac: 1.5 GeV, {approx}8 MW in total beam power, compressor ring: {approx}5 MW). We discuss the target neutronics, moderators and their layout. The sources is designed to have at least 30 beam lines equipped with more than 40 instruments, which are selected tentatively to the present knowledge. (author)

Role of nuclear couplings in the inelastic excitation of weakly-bound neutron-rich nuclei

Energy Technology Data Exchange (ETDEWEB)

Dasso, C.H. [Niels Bohr Institute, Copenhagen (Denmark); Lenzi, S.M.; Vitturi, A. [Universita di Padova (Italy)

1996-12-31

Much effort is presently devoted to the study of nuclear systems far from the stability line. Particular emphasis has been placed in light systems such as {sup 11}Li, {sup 8}B and others, where the very small binding energy of the last particles causes their density distribution to extend considerably outside of the remaining nuclear core. Some of the properties associated with this feature are expected to characterize also heavier systems in the vicinity of the proton or neutron drip lines. It is by now well established that low-lying concentrations of multipole strength arise from pure configurations in which a peculiar matching between the wavelength of the continuum wavefunction of the particles and the range of the weakly-bound hole states occurs. To this end the authors consider the break-up of a weakly-bound system in a heavy-ion collision and focus attention in the inelastic excitation of the low-lying part of the continuum. They make use of the fact that previous investigations have shown that the multipole response in this region is not of a collective nature and describe their excited states as pure particle-hole configurations. Since the relevant parameter determining the strength distributions is the binding energy of the last bound orbital they find it most convenient to use single-particle wavefunctions generated by a sperical square-well potential with characteristic nuclear dimensions and whose depth has been adjusted to give rise to a situation in which the last occupied neutron orbital is loosely-bound. Spin-orbit couplings are, for the present purpose, ignored. The results of this investigation clearly indicate that nuclear couplings have the predominant role in causing projectile dissociation in many circumstances, even at bombarding energies remarkably below the Coulomb barrier.

Prototype coupling of the CFD software ansys CFX with the 3D neutron kinetic core model DYN3D - 249

International Nuclear Information System (INIS)

Kliem, S.; Rohde, U.; Schutze, J.; Frank, Th.

2010-01-01

The CFD code ANSYS CFX has been coupled with the neutron-kinetic core model DYN3D. ANSYS CFX calculates the fluid dynamics and related transport phenomena in the reactor's coolant and provides the corresponding data to DYN3D. In the fluid flow simulation of the coolant, the core itself is modeled within the porous body approach. DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the prototype that is currently available, the coupling is restricted to single-phase flow problems. In the time domain an explicit coupling of the codes has been implemented so far. Steady-state and transient verification calculations for a small-size test problem confirm the correctness of the implementation of the prototype coupling. This test problem was a mini-core consisting of nine real-size fuel assemblies. Comparison was performed with the DYN3D standalone code. In the steady state, the effective multiplication factor obtained by the ANSYS CFX/DYN3D codes shows a deviation of 9.8 pcm from the DYN3D stand-alone solution. This difference can be attributed to the use of different water property packages in the two codes. The transient test case simulated the withdrawal of the control rod from the central fuel assembly at hot zero power. Power increase during the introduction of positive reactivity and power reduction due to fuel temperature increase are calculated in the same manner by the coupled and the stand-alone codes. The maximum values reached during the power rise differ by about 1 MW at a power level of 50 MW. Beside the different water property packages, these differences are caused by the use of different flow solvers. (authors)

Search for EC-decayed neutron-deficient actinide isotopes using gas-jet coupled JAERI-ISOL

Energy Technology Data Exchange (ETDEWEB)

Tsukada, Kazuaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1997-07-01

To study the nuclear properties of unknown neutron deficient actinide isotopes which decay mainly via orbital electron capture (EC), we have developed a composite system consisting of a gas-jet transport apparatus and a thermal ion-source at the JAERI-ISOL. With this system, search for {sup 236}Am produced in the {sup 235}U({sup 6}Li, 5n) reaction has been performed. Pu KX-rays associated with the EC decay of {sup 236}Am are observed at the mass-236 fraction. The half-life of {sup 236}Am is evaluated to be 4.4min. The outline of the gas-jet coupled JAERI-ISOL system and typical performance are given. (author)

Study of sampling rate influence on neutron-gamma discrimination with stilbene coupled to a silicon photomultiplier.

Science.gov (United States)

Zhang, Jinglong; Moore, Michael E; Wang, Zhonghai; Rong, Zhou; Yang, Chaowen; Hayward, Jason P

2017-10-01

Choosing a digitizer with an appropriate sampling rate is often a trade-off between performance and economy. The influence of sampling rates on the neutron-gamma Pulse Shape Discrimination (PSD) with a solid stilbene scintillator coupled to a Silicon Photomultiplier was investigated in this work. Sampling rates from 125MSPS to 2GSPS from a 10-bit digitizer were used to collect detector pulses produced by the interactions of a Cf-252 source. Due to the decreased signal-to-noise ratio (SNR), the PSD performance degraded with reduced sampling rates. The reason of PSD performance degradation was discussed. Then, an efficient combination of filtering and digital signal processing (DSP) was then applied to suppress the timing noise and electronic background noise. The results demonstrate an improved PSD performance especially at low sampling rates, down to 125MSPS. Using filtering and DSP, the ascribed Figure of Merit (FOM) at 125keV ee (± 10keV ee ) increased from 0.95 to 1.02 at 125MSPS. At 300keV ee and above, all the FOMs are better than 2.00. Our study suggests that 250MSPS is a good enough sampling rate for neutron-gamma discrimination in this system in order to be sensitive to neutrons at and above ~ 125keV ee . Copyright © 2017 Elsevier Ltd. All rights reserved.

Coupling calculation of CFD-ACE computational fluid dynamics code and DeCART whole-core neutron transport code for development of numerical reactor

Energy Technology Data Exchange (ETDEWEB)

Shin, Chang Hwan; Seo, Kyong Won; Chun, Tae Hyun; Kim, Kang Seog

2005-03-15

Code coupling activities have so far focused on coupling the neutronics modules with the CFD module. An interface module for the CFD-ACE/DeCART coupling was established as an alternative to the original STAR-CD/DeCART interface. The interface module for DeCART/CFD-ACE was validated by single-pin model. The optimized CFD mesh was decided through the calculation of multi-pin model. It was important to consider turbulent mixing of subchannels for calculation of fuel temperature. For the parallel calculation, the optimized decompose process was necessary to reduce the calculation costs and setting of the iteration and convergence criterion for each code was important, too.

Coupling calculation of CFD-ACE computational fluid dynamics code and DeCART whole-core neutron transport code for development of numerical reactor

International Nuclear Information System (INIS)

Shin, Chang Hwan; Seo, Kyong Won; Chun, Tae Hyun; Kim, Kang Seog

2005-03-01

Code coupling activities have so far focused on coupling the neutronics modules with the CFD module. An interface module for the CFD-ACE/DeCART coupling was established as an alternative to the original STAR-CD/DeCART interface. The interface module for DeCART/CFD-ACE was validated by single-pin model. The optimized CFD mesh was decided through the calculation of multi-pin model. It was important to consider turbulent mixing of subchannels for calculation of fuel temperature. For the parallel calculation, the optimized decompose process was necessary to reduce the calculation costs and setting of the iteration and convergence criterion for each code was important, too

Active neutron multiplicity analysis and Monte Carlo calculations

International Nuclear Information System (INIS)

Krick, M.S.; Ensslin, N.; Langner, D.G.; Miller, M.C.; Siebelist, R.; Stewart, J.E.; Ceo, R.N.; May, P.K.; Collins, L.L. Jr

1994-01-01

Active neutron multiplicity measurements of high-enrichment uranium metal and oxide samples have been made at Los Alamos and Y-12. The data from the measurements of standards at Los Alamos were analyzed to obtain values for neutron multiplication and source-sample coupling. These results are compared to equivalent results obtained from Monte Carlo calculations. An approximate relationship between coupling and multiplication is derived and used to correct doubles rates for multiplication and coupling. The utility of singles counting for uranium samples is also examined

Proton impurity in the neutron matter: a nuclear polaron problem

Energy Technology Data Exchange (ETDEWEB)

Kutschera, M [Institute of Nuclear Physics, Cracow (Poland); Wojcik, W [Politechnika Krakowska, Cracow (Poland)

1992-10-01

We study interactions of a proton impurity with density oscillations of the neutron matter in a Debye approximation. The proton-phonon coupling is of the deformation-potential type at long wavelengths. It is weak at low density and increases with the neutron matter density. We calculate the proton`s effective mass perturbatively for a weak coupling, and use a canonical transformation technique for stronger couplings. The proton`s effective mass grows significantly with density, and at higher densities the proton impurity can be localized. This behaviour is similar to that of the polaron in solids. We obtain properties of the localized proton in the strong coupling regime from variational calculations, treating the neutron in the Thomas-Fermi approximation. (author). 14 refs, 8 figs.

Study on the novel neutron-to-proton convertor for improving the detection efficiency of a triple GEM based fast neutron detector

International Nuclear Information System (INIS)

Wang Xiaodong; Yang Lei; Zhang Chunhui; Hu Bitao; Yang Herun; Zhang Junwei; Ren Zhongguo; Ha Ri-Ba-La; An Luxing

2015-01-01

A high-efficiency fast neutron detector prototype based on a triple Gas Electron Multiplier (GEM) detector, which, coupled with a novel multi-layered high-density polyethylene (HDPE) as a neutron-to-proton converter for improving the neutron detection efficiency, is introduced and tested with the Am-Be neutron source in the Institute of Modern Physics (IMP) at Lanzhou in the present work. First, the developed triple GEM detector is tested by measuring its effective gain and energy resolution with "5"5Fe X-ray source to ensure that it has a good performance. The effective gain and obtained energy resolution is 5.0 × 10"4 and around 19.2%, respectively. Secondly, the novel multi-layered HDPE converter is coupled with the cathode of the triple GEM detector making it a high-efficiency fast neutron detector. Its effective neutron response is four times higher than that of the traditional single-layered conversion technique when the converter layer number is 38. (authors)

Recent advances in neutron tomography

International Nuclear Information System (INIS)

McFarland, E.; Massachusetts Inst. of Technology, Cambridge, MA; Lanza, R.

1993-01-01

Neutron imaging has been shown to be an excellent imaging tool for many nondestructive evaluation applications. Significantly improved contrast over X-ray images is possible for materials commonly found in engineering assemblies. The major limitations have been the neutron source and detection. A low cost, position sensitive neutron tomography detector system has been designed and built based on an electro-optical detector system using a LiF-ZnS scintillator screen and a cooled charge coupled device. This detector system can be used for neutron radiography as well as two and three-dimensional neutron tomography. Calculated performance of the system predicted near-quantum efficiency for position sensitive neutron detection. Experimental data was recently taken using this system at McClellan Air Force Base, Air Logistics Center, Sacramento, CA. With increased availability of low cost neutron sources and advanced image processing, neutron tomography will become an increasingly important nondestructive imaging method

Set of thermal neutron-scattering experiments for the Weapons Neutron Research Facility

International Nuclear Information System (INIS)

Brugger, R.M.

1975-12-01

Six classes of experiments form the base of a program of thermal neutron scattering at the Weapons Neutron Research (WNR) Facility. Three classes are to determine the average microscopic positions of atoms in materials and three are to determine the microscopic vibrations of these atoms. The first three classes concern (a) powder sample neutron diffraction, (b) small angle scattering, and (c) single crystal Laue diffraction. The second three concern (d) small kappa inelastic scattering, (e) scattering surface phonon measurements, and (f) line widths. An instrument to couple with the WNR pulsed source is briefly outlined for each experiment

Two-quasineutron states in 98248Cf and 98250Cf and the neutron-neutron residual interactions

International Nuclear Information System (INIS)

Katori, K.; Ahmad, I.; Friedman, A. M.

2008-01-01

Two-quasineutron states in 248 Cf and 250 Cf were investigated by single-neutron transfer reactions, 249 Cf(d,t) 248 Cf and 249 Cf(d,p) 250 Cf. The majority of levels observed were assigned to 12 bands in 248 Cf and six bands in 250 Cf, constructed from the single-particle states in neighboring Cf nuclei. The effective two-body interactions between two odd neutrons coupled outside a deformed core were deduced from the differences in the energies of the bandheads formed by the parallel and antiparallel coupling of the intrinsic spins of the two single-particle states

Layered magnets: polarized neutron reflection studies

Energy Technology Data Exchange (ETDEWEB)

Zabel, H; Schreyer, A [Ruhr-Univ. Bochum, Lehrstuhl fuer Experimentalphysik/Festkoerperphysik, Bochum (Germany)

1996-11-01

Neutron reflectivity measurements from extended surfaces, thin films and superlattices provide information on the chemical profile parallel to the film normal, including film thicknesses, average composition and interfacial roughness parameters. Reflectivity measurements with polarized neutrons are particularly powerful for analyzing the magnetic density profiles in thin films and superlattices in addition to chemical profiles. The basic theory of polarized neutron reflectivity is provided, followed by some examples and more recent applications concerning polarized neutron reflectivity studies from exchange coupled Fe/Cr superlattices. (author) 5 figs., 13 refs.

The OECD/NEA/NSC PBMR 400 MW coupled neutronics thermal hydraulics transient benchmark: transient results - 290

International Nuclear Information System (INIS)

Strydom, G.; Reitsma, F.; Ngeleka, P.T.; Ivanov, K.N.

2010-01-01

The PBMR is a High-Temperature Gas-cooled Reactor (HTGR) concept developed to be built in South Africa. The analysis tools used for core neutronic design and core safety analysis need to be verified and validated, and code-to-code comparisons are an essential part of the V and V plans. As part of this plan the PBMR 400 MWth design and a representative set of transient exercises are defined as an OECD benchmark. The scope of the benchmark is to establish a series of well defined multi-dimensional computational benchmark problems with a common given set of cross sections, to compare methods and tools in coupled neutronics and thermal hydraulics analysis with a specific focus on transient events. This paper describes the current status of the benchmark project and shows the results for the six transient exercises, consisting of three Loss of Cooling Accidents, two Control Rod Withdrawal transients, a power load-follow transient, and a Helium over-cooling Accident. The participants' results are compared using a statistical method and possible areas of future code improvement are identified. (authors)

Neutronics/Thermo-fluid Coupled Analysis of PMR-200 Equilibrium Cycle by CAPP/GAMMA+ Code System

Energy Technology Data Exchange (ETDEWEB)

Lee, Hyun Chul; Tak, Nam-il [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

The equilibrium core was obtained by performing CAPP stand-alone multi-cycle depletion calculation with critical rod position search. In this work, a code system for coupled neutronics and thermo-fluids simulation was developed using CAPP and GAMMA+ codes. A server program, INTCA, controls the two codes for coupled calculations and performs the mapping between the variables of the two codes based on the nodalization of the two codes. In order to extend the knowledge about the coupled behavior of a prismatic VHTR, the CAPP/GAMMA+ code system was applied to steady state performance analysis of PMR-200. The coupled calculation was carried out for the equilibrium core of PMR-200 from BOC to EOC. The peak fuel temperature was predicted to be 1372 .deg. C near MOC. However, the cycle-average fuel temperature was calculated as 1230 .deg. C, which is slightly below the design target of 1250 .deg. C. In addition, significant impact of the bypass flow on the central reflector temperature was found. Without bypass flow, the temperature of the active core region was slightly decreased while the temperature of the central and side reflector region was increased much. The both changes in the temperature increase the multiplication factor and the total change of the multiplication factor was more than 300 pcm. On the other hand, the effect of the bypass flow on the power density profile was not significant.

Experimental possibilities and fast neutron dose map of the fast neutron fields at the RB reactor facility

International Nuclear Information System (INIS)

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

1993-01-01

The RB is an unshielded, zero power nuclear facility with natural and enriched uranium fuel (2% and 80%) and D 2 O as moderator. It is possible to create different configurations of non-reflected and partially reflected critical systems and to make experiments in the fields of thermal neutrons. The fields of fast neutrons with 'softened' fission spectrum are made by modifying the system: modified experimental fuel channel EFC, coupled fast-thermal system in two configurations CFTS-1 and CFTS-2, coupled fast-thermal core HERBE. The intermediate and fast neutron absorbed doses in fast neutron fields are given. In first configuration of RB reactor it was almost impossible to perform dosimetric and other experiments. By creating these fields, with in our circumstances available fuel elements, the possibilities for different experiments are greatly improved. Now we can irradiate food samples, soil samples, electronic devices, study material properties, perform various dosimetry experiments, etc. (1 tab.)

Nonlinear mode coupling in rotating stars and the r-mode instability in neutron stars

International Nuclear Information System (INIS)

Schenk, A.K.; Arras, P.; Flanagan, E.E.; Teukolsky, S.A.; Wasserman, I.

2002-01-01

We develop the formalism required to study the nonlinear interaction of modes in rotating Newtonian stars, assuming that the mode amplitudes are only mildly nonlinear. The formalism is simpler than previous treatments of mode-mode interactions for spherical stars, and simplifies and corrects previous treatments for rotating stars. At linear order, we elucidate and extend slightly a formalism due to Schutz, show how to decompose a general motion of a rotating star into a sum over modes, and obtain uncoupled equations of motion for the mode amplitudes under the influence of an external force. Nonlinear effects are added perturbatively via three-mode couplings, which suffices for moderate amplitude modal excitations; the formalism is easy to extend to higher order couplings. We describe a new, efficient way to compute the modal coupling coefficients, to zeroth order in the stellar rotation rate, using spin-weighted spherical harmonics. The formalism is general enough to allow computation of the initial trends in the evolution of the spin frequency and differential rotation of the background star. We apply this formalism to derive some properties of the coupling coefficients relevant to the nonlinear interactions of unstable r modes in neutron stars, postponing numerical integrations of the coupled equations of motion to a later paper. First, we clarify some aspects of the expansion in stellar rotation frequency Ω that is often used to compute approximate mode functions. We show that, in zero-buoyancy stars, the rotational modes (those modes whose frequencies vanish as Ω→0) are orthogonal to zeroth order in Ω. From an astrophysical viewpoint, the most interesting result of this paper is that many couplings of r modes to other rotational modes are small: either they vanish altogether because of various selection rules, or they vanish to lowest order in Ω or in compressibility. In particular, in zero-buoyancy stars, the coupling of three r modes is forbidden

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

Research trends in neutron physics

International Nuclear Information System (INIS)

Lynn, J.E.

1976-01-01

The trends in neutron research are discussed from the viewpoints of development of pulsed neutron sources, the ingenuity of specialization of instrumentation and experimental techniques, and research programs. The latter comprise the large and still expanding requirements of nuclear data for nuclear power technology, the requirements of other fundamental sciences, and the experimental and theoretical developments required for a more fundamental understanding of the subject of neutron and related nuclear reactions itself. The general conclusion is that high energy resolution coupled with high intensity for detecting weak reactions provides the key to further progress, and that (provided financial limitations do not stifle the further development of experimental facilities, particularly neutron sources) the subject of neutron physics still has a long and fruitful future

Systems and methods for neutron detection using scintillator nano-materials

Science.gov (United States)

Letant, Sonia Edith; Wang, Tzu-Fang

2016-03-08

In one embodiment, a neutron detector includes a three dimensional matrix, having nanocomposite materials and a substantially transparent film material for suspending the nanocomposite materials, a detector coupled to the three dimensional matrix adapted for detecting a change in the nanocomposite materials, and an analyzer coupled to the detector adapted for analyzing the change detected by the detector. In another embodiment, a method for detecting neutrons includes receiving radiation from a source, converting neutrons in the radiation into alpha particles using converter material, converting the alpha particles into photons using quantum dot emitters, detecting the photons, and analyzing the photons to determine neutrons in the radiation.

Analysis of the ITER computational shielding benchmark with the Monte Carlo TRIPOLI-4® neutron gamma coupled calculations

International Nuclear Information System (INIS)

Lee, Yi-Kang

2016-01-01

Highlights: • Verification and validation of TRIPOLI-4 radiation transport calculations for ITER shielding benchmark. • Evaluation of CEA-V5.1.1 and FENDL-3.0 nuclear data libraries on D–T fusion neutron continuous energy transport calculations. • Advances in nuclear analyses for nuclear heating and radiation damage in iron. • This work also demonstrates that the “safety factors” concept is necessary in the nuclear analyses of ITER. - Abstract: With the growing interest in using the continuous-energy TRIPOLI-4 ® Monte Carlo radiation transport code for ITER applications, a key issue that arises is whether or not the released TRIPOLI-4 code and its associated nuclear data libraries are verified and validated for the D–T fusion neutronics calculations. Previous published benchmark results of TRIPOLI-4 code on the ITER related activities have concentrated on the first wall loading, the reactor dosimetry, the nuclear heating, and the tritium breeding ratio. To enhance the TRIPOLI-4 verification and validation on neutron-gamma coupled calculations for fusion device application, the computational ITER shielding benchmark of M. E. Sawan was performed in this work by using the 2013 released TRIPOLI-4.9S code and the associated CEA-V5.1.1 data library. First wall, blanket, vacuum vessel and toroidal field magnet of the inboard and outboard components were fully modelled in this 1-D toroidal cylindrical benchmark. The 14.1 MeV source neutrons were sampled from a uniform isotropic distribution in the plasma zone. Nuclear responses including neutron and gamma fluxes, nuclear heating, and material damage indicator were benchmarked against previous published results. The capabilities of the TRIPOLI-4 code on the evaluation of above physics parameters were presented. The nuclear data library from the new FENDL-3.0 evaluation was also benchmarked against the CEA-V5.1.1 results for the neutron transport calculations. The results show that both data libraries can be

A neutron irradiator applied to cancer treatment

International Nuclear Information System (INIS)

Campos, Tarcisio P.R.; Andrade, Ana P. de

2000-01-01

Cancer and the way of treating it with neutron capture therapy are addressed. This paper discusses also the type of neutron facilities used to treat cancer around the world, as follow: discrete neutron sources, accelerators, and nuclear reactors. The major features of an epithermal neutron irradiation facility applied to BNCT treatment are addressed. The main goal is to give another choice of neutron irradiators to be set in a hospital. The irradiation facility embeds a set of 252 Cf neutron source coupled with a homogeneous mixture of uranium-zirconium hydride alloy containing 8.4 wt % uranium enriched to 20% U 235 . The facility delivers an epithermal neutron beam with low background of fast neutron and gamma rays. The N particle transport code (MCNP-4A) has been used during the simulation in order to achieve the desired configurations and to estimate the multiplication factor, k eff . The present facility loaded with 30 mg of 252 Cf neutron source generates an external beam with an intensity of 10 7 n/cm 2 .s on the spectrum of 4 eV to 40 KeV. The 252 Cf - facility coupled with fissile material was able to amplify the epithermal flux to 10 8 n/cm 2 .s, maintaining the figure-of-merits represented by the ratios of the fast dose and gamma dose in air per epithermal neutron flux closed to those values presented by BMRR, MITR-II and Petten Reactor. The medical irradiation facility loaded with 252 Cf- 235 U can be a choice for BNCT. (author)

Review on individual neutron dosimetry

International Nuclear Information System (INIS)

Portal, M.

1983-01-01

Up to now, nuclear energy workers in relation to neutron radiations were few. Fast development of nuclear energy lead us to study, for future, individual dosimetry techniques which are autonomous, more accurate and cheaper. The future dosemeter will be a couple: fast neutron dosemeter and slow neutron dosemeter. The different current studies concerning this ''composite'' dosemeter are described. In 1984-1985, operation of a ''non-homogeneous, composite'' dosemeter is foreseen; later on, an ''homogeneous composite'' dosemeter that is to say a dosemeter which needs same basis techniques [fr

Neutron Electric Dipole Moment from colored scalars⋆

Directory of Open Access Journals (Sweden)

Fajfer Svjetlana

2014-01-01

Full Text Available We present new contributions to the neutron electric dipole moment induced by a color octet, weak doublet scalar, accommodated within a modified Minimal Flavor Violating framework. These flavor non-diagonal couplings of the color octet scalar might account for an assymmetry of order 3 × 10−3 for aCP(D0 → K−K+ − aCP(D0 → π+π− at tree level. The same couplings constrained by this assymmetry also induce two-loop contributions to the neutron electric dipole moment. We find that the direct CP violating asymmetry in neutral D-meson decays is more constraining on the allowed parameter space than the current experimental bound on neutron electric dipole moment.

Application of the decoupling scheme on complex neutron-gamma shielding problems

Energy Technology Data Exchange (ETDEWEB)

Feher, S. [Institute of Nuclear Technology, Technical University of Budapest, Budapest (Hungary); Leege, P.F.A. de; Hoogenboom, J.E.; Kloosterman, J.L. [Interfaculty Reactor Institute, Delft University of Technology, Delft (Netherlands)

2000-03-01

Coupled neutron-gamma shielding calculations using S{sub n} transport theory can be time consuming, especially for two- and three-dimensional geometries. In general, the CPU time of these calculations increases stronger than linear with increasing number of neutron and gamma energy groups, and depends on the order of Legendre expansion and number of S{sub n} directions used. This fact induced the idea of the decoupling method, which seems applicable to accelerate coupled neutron-gamma shielding calculations. The data included in a combined neutron-gamma library can be readily separated into a library containing neutron data only and another library containing gamma data only. Separate calculations for neutrons and gammas are performed on complex geometries using a different Legendre order expansion for neutrons and gammas. CPU savings of 60 to 85% can be achieved for the two-dimensional DORT and three-dimensional TORT calculations respectively. (author)

Low dimensional neutron moderators for enhanced source brightness

DEFF Research Database (Denmark)

Mezei, Ferenc; Zanini, Luca; Takibayev, Alan

2014-01-01

In a recent numerical optimization study we have found that liquid para-hydrogen coupled cold neutron moderators deliver 3–5 times higher cold neutron brightness at a spallation neutron source if they take the form of a flat, quasi 2-dimensional disc, in contrast to the conventional more voluminous...... for cold neutrons. This model leads to the conclusions that the optimal shape for high brightness para-hydrogen neutron moderators is the quasi 1-dimensional tube and these low dimensional moderators can also deliver much enhanced cold neutron brightness in fission reactor neutron sources, compared...... to the much more voluminous liquid D2 or H2 moderators currently used. Neutronic simulation calculations confirm both of these theoretical conclusions....

ARCADIAR - A New Generation of Coupled Neutronics / Core Thermal- Hydraulics Code System at AREVA NP

International Nuclear Information System (INIS)

Curca-Tivig, Florin; Merk, Stephan; Pautz, Andreas; Thareau, Sebastien

2007-01-01

Anticipating future needs of our customers and willing to concentrate synergies and competences existing in the company for the benefit of our customers, AREVA NP decided in 2002 to develop the next generation of coupled neutronics/ core thermal-hydraulic (TH) code systems for fuel assembly and core design calculations for both, PWR and BWR applications. The global CONVERGENCE project was born: after a feasibility study of one year (2002) and a conceptual phase of another year (2003), development was started at the beginning of 2004. The present paper introduces the CONVERGENCE project, presents the main feature of the new code system ARCADIA R and concludes on customer benefits. ARCADIA R is designed to meet AREVA NP market and customers' requirements worldwide. Besides state-of-the-art physical modeling, numerical performance and industrial functionality, the ARCADIA R system is featuring state-of-the-art software engineering. The new code system will bring a series of benefits for our customers: e.g. improved accuracy for heterogeneous cores (MOX/ UOX, Gd...), better description of nuclide chains, and access to local neutronics/ thermal-hydraulics and possibly thermal-mechanical information (3D pin by pin full core modeling). ARCADIA is a registered trademark of AREVA NP. (authors)

A Long-Pulse Spallation Source at Los Alamos: Facility description and preliminary neutronic performance for cold neutrons

International Nuclear Information System (INIS)

Russell, G.J.; Weinacht, D.J.; Pitcher, E.J.; Ferguson, P.D.

1998-03-01

The Los Alamos National Laboratory has discussed installing a new 1-MW spallation neutron target station in an existing building at the end of its 800-MeV proton linear accelerator. Because the accelerator provides pulses of protons each about 1 msec in duration, the new source would be a Long Pulse Spallation Source (LPSS). The facility would employ vertical extraction of moderators and reflectors, and horizontal extraction of the spallation target. An LPSS uses coupled moderators rather than decoupled ones. There are potential gains of about a factor of 6 to 7 in the time-averaged neutron brightness for cold-neutron production from a coupled liquid H 2 moderator compared to a decoupled one. However, these gains come at the expense of putting ''tails'' on the neutron pulses. The particulars of the neutron pulses from a moderator (e.g., energy-dependent rise times, peak intensities, pulse widths, and decay constant(s) of the tails) are crucial parameters for designing instruments and estimating their performance at an LPSS. Tungsten is the reference target material. Inconel 718 is the reference target canister and proton beam window material, with Al-6061 being the choice for the liquid H 2 moderator canister and vacuum container. A 1-MW LPSS would have world-class neutronic performance. The authors describe the proposed Los Alamos LPSS facility, and show that, for cold neutrons, the calculated time-averaged neutronic performance of a liquid H 2 moderator at the 1-MW LPSS is equivalent to about 1/4th the calculated neutronic performance of the best liquid D 2 moderator at the Institute Laue-Langevin reactor. They show that the time-averaged moderator neutronic brightness increases as the size of the moderator gets smaller

DORT-TD/THERMIX solutions for the OECD/NEA/NSC PBMR400 MW coupled neutronics thermal hydraulics transient benchmark

International Nuclear Information System (INIS)

Tyobeka, Bismark; Pautz, Andreas; Ivanov, Kostadin

2008-01-01

In new reactor designs that are still under review such as the PBMR, not much experimental data exists to benchmark newly developed computer codes against. Such a situation requires that nuclear engineers and designers of this novel reactor design must resort to the validation of a newly developed code through a code-to-code benchmarking exercise because there are validated codes that are currently in use to analyze this reactor design, albeit very few of them. There are numerous HTR core physics benchmarks that are currently being pursued by different organizations, for different purposes. One such benchmark exercise is the PBMR-400 MW OECD/NEA/NSC coupled neutronics/thermal hydraulics transient benchmark. In this paper, a newly developed coupled neutronics thermal hydraulics code system, DORT-TD/THERMIX with both transport and diffusion theory options, is used to simulate the transient scenarios in the above-mentioned benchmark problem. Steady-state calculations results are compared with selected participants' results as well as transient models in which the diffusion and transport theory solutions of the same code system are directly compared. Several sensitivity studies are also shown in order to determine how much the change in certain parameters influences the overall behaviour of a given transient. It is shown in this paper that DORT-TD/THERMIX is a versatile tool which can be deployed for design and safety analyses of high temperature reactors of pebble-bed type. (authors)

Few-MeV neutrons incident on yttrium

International Nuclear Information System (INIS)

Budtz-Jorgensen, C.; Guenther, P.; Smith, A.; Whalen, J.

1982-09-01

Neutron total and scattering cross sections of elemental yttrium are measured in the few-MeV region with broad resolutions. The total-cross section measurements extend from approx. = 0.5 to 4.2 MeV in steps of less than or equal to 0.1 MeV. Neutron elastic- and inelastic-scattering cross sections are measured from approx. = 1.5 to 4.0 MeV, at incident-neutron energy intervals of less than or equal to 50 keV and at ten or more scattering angles distributed between 20 and 160 deg. Inelastically-scattered neutron groups are observed corresponding to the excitation of levels at 909 +- 23, 1504 +- 20, 1747 +- 16, 2224 +- 16, 2567 +- 26, 2889 +- 12 and 3104 +- 10 keV. The experimental results are discussed in terms of the spherical optical-statistical, coupled-channels and core-coupling models and compared with corresponding quantities given in the evaluated nuclear data file ENDF/B-V

Modification of NUR II neutron beam profile of MINT TRIGA MARK II research reactor for digital neutron radiography

International Nuclear Information System (INIS)

Muhammad Rawi Mohamed Zin; Azali Muhammad; Abdul Aziz Mohamed; Rafhayudi Jamro; Syed Nasaruddin Syed Idris; Ng Aik Hao; Rosly Jaafar

2006-01-01

A cone neutron beam collimated by a 5.4 cm aperture produced in the Neutron Radiography II (NUR II) via a step divergence collimator had to be modified to fulfill 5 cm x 6 cm dimension of the scintillation screen placed in the charge couple device (ccd) camera. The required convergence neutron beam was obtained by a simple collimator-beam plug plugged in front of the NUR II beam port. The calculations involved in designing the collimator-beam plug had to take into account not only the neutron beam profiling but also the neutron and gamma shielding and are discussed in this article. (Author)

Performance of the coupled thermalhydraulics/neutron kinetics code R/P/C on workstation clusters and multiprocessor systems

International Nuclear Information System (INIS)

Hammer, C.; Paffrath, M.; Boeer, R.; Finnemann, H.; Jackson, C.J.

1996-01-01

The light water reactor core simulation code PANBOX has been coupled with the transient analysis code RELAP5 for the purpose of performing plant safety analyses with a three-dimensional (3-D) neutron kinetics model. The system has been parallelized to improve the computational efficiency. The paper describes the features of this system with emphasis on performance aspects. Performance results are given for different types of parallelization, i. e. for using an automatic parallelizing compiler, using the portable PVM platform on a workstation cluster, using PVM on a shared memory multiprocessor, and for using machine dependent interfaces. (author)

Online monitoring of fast neutron (DT/DD) at Purnima neutron generator

International Nuclear Information System (INIS)

Bishnoi, S.; Patel, T.; Shukla, M.; Adhikari, P.S.; Sinha, A.

2012-01-01

A neutron generator (NG) at Purnima Labs, BARC has been developed for DT accelerator driven zero power subcritical (ADSS) system. Subcritical core of ADSS will be coupled to the NG for benchmarking experiments. Kinetic parameters of ADSS such as K-source, flux, power etc depends on this external neutron source strength injected to the core. However the neutron emission rate of NG does not remain stable throughout its operation. In view of this a reliable, precise and online monitoring of NG's neutron emission rate is required. An online neutron monitoring system based on associated particle method has been designed, developed and installed at NG. The monitoring unit consists of an ion implanted planar silicon detector, placed inside the drift tube of NG at an angle with respect to D + beam direction. A series of experiments were carried out with increasing neutron yield to optimize the position of detector such that it has sufficient counting statistics and minimum pileup. A complementary calibration procedure for validating these results based on activation technique was also carried out with standard Cu foil. The reaction rate monitored with online monitor and foil activation technique were compared, their variations with the predicted (theoretical) results were within 16%. This paper deals with the development and performance of online neutron monitoring system for DT and DD neutrons

Burnup dependent core neutronic calculations for research and training reactors via SCALE4.4

International Nuclear Information System (INIS)

Tombakoglu, M.; Cecen, Y.

2001-01-01

In this work, the full core modelling is performed to improve neutronic analyses capability for nuclear research reactors using SCALE4.4 code system. KENOV.a module of SCALE4.4 code system is utilized for full core neutronic analysis. The ORIGEN-S module is coupled with the KENOV.a module to perform burnup dependent neutronic analyses. Results of neutronic calculations for 1 st cycle of Cekmece TR-2 research reactor are presented. In particular, coupling of KENOV.a and ORIGEN-S modules of SCALE4.4 is discussed. The preliminary results of 2-D burnup dependent neutronic calculations are also given. These results are extended to burnup dependent core calculations of TRIGA Mark-II research reactors. The code system developed here is similar to the code system that couples MCNP and ORIGEN2.(author)

Coupling of the computational fluid dynamics code ANSYS CFX with the 3D neutron kinetic core model DYN3D

International Nuclear Information System (INIS)

Kliem, S.; Grahn, A.; Rohde, U.; Schuetze, J.; Frank, Th.

2010-01-01

The computational fluid dynamics code ANSYS CFX has been coupled with the neutron-kinetic core model DYN3D. ANSYS CFX calculates the fluid dynamics and related transport phenomena in the reactors coolant and provides the corresponding data to DYN3D. In the fluid flow simulation of the coolant, the core itself is modeled within the porous body approach. DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the prototype that is currently available, the coupling is restricted to single-phase flow problems. In the time domain an explicit coupling of the codes has been implemented so far. Steady-state and transient verification calculations for two small-size test problems confirm the correctness of the implementation of the prototype coupling. The first test problem was a mini-core consisting of nine real-size fuel assemblies with quadratic cross section. Comparison was performed with the DYN3D stand-alone code. In the steady state, the effective multiplication factor obtained by the DYN3D/ANSYS CFX codes hows a deviation of 9.8 pcm from the DYN3D stand-alone solution. This difference can be attributed to the use of different water property packages in the two codes. The transient test case simulated the withdrawal of the control rod from the central fuel assembly at hot zero power in the same mini-core. Power increase during the introduction of positive reactivity and power reduction due to fuel temperature increase are calculated in the same manner by the coupled and the stand-alone codes. The maximum values reached during the power rise differ by about 1 MW at a power level of 50 MW. Beside the different water property packages, these differences are caused by the use of different flow solvers. The same calculations were carried for a mini-core with seven real-size fuel assemblies with hexagonal cross section in

Neutron electric dipole moment from supersymmetric anomalous W-boson coupling

International Nuclear Information System (INIS)

Kadoyoshi, T.; Oshimo, N.

1997-01-01

In the supersymmetric standard model (SSM) the W boson could have a nonvanishing electric dipole moment (EDM) through a one-loop diagram mediated by the charginos and neutralinos. Then the W-boson EDM induces the EDMs of the neutron and the electron. We discuss these EDMs, taking into consideration the constraints from the neutron and electron EDMs at the one-loop level induced by the charginos and squarks or sleptons. It is shown that the neutron and the electron could, respectively, have EDMs of the order of 10 -26 ecm and 10 -27 ecm, solely owing to the W-boson EDM. Since these EDMs do not depend on the values of the SSM parameters for the squark or slepton sector, they provide less ambiguous predictions for CP violation in the SSM. copyright 1997 The American Physical Society

Initial Coupling of the RELAP-7 and PRONGHORN Applications

Energy Technology Data Exchange (ETDEWEB)

J. Ortensi; D. Andrs; A.A. Bingham; R.C. Martineau; J.W. Peterson

2012-10-01

Modern nuclear reactor safety codes require the ability to solve detailed coupled neutronic- thermal fluids problems. For larger cores, this implies fully coupled higher dimensionality spatial dynamics with appropriate feedback models that can provide enough resolution to accurately compute core heat generation and removal during steady and unsteady conditions. The reactor analysis code PRONGHORN is being coupled to RELAP-7 as a first step to extend RELAP’s current capabilities. This report details the mathematical models, the type of coupling, and the testing results from the integrated system. RELAP-7 is a MOOSE-based application that solves the continuity, momentum, and energy equations in 1-D for a compressible fluid. The pipe and joint capabilities enable it to model parts of the power conversion unit. The PRONGHORN application, also developed on the MOOSE infrastructure, solves the coupled equations that define the neutron diffusion, fluid flow, and heat transfer in a full core model. The two systems are loosely coupled to simplify the transition towards a more complex infrastructure. The integration is tested on a simplified version of the OECD/NEA MHTGR-350 Coupled Neutronics-Thermal Fluids benchmark model.

Recent advances in laser-driven neutron sources

Science.gov (United States)

Alejo, A.; Ahmed, H.; Green, A.; Mirfayzi, S. R.; Borghesi, M.; Kar, S.

2016-11-01

Due to the limited number and high cost of large-scale neutron facilities, there has been a growing interest in compact accelerator-driven sources. In this context, several potential schemes of laser-driven neutron sources are being intensively studied employing laser-accelerated electron and ion beams. In addition to the potential of delivering neutron beams with high brilliance, directionality and ultra-short burst duration, a laser-driven neutron source would offer further advantages in terms of cost-effectiveness, compactness and radiation confinement by closed-coupled experiments. Some of the recent advances in this field are discussed, showing improvements in the directionality and flux of the laser-driven neutron beams.

Device for Writing the Time Tail from Spallation Neutron Pulses

International Nuclear Information System (INIS)

Langan, P.; Schoenborn, Benno P.; Daemen, L.L.

2001-01-01

Recent work at Los Alamos Neutron Science Center (LANSCE), has shown that there are large gains in neutron beam intensity to be made by using coupled moderators at spallation neutron sources. Most of these gains result from broadening the pulse-width in time. However the accompanying longer exponential tail at large emission times can be a problem in that it introduces relatively large beam-related backgrounds at high resolutions. We have designed a device that can reshape the moderated neutron beam by cutting the time-tail so that a sharp time resolution can be re-established without a significant loss in intensity. In this work the basic principles behind the tail-cutter and some initial results of Monte Carlo simulations are described. Unwanted neutrons in the long time-tail are diffracted out of the transmitted neutron beam by a nested stack of aperiodic multi-layers, rocking at the same frequency as the source. Nested aperiodic multi-layers have recently been used at X-ray sources and as band-pass filters in quasi-Laue neutron experiments at reactor neutron sources. Optical devices that rock in synchronization with a pulsed neutron beam are relatively new but are already under construction at LANSCE. The tail-cutter described here is a novel concept that uses existing multi-layer technology in a new way for spallation neutrons. Coupled moderators in combination with beam shaping devices offer the means of increasing flux whilst maintaining a sharp time distribution. A prototype device is being constructed for the protein crystallography station at LANSCE. The protein crystallography station incorporates a water moderator that has been judiciously coupled in order to increase the flux over neutron energies that are important to structural biology (3-80meV). This development in moderator design is particularly important because protein crystallography is flux limited and because conventional ambient water and cold hydrogen moderators do not provide relatively

Analysis of the ITER computational shielding benchmark with the Monte Carlo TRIPOLI-4{sup ®} neutron gamma coupled calculations

Energy Technology Data Exchange (ETDEWEB)

Lee, Yi-Kang, E-mail: yi-kang.lee@cea.fr

2016-11-01

Highlights: • Verification and validation of TRIPOLI-4 radiation transport calculations for ITER shielding benchmark. • Evaluation of CEA-V5.1.1 and FENDL-3.0 nuclear data libraries on D–T fusion neutron continuous energy transport calculations. • Advances in nuclear analyses for nuclear heating and radiation damage in iron. • This work also demonstrates that the “safety factors” concept is necessary in the nuclear analyses of ITER. - Abstract: With the growing interest in using the continuous-energy TRIPOLI-4{sup ®} Monte Carlo radiation transport code for ITER applications, a key issue that arises is whether or not the released TRIPOLI-4 code and its associated nuclear data libraries are verified and validated for the D–T fusion neutronics calculations. Previous published benchmark results of TRIPOLI-4 code on the ITER related activities have concentrated on the first wall loading, the reactor dosimetry, the nuclear heating, and the tritium breeding ratio. To enhance the TRIPOLI-4 verification and validation on neutron-gamma coupled calculations for fusion device application, the computational ITER shielding benchmark of M. E. Sawan was performed in this work by using the 2013 released TRIPOLI-4.9S code and the associated CEA-V5.1.1 data library. First wall, blanket, vacuum vessel and toroidal field magnet of the inboard and outboard components were fully modelled in this 1-D toroidal cylindrical benchmark. The 14.1 MeV source neutrons were sampled from a uniform isotropic distribution in the plasma zone. Nuclear responses including neutron and gamma fluxes, nuclear heating, and material damage indicator were benchmarked against previous published results. The capabilities of the TRIPOLI-4 code on the evaluation of above physics parameters were presented. The nuclear data library from the new FENDL-3.0 evaluation was also benchmarked against the CEA-V5.1.1 results for the neutron transport calculations. The results show that both data libraries

IMPROVED COMPUTATIONAL CHARACTERIZATION OF THE THERMAL NEUTRON SOURCE FOR NEUTRON CAPTURE THERAPY RESEARCH AT THE UNIVERSITY OF MISSOURI

Energy Technology Data Exchange (ETDEWEB)

Stuart R. Slattery; David W. Nigg; John D. Brockman; M. Frederick Hawthorne

2010-05-01

Parameter studies, design calculations and initial neutronic performance measurements have been completed for a new thermal neutron beamline to be used for neutron capture therapy cell and small-animal radiobiology studies at the University of Missouri Research Reactor. The beamline features the use of single-crystal silicon and bismuth sections for neutron filtering and for reduction of incident gamma radiation. The computational models used for the final beam design and performance evaluation are based on coupled discrete-ordinates and Monte Carlo techniques that permit detailed modeling of the neutron transmission properties of the filtering crystals with very few approximations. This is essential for detailed dosimetric studies required for the anticipated research program.

Preliminary optimization experiments of coupled liquid hydrogen moderator for KENS-II

International Nuclear Information System (INIS)

Watanabe, N.; Kiyanagi, Y.; Inoue, K.; Furusaka, M.; Ikeda, S.; Arai, M.; Iwasa, H.

1989-01-01

As a preliminary optimization experiment on the cold-neutron source for KENS-II, energy and time distributions of cold neutrons emanating from coupled liquid-hydrogen moderators with and without a premoderator in a graphite reflector were measured and compared with those from a decoupled liquid-hydrogen moderator. The results showed that the energy spectra from the coupled liquid-hydrogen moderators are almost the same as those from a decoupled one. Relative gain of the former to the latter is fairly high, more than 5, and further increases with increasing wavelength. The broadening of the neutron pulse width in coupled moderators at the cold-neutron region is not so significant and only 1.5 times compared to the solid methane moderator presently operated at KENS-II. 2 refs., 12 figs., 1 tab

Status report on SHARP coupling framework.

Energy Technology Data Exchange (ETDEWEB)

Caceres, A.; Tautges, T. J.; Lottes, J.; Fischer, P.; Rabiti, C.; Smith, M. A.; Siegel, A.; Yang, W. S.; Palmiotti, G.

2008-05-30

This report presents the software engineering effort under way at ANL towards a comprehensive integrated computational framework (SHARP) for high fidelity simulations of sodium cooled fast reactors. The primary objective of this framework is to provide accurate and flexible analysis tools to nuclear reactor designers by simulating multiphysics phenomena happening in complex reactor geometries. Ideally, the coupling among different physics modules (such as neutronics, thermal-hydraulics, and structural mechanics) needs to be tight to preserve the accuracy achieved in each module. However, fast reactor cores in steady state mode represent a special case where weak coupling between neutronics and thermal-hydraulics is usually adequate. Our framework design allows for both options. Another requirement for SHARP framework has been to implement various coupling algorithms that are parallel and scalable to large scale since nuclear reactor core simulations are among the most memory and computationally intensive, requiring the use of leadership-class petascale platforms. This report details our progress toward achieving these goals. Specifically, we demonstrate coupling independently developed parallel codes in a manner that does not compromise performance or portability, while minimizing the impact on individual developers. This year, our focus has been on developing a lightweight and loosely coupled framework targeted at UNIC (our neutronics code) and Nek (our thermal hydraulics code). However, the framework design is not limited to just using these two codes.

ZZ-PBMR-400, OECD/NEA PBMR Coupled Neutronics/Thermal Hydraulics Transient Benchmark - The PBMR-400 Core Design

International Nuclear Information System (INIS)

Reitsma, Frederik

2007-01-01

Description of benchmark: This international benchmark, concerns Pebble-Bed Modular Reactor (PBMR) coupled neutronics/thermal hydraulics transients based on the PBMR-400 MW design. The deterministic neutronics, thermal-hydraulics and transient analysis tools and methods available to design and analyse PBMRs lag, in many cases, behind the state of the art compared to other reactor technologies. This has motivated the testing of existing methods for HTGRs but also the development of more accurate and efficient tools to analyse the neutronics and thermal-hydraulic behaviour for the design and safety evaluations of the PBMR. In addition to the development of new methods, this includes defining appropriate benchmarks to verify and validate the new methods in computer codes. The scope of the benchmark is to establish well-defined problems, based on a common given set of cross sections, to compare methods and tools in core simulation and thermal hydraulics analysis with a specific focus on transient events through a set of multi-dimensional computational test problems. The benchmark exercise has the following objectives: - Establish a standard benchmark for coupled codes (neutronics/thermal-hydraulics) for PBMR design; - Code-to-code comparison using a common cross section library ; - Obtain a detailed understanding of the events and the processes; - Benefit from different approaches, understanding limitations and approximations. Major Design and Operating Characteristics of the PBMR (PBMR Characteristic and Value): Installed thermal capacity: 400 MW(t); Installed electric capacity: 165 MW(e); Load following capability: 100-40-100%; Availability: ≥ 95%; Core configuration: Vertical with fixed centre graphite reflector; Fuel: TRISO ceramic coated U-235 in graphite spheres; Primary coolant: Helium; Primary coolant pressure: 9 MPa; Moderator: Graphite; Core outlet temperature: 900 C.; Core inlet temperature: 500 C.; Cycle type: Direct; Number of circuits: 1; Cycle

Influence of an SN solver in a fine-mesh neutronics/thermal-hydraulics framework

International Nuclear Information System (INIS)

Jareteg, Klas; Vinai, Paolo; Demaziere, Christophe; Sasic, Srdjan

2015-01-01

In this paper a study on the influence of a neutron discrete ordinates (S N ) solver within a fine-mesh neutronic/thermal-hydraulic methodology is presented. The methodology consists of coupling a neutronic solver with a single-phase fluid solver, and it is aimed at computing the two fields on a three-dimensional (3D) sub-pin level. The cross-sections needed for the neutron transport equations are pre-generated using a Monte Carlo approach. The coupling is resolved in an iterative manner with full convergence of both fields. A conservative transfer of the full 3D information is achieved, allowing for a proper coupling between the neutronic and the thermal-hydraulic meshes on the finest calculated scales. The discrete ordinates solver is benchmarked against a Monte Carlo reference solution for a two-dimensional (2D) system. The results confirm the need of a high number of ordinates, giving a satisfactory accuracy in k eff and scalar flux profile applying S 16 for 16 energy groups. The coupled framework is used to compare the S N implementation and a solver based on the neutron diffusion approximation for a full 3D system of a quarter of a symmetric, 7x7 array in an infinite lattice setup. In this case, the impact of the discrete ordinates solver shows to be significant for the coupled system, as demonstrated in the calculations of the temperature distributions. (author)

Neutron Scattering Investigations of Correlated Electron Systems and Neutron Instrumentation

DEFF Research Database (Denmark)

Holm, Sonja Lindahl

are a unique probe for studying the atomic and molecular structure and dynamics of materials. Even though neutrons are very expensive to produce, the advantages neutrons provide overshadow the price. As neutrons interact weakly with materials compared to many other probes, e.g. electrons or photons...... contains antiferromagnetically coupled Cu2+ S = 1=2 ions forming truncated 24-spin cube clusters of linked triangles. The clusters in boleite afford a situation intermediate between molecular and bulk magnetism, accessible to both experiment and numerical theory, in which a spin liquid can be studied...... the impact of the time structure (pulse length and repetition frequency) choice for ESS are appended. McStas simulations of a low resolution cold powder diffractometer and high resolution thermal powder diffractometer with wavelength frame multiplication have been carried out for 20 different settings...

Neutron skin of 208 Pb in consistency with neutron star observations

CERN Document Server

Miyazaki, K

2007-01-01

The renormalized meson-nucleon couplings are applied to the relativistic optical model of p-208Pb elastic scattering at T_{lab}=200MeV. We calculate the strength of the vector potential at nuclear center as varying the neutron radius of 208Pb. The neutron skin thickness S_{n} is determined in the comparison of the calculated potential with the phenomenological one. We find a value S_{n}=0.118fm being consistent with the astronomical observations of massive neutron stars (NSs), the standard scenario of NS cooling and the experimental nuclear symmetry energy in terrestrial laboratory. The value is complementary to the previous result S_{n}=0.119fm in the analysis of elastic scattering above T_{lab}=500MeV within the relativistic impulse optical model.

A test-type hyper-thermal neutron generator for neutron capture therapy - estimation of neutron energy spectrum by simulation calculations and TOF experiments

International Nuclear Information System (INIS)

Sakurai, Yoshinori; Kobayashi, Tooru; Kobayashi, Katsuhei

1999-01-01

In order to clarify the irradiation characteristics of hyper-thermal neutrons and the feasibility of a hyper-thermal neutron irradiation field for neutron capture therapy, a 'test-type' hyper-thermal neutron generator was designed and made. Graphite of 6 cm thickness and 21 cm diameter was selected as the high temperature scatterer. The scatterer is heated up to 1200 deg. C maximum using molybdenum heaters. The radiation heat is shielded by reflectors of molybdenum and stainless steel. The temperature is measured using three R-type thermo-couples and controlled by a program controller. The total thickness of the generator is designed to be as thin as possible, 20 cm in maximum, in the standing point of the neutron beam intensity. The thermal stability, controllability and safety of the generator at high temperature employment were confirmed by the heating tests. As one of the experiments for the characteristics estimation, the neutron energy spectrum dependent on the scatterer temperature was measured by the TOF (time of flight) method using the LINAC neutron generator. The estimations by simulation calculations were also performed. From the experiment and calculation results, it was confirmed that the neutron temperature shifted higher as the scatterer temperature was higher. The prospect of the feasibility of the 'hyper-thermal neutron irradiation field for NCT' was opened from the estimation results of the generator characteristics by the simulation calculations and experiments

Coupling of the 3D neutron kinetic core model DYN3D with the CFD software ANSYS-CFX

International Nuclear Information System (INIS)

Grahn, Alexander; Kliem, Sören; Rohde, Ulrich

2015-01-01

Highlights: • Improved thermal hydraulic description of nuclear reactor cores. • Possibility of three-dimensional flow phenomena in the core, such as cross flow, flow reversal, flow around obstacles. • Simulation at higher spatial resolution as compared to system codes. - Abstract: This article presents the implementation of a coupling between the 3D neutron kinetic core model DYN3D and the commercial, general purpose computational fluid dynamics (CFD) software ANSYS-CFX. In the coupling approach, parts of the thermal hydraulic calculation are transferred to CFX for its better ability to simulate the three-dimensional coolant redistribution in the reactor core region. The calculation of the heat transfer from the fuel into the coolant remains with DYN3D, which incorporates well tested and validated heat transfer models for rod-type fuel elements. On the CFX side, the core region is modeled based on the porous body approach. The implementation of the code coupling is verified by comparing test case results with reference solutions of the DYN3D standalone version. Test cases cover mini and full core geometries, control rod movement and partial overcooling transients

Particle physics with cold neutrons

International Nuclear Information System (INIS)

Dubbers, D.

1991-01-01

Slow neutrons are used in a large number of experiments to study the physics of particles and their fundamental interactions. Some of these experiments search for manifestations of ''new physics'' like baryon- or lepton-number nonconservation, time reversal nonconservation, new particles, right-handed currents, nonzero neutron charge, nonlinear terms in the Schrodinger equation, exotic e + e - states, and others. Other slow neutron experiments test the present Standard Model. The parity nonconserving weak neutron-nucleon interaction is studied in a variety of experiments. Free neutron beta decay gives precise values for the weak vector and axialvector coupling constants, which allow precise tests of basic symmetries like the conservation of the weak vector current, the unitarity of the weak quark mixing matrix, SU(3) flavour symmetry, and right-handed currents. Neutron beta decay data are further needed to calculate weak cross-sections, for applications, in big bang cosmology, in astrophysics, in solar physics and the solar neutrino problem, and in such mundane things as neutrino detection efficiencies in neutrino oscillation or proton decay experiments. Neutron-nucleon, neutron-nucleus and neutron-electron scattering lengths are determined in high precision experiments, which use methods like neutron interferometry or neutron gravity spectrometry. The experiments give information on quantities like the neutron charge radius or the neutron electric polarizability. Precision measurements of other fundamental constants lead to a better, model-independent value of the fine structure constant. Finally, the fundamental experiments on quantum mechanics, like spinor 4π -rotation, Berry's phase, dressed neutrons, Aharanov - Casher effect, or gravitational effects on the neutron's phase will be briefly discussed. (author)

Transient thermal-hydraulic/neutronic analysis in a VVER-1000 reactor core

International Nuclear Information System (INIS)

Seyed khalil Mousavian; Mohammad Mohsen Ertejaei; Majid Shahabfar

2005-01-01

Full text of publication follows: Nowadays, coupled thermal-hydraulic and three-dimensional neutronic codes in order to consider different feedback effects is state of the art subject in nuclear engineering researches. In this study, RELAP5/COBRA and WIMS/CITATION codes are implemented to investigate the VVER-1000 reactor core parameters during Large Break Loss of Coolant Accident (LB-LOCA). In a LB-LOCA, the primary side pressure, coolant density and fuel temperature strongly decrease but the cladding temperature experiences a strong peak. For this purpose, the RELAP5 Best Estimate (BE) system code is used to simulate the LB-LOCA analysis in VVER-1000 nuclear thermal-hydraulic loops. Also, the modified COBRA-IIIc software as a sub-channel analysis code is applied for modeling of VVER-1000 reactor core. Moreover, WIMS and CITATION as a cross section and 3-D neutron flux codes are coupled with thermal-hydraulic codes with the aim of consider the spatial effects through the reactor core. For this reason, suitable software is developed to link and speed up the coupled thermalhydraulic and three-dimensional neutronic calculations. This software utilizes of external coupling concept in order to integrate thermal-hydraulic and neutronic calculations. (authors)

7Li neutron-induced elastic scattering cross section measurement using a slowing-down spectrometer

Directory of Open Access Journals (Sweden)

Heusch M.

2010-10-01

Full Text Available A new integral measurement of the 7Li neutron induced elastic scattering cross section was determined in a wide neutron energy range. The measurement was performed on the LPSC-PEREN experimental facility using a heterogeneous graphite-LiF slowing-down time spectrometer coupled with an intense pulsed neutron generator (GENEPI-2. This method allows the measurement of the integral elastic scattering cross section in a slowing-down neutron spectrum. A Bayesian approach coupled to Monte Carlo calculations was applied to extract naturalC, 19F and 7Li elastic scattering cross sections.

Testing Universal Relations of Neutron Stars with a Nonlinear Matter-Gravity Coupling Theory

Science.gov (United States)

Sham, Y.-H.; Lin, L.-M.; Leung, P. T.

2014-02-01

Due to our ignorance of the equation of state (EOS) beyond nuclear density, there is still no unique theoretical model for neutron stars (NSs). It is therefore surprising that universal EOS-independent relations connecting different physical quantities of NSs can exist. Lau et al. found that the frequency of the f-mode oscillation, the mass, and the moment of inertia are connected by universal relations. More recently, Yagi and Yunes discovered the I-Love-Q universal relations among the mass, the moment of inertia, the Love number, and the quadrupole moment. In this paper, we study these universal relations in the Eddington-inspired Born-Infeld (EiBI) gravity. This theory differs from general relativity (GR) significantly only at high densities due to the nonlinear coupling between matter and gravity. It thus provides us an ideal case to test how robust the universal relations of NSs are with respect to the change of the gravity theory. Due to the apparent EOS formulation of EiBI gravity developed recently by Delsate and Steinhoff, we are able to study the universal relations in EiBI gravity using the same techniques as those in GR. We find that the universal relations in EiBI gravity are essentially the same as those in GR. Our work shows that, within the currently viable coupling constant, there exists at least one modified gravity theory that is indistinguishable from GR in view of the unexpected universal relations.

Testing universal relations of neutron stars with a nonlinear matter-gravity coupling theory

International Nuclear Information System (INIS)

Sham, Y.-H.; Lin, L.-M.; Leung, P. T.

2014-01-01

Due to our ignorance of the equation of state (EOS) beyond nuclear density, there is still no unique theoretical model for neutron stars (NSs). It is therefore surprising that universal EOS-independent relations connecting different physical quantities of NSs can exist. Lau et al. found that the frequency of the f-mode oscillation, the mass, and the moment of inertia are connected by universal relations. More recently, Yagi and Yunes discovered the I-Love-Q universal relations among the mass, the moment of inertia, the Love number, and the quadrupole moment. In this paper, we study these universal relations in the Eddington-inspired Born-Infeld (EiBI) gravity. This theory differs from general relativity (GR) significantly only at high densities due to the nonlinear coupling between matter and gravity. It thus provides us an ideal case to test how robust the universal relations of NSs are with respect to the change of the gravity theory. Due to the apparent EOS formulation of EiBI gravity developed recently by Delsate and Steinhoff, we are able to study the universal relations in EiBI gravity using the same techniques as those in GR. We find that the universal relations in EiBI gravity are essentially the same as those in GR. Our work shows that, within the currently viable coupling constant, there exists at least one modified gravity theory that is indistinguishable from GR in view of the unexpected universal relations.

Application of coupled codes for safety analysis and licensing issues

International Nuclear Information System (INIS)

Langenbuch, S.; Velkov, K.

2006-01-01

An overview is given on the development and the advantages of coupled codes which integrate 3D neutron kinetics into thermal-hydraulic system codes. The work performed within GRS by coupling the thermal-hydraulic system code ATHLET and the 3D neutronics code QUABOX/CUBBOX is described as an example. The application of the coupled codes as best-estimate simulation tools for safety analysis is discussed. Some examples from German licensing practices are given which demonstrate how the improved analytical methods of coupled codes have contributed to solve licensing issues related to optimized and more economical use of fuel. (authors)

Constraints on mirror models of dark matter from observable neutron-mirror neutron oscillation

Science.gov (United States)

Mohapatra, Rabindra N.; Nussinov, Shmuel

2018-01-01

The process of neutron-mirror neutron oscillation, motivated by symmetric mirror dark matter models, is governed by two parameters: n -n‧ mixing parameter δ and n -n‧ mass splitting Δ. For neutron mirror neutron oscillation to be observable, the splitting between their masses Δ must be small and current experiments lead to δ ≤ 2 ×10-27 GeV and Δ ≤10-24 GeV. We show that in mirror universe models where this process is observable, this small mass splitting constrains the way that one must implement asymmetric inflation to satisfy the limits of Big Bang Nucleosynthesis on the number of effective light degrees of freedom. In particular we find that if asymmetric inflation is implemented by inflaton decay to color or electroweak charged particles, the oscillation is unobservable. Also if one uses SM singlet fields for this purpose, they must be weakly coupled to the SM fields.

Excitations of one-valence-proton, one-valence-neutron nucleus {sup 210}Bi from cold-neutron capture

Energy Technology Data Exchange (ETDEWEB)

Cieplicka-Oryńczak, N. [INFN sezione di Milano, Via Celoria 16, 20133 Milano (Italy); Institute of Nuclear Physics, Polish Academy of Sciences, PL-31342 Kraków (Poland); Fornal, B.; Szpak, B. [Institute of Nuclear Physics, Polish Academy of Sciences, PL-31342 Kraków (Poland); Leoni, S.; Bottoni, S. [INFN sezione di Milano, Via Celoria 16, 20133 Milano (Italy); Università degli Studi di Milano, Via Celoria 16, 20133 Milano (Italy); Bazzacco, D. [Dipartimento di Fisica e Astronomia dell’Università, I-35131 Padova (Italy); INFN Sezione di Padova, I-35131 Padova (Italy); Blanc, A.; Jentschel, M.; Köster, U.; Mutti, P.; Soldner, T. [Institute Laue-Langevin, 6, rue Jules Horowitz, 38042 Grenoble Cedex 9 (France); Bocchi, G. [Università degli Studi di Milano, Via Celoria 16, 20133 Milano (Italy); France, G. de [GANIL, Bd. Becquerel, BP 55027, 14076 CAEN Cedex 05 (France); Simpson, G. [LPSC, Université Joseph Fourier Grenoble 1, CNRS/IN2P3, Institut National Polytechnique de Grenoble, F-38026 Grenoble Cedex (France); Ur, C. [INFN Sezione di Padova, Via F. Marzolo 8, I-35131 Padova (Italy); Urban, W. [Faculty of Physics, University of Warsaw, ul. Hoża 69, 02-681, Warszawa (Poland)

2015-10-15

The low-spin structure of one-proton, one-neutron {sup 210}Bi nucleus was investigated in cold-neutron capture reaction on {sup 209}Bi. The γ-coincidence measurements were performed with use of EXILL array consisted of 16 HPGe detectors. The experimental results were compared to shell-model calculations involving valence particles excitations. The {sup 210}Bi nucleus offers the potential to test the effective proton-neutron interactions because most of the states should arise from the proton-neutron excitations. Additionally, it was discovered that a few states should come from the couplings of valence particles to the 3{sup −} octupole vibration in {sup 208}Pb which provides also the possibility of testing the calculations involving the core excitations.

Neutron stars in relativistic mean field theory with isovector scalar meson

International Nuclear Information System (INIS)

Kubis, S.; Kutschera, M.; Stachniewicz, S.

1998-01-01

We study the equation of state (EOS) of β-stable dense matter and models of neutron stars in the relativistic mean field (RMF) theory with the isovector scalar mean field corresponding to the δ-meson (a 0 (980)). A range of values of the δ-meson coupling compatible with the Bonn potentials is explored. Parameters of the model in the isovector sector are constrained to fit the nuclear symmetry energy, E s ∼30 MeV. We find that the quantity most sensitive to the δ-meson coupling is the proton fraction of neutron star matter. It increases significantly in the presence of the δ-field. The energy per baryon also increases but the effect is smaller. The EOS becomes slightly stiffer and the maximum neutron star mass increases for stronger δ-meson coupling. (author)

PANDORA, a large volume low-energy neutron detector with real-time neutron-gamma discrimination

Science.gov (United States)

Stuhl, L.; Sasano, M.; Yako, K.; Yasuda, J.; Baba, H.; Ota, S.; Uesaka, T.

2017-09-01

The PANDORA (Particle Analyzer Neutron Detector Of Real-time Acquisition) system, which was developed for use in inverse kinematics experiments with unstable isotope beams, is a neutron detector based on a plastic scintillator coupled to a digital readout. PANDORA can be used for any reaction study involving the emission of low energy neutrons (100 keV-10 MeV) where background suppression and an increased signal-to-noise ratio are crucial. The digital readout system provides an opportunity for pulse shape discrimination (PSD) of the detected particles as well as intelligent triggering based on PSD. The figure of merit results of PANDORA are compared to the data in literature. Using PANDORA, 91 ± 1% of all detected neutrons can be separated, while 91 ± 1% of the detected gamma rays can be excluded, reducing the gamma ray background by one order of magnitude.

Coupled variations of fundamental couplings and primordial nucleosynthesis

International Nuclear Information System (INIS)

Coc, Alain; Nunes, Nelson J.; Olive, Keith A.; Uzan, Jean-Philippe; Vangioni, Elisabeth

2006-10-01

The effect of variations of the fundamental nuclear parameters on big-bang nucleosynthesis are modeled and discussed in detail taking into account the interrelations between the fundamental parameters arising in unified theories. Considering only 4 He, strong constraints on the variation of the neutron lifetime, neutron-proton mass difference are set. These constraints are then translated into constraints on the time variation of the Yukawa couplings and the fine structure constant. Furthermore, we show that a variation of the deuterium binding energy is able to reconcile the 7 Li abundance deduced from the WMAP analysis with its spectroscopically determined value while maintaining concordance with D and 4 He. (authors)

High-speed motion neutron radiography

International Nuclear Information System (INIS)

Bossi, R.H.; Barton, J.P.; Robinson, A.H.

1982-01-01

A system has been developed to perform neutron radiographic analysis of dynamic events having a duration of several milliseconds. The system has been operated in the range of 2000 to 10,000 frames. Synchronization has provided high-speed motion neutron radiographs for evaluation of the firing cycles of 7.62-mm munition rounds within a thick steel rifle barrel. The system has also been used to demonstrate its ability to produce neutron radiographic movies of two-phase flow. The equipment includes a TRIGA reactor capable of pulsing to a peak power of 3000 MW, a neutron beam collimator, a scintillator neutron conversion screen coupled to an image intensifier, and a 16-mm high-speed movie camera. The peak neutron flux incident at the object position is about 4 X 10 11 n/cm 2 X s with a pulse, full-width at half-maximum, of 9 ms. Modulation transfer function techniques have been used to assist optimization of the system performance. Special studies have been performed on the scintillator conversion screens and on the effects of statistical limitations on information availability

A coupled diffusion-transport computational method and its application for the determination of space dependent angular flux distributions at a cold neutron source

International Nuclear Information System (INIS)

Turgut, M.H.

1985-01-01

A fast calculation program ''BRIDGE'' was developed for the calculation of a Cold Neutron Source (CNS) at a radial beam tube of the FRG-I reactor, which couples a total assembly diffusion calculation to a transport calculation for a certain subregion. For the coupling flux and current boundary values at the common surfaces are taken from the diffusion calculation and are used as driving conditions in the transport calculation. 'Equivalence Theorie' is used for the transport feedback effect on the diffusion calculation to improve the consistency of the boundary values. The optimization of a CNS for maximizing the subthermal flux in the wavelength range 4 - 6 A is discussed. (orig.) [de

Neutron-induced photon production in MCNP

International Nuclear Information System (INIS)

Little, R.C.; Seamon, R.E.

1983-01-01

An improved method of neutron-induced photon production has been incorporated into the Monte Carlo transport code MCNP. The new method makes use of all partial photon-production reaction data provided by ENDF/B evaluators including photon-production cross sections as well as energy and angular distributions of secondary photons. This faithful utilization of sophisticated ENDF/B evaluations allows more precise MCNP calculations for several classes of coupled neutron-photon problems

Neutron diffraction measurements at the INES diffractometer using a neutron radiative capture based counting technique

Energy Technology Data Exchange (ETDEWEB)

Festa, G. [Centro NAST, Universita degli Studi di Roma Tor Vergata, Roma (Italy); Pietropaolo, A., E-mail: antonino.pietropaolo@roma2.infn.it [Centro NAST, Universita degli Studi di Roma Tor Vergata, Roma (Italy); Grazzi, F.; Barzagli, E. [CNR-ISC Firenze (Italy); Scherillo, A. [CNR-ISC Firenze (Italy); ISIS facility Rutherford Appleton Laboratory (United Kingdom); Schooneveld, E.M. [ISIS facility Rutherford Appleton Laboratory (United Kingdom)

2011-10-21

The global shortage of {sup 3}He gas is an issue to be addressed in neutron detection. In the context of the research and development activity related to the replacement of {sup 3}He for neutron counting systems, neutron diffraction measurements performed on the INES beam line at the ISIS pulsed spallation neutron source are presented. For these measurements two different neutron counting devices have been used: a 20 bar pressure squashed {sup 3}He tube and a Yttrium-Aluminum-Perovskite scintillation detector. The scintillation detector was coupled to a cadmium sheet that registers the prompt radiative capture gamma rays generated by the (n,{gamma}) nuclear reactions occurring in cadmium. The assessment of the scintillator based counting system was done by performing a Rietveld refinement analysis on the diffraction pattern from an ancient Japanese blade and comparing the results with those obtained by a {sup 3}He tube placed at the same angular position. The results obtained demonstrate the considerable potential of the proposed counting approach based on the radiative capture gamma rays at spallation neutron sources.

Dual-fission chamber and neutron beam characterization for fission product yield measurements using monoenergetic neutrons

Science.gov (United States)

Bhatia, C.; Fallin, B.; Gooden, M. E.; Howell, C. R.; Kelley, J. H.; Tornow, W.; Arnold, C. W.; Bond, E. M.; Bredeweg, T. A.; Fowler, M. M.; Moody, W. A.; Rundberg, R. S.; Rusev, G.; Vieira, D. J.; Wilhelmy, J. B.; Becker, J. A.; Macri, R.; Ryan, C.; Sheets, S. A.; Stoyer, M. A.; Tonchev, A. P.

2014-09-01

A program has been initiated to measure the energy dependence of selected high-yield fission products used in the analysis of nuclear test data. We present out initial work of neutron activation using a dual-fission chamber with quasi-monoenergetic neutrons and gamma-counting method. Quasi-monoenergetic neutrons of energies from 0.5 to 15 MeV using the TUNL 10 MV FM tandem to provide high-precision and self-consistent measurements of fission product yields (FPY). The final FPY results will be coupled with theoretical analysis to provide a more fundamental understanding of the fission process. To accomplish this goal, we have developed and tested a set of dual-fission ionization chambers to provide an accurate determination of the number of fissions occurring in a thick target located in the middle plane of the chamber assembly. Details of the fission chamber and its performance are presented along with neutron beam production and characterization. Also presented are studies on the background issues associated with room-return and off-energy neutron production. We show that the off-energy neutron contribution can be significant, but correctable, while room-return neutron background levels contribute less than <1% to the fission signal.

Detection of fast neutrons in a plastic scintillator using digital pulse processing to reject gammas

International Nuclear Information System (INIS)

Reeder, P.L.; Peurrung, A.J.; Hansen, R.R.; Stromswold, D.C.; Hensley, W.K.; Hubbard, C.W.

1999-01-01

We report on neutron-gamma discrimination in a plastic scintillator based on the time delay inherent in second and third chance neutron scattering. Because of the time delay (∼3 ns) between the first and second scattering of a neutron, calculations of gammas and neutrons in a plastic scintillator predict that a neutron signal should be significantly broader than a pulse from a gamma event. Experimentally, we have used a fast digital oscilloscope coupled to a computer to examine individual pulses from neutron or gamma induced signals in fast scintillators coupled to a fast PMT. Individual neutron-induced signals were consistent with the predictions of our model, but gamma pulses were broader than expected. We present various tests to understand this phenomenon and discuss a way to overcome this problem

Approximate solution for the reactor neutron probability distribution

International Nuclear Information System (INIS)

Ruby, L.; McSwine, T.L.

1985-01-01

Several authors have studied the Kolmogorov equation for a fission-driven chain-reacting system, written in terms of the generating function G(x,y,z,t) where x, y, and z are dummy variables referring to the neutron, delayed neutron precursor, and detector-count populations, n, m, and c, respectively. Pal and Zolotukhin and Mogil'ner have shown that if delayed neutrons are neglected, the solution is approximately negative binomial for the neutron population. Wang and Ruby have shown that if the detector effect is neglected, the solution, including the effect of delayed neutrons, is approximately negative binomial. All of the authors assumed prompt-neutron emission not exceeding two neutrons per fission. An approximate method of separating the detector effect from the statistics of the neutron and precursor populations has been proposed by Ruby. In this weak-coupling limit, it is assumed that G(x,y,z,t) = H(x,y)I(z,t). Substitution of this assumption into the Kolmogorov equation separates the latter into two equations, one for H(x,y) and the other for I(z,t). Solution of the latter then gives a generating function, which indicates that in the weak-coupling limit, the detector counts are Poisson distributed. Ruby also showed that if the detector effect is neglected in the equation for H(x,y), i.e., the detector efficiency is set to zero, then the resulting equation is identical with that considered by Wang and Ruby. The authors present here an approximate solution for H(x,y) that does not set the detector efficiency to zero

Nuclear spin content and constraints on exotic spin-dependent couplings

International Nuclear Information System (INIS)

Kimball, D F Jackson

2015-01-01

There are numerous recent and ongoing experiments employing a variety of atomic species to search for couplings of atomic spins to exotic fields. In order to meaningfully compare these experimental results, the coupling of the exotic field to the atomic spin must be interpreted in terms of the coupling to electron, proton, and neutron spins. Traditionally, constraints from atomic experiments on exotic couplings to neutron and proton spins have been derived using the single-particle Schmidt model for nuclear spin. In this model, particular atomic species are sensitive to either neutron or proton spin couplings, but not both. More recently, semi-empirical models employing nuclear magnetic moment data have been used to derive new constraints for non-valence nucleons. However, comparison of such semi-empirical models to detailed large-scale nuclear shell model calculations and analysis of known physical effects in nuclei show that existing semi-empirical models cannot reliably be used to predict the spin polarization of non-valence nucleons. The results of our re-analysis of nuclear spin content are applied to searches for exotic long-range monopole–dipole and dipole–dipole couplings of nuclei leading to significant revisions of some published constraints. (paper)

Thermal-hydraulic and neutron-physical characteristics of a new SCWR fuel assembly

International Nuclear Information System (INIS)

Liu, X.J.; Cheng, X.

2009-01-01

A new fuel assembly design for a thermal supercritical water cooled reactor (SCWR) core is proposed. Compared to the existing fuel assemblies, the present fuel assembly has two-rows of fuel rods between the moderator channels, to achieve a more uniform moderation for all fuel rod cells, and subsequently, a more uniform radial power distribution. In addition, a neutron-kinetics/thermal-hydraulics coupling method is developed, to analyze the neutron-physical and thermal-hydraulic behavior of the fuel assembly designs. This coupling method is based on the sub-channel analysis code COBRA-IV for thermal-hydraulics and the neutron-kinetics code SKETCH-N for neutron-physics. Both the COBRA-IV code and the SKETCH-N code are accordingly modified. An interface is established for the data transfer between these two codes. This coupling method is applied to both the one-row fuel assemblies (previous design) and the two-row fuel assemblies (present design). The performance of the two types of fuel assemblies is compared. The results show clearly that the two-row fuel assembly has more favorable neutron-physical and thermal-hydraulic characteristics than the one-row fuel assembly. The effect of various parameters on the fuel assembly performance is discussed. The coupling method is proven to be well suitable for further applications to SCWR fuel assembly design analysis

Dispersion and decay of collective modes in neutron star cores

OpenAIRE

Kobyakov, D. N.; Pethick, C. J.; Reddy, S.; Schwenk, A.

2017-01-01

We calculate the frequencies of collective modes of neutrons, protons and electrons in the outer core of neutron stars. The neutrons and protons are treated in a hydrodynamic approximation and the electrons are regarded as collisionless. The coupling of the nucleons to the electrons leads to Landau damping of the collective modes and to significant dispersion of the low-lying modes. We investigate the sensitivity of the mode frequencies to the strength of entrainment between neutrons and prot...

Constraints on mirror models of dark matter from observable neutron-mirror neutron oscillation

Directory of Open Access Journals (Sweden)

Rabindra N. Mohapatra

2018-01-01

Full Text Available The process of neutron-mirror neutron oscillation, motivated by symmetric mirror dark matter models, is governed by two parameters: n−n′ mixing parameter δ and n−n′ mass splitting Δ. For neutron mirror neutron oscillation to be observable, the splitting between their masses Δ must be small and current experiments lead to δ≤2×10−27 GeV and Δ≤10−24 GeV. We show that in mirror universe models where this process is observable, this small mass splitting constrains the way that one must implement asymmetric inflation to satisfy the limits of Big Bang Nucleosynthesis on the number of effective light degrees of freedom. In particular we find that if asymmetric inflation is implemented by inflaton decay to color or electroweak charged particles, the oscillation is unobservable. Also if one uses SM singlet fields for this purpose, they must be weakly coupled to the SM fields.

Type-I superconductivity and neutron star precession

International Nuclear Information System (INIS)

Sedrakian, Armen

2005-01-01

Type-I proton superconducting cores of neutron stars break up in a magnetic field into alternating domains of superconducting and normal fluids. We examine two channels of superfluid-normal fluid friction where (i) rotational vortices are decoupled from the nonsuperconducting domains and the interaction is due to the strong force between protons and neutrons; (ii) the nonsuperconducting domains are dynamically coupled to the vortices and the vortex motion generates transverse electric fields within them, causing electronic current flow and Ohmic dissipation. The obtained dissipation coefficients are consistent with the Eulerian precession of neutron stars

Neutron stars in relativistic mean field theory with isovector scalar meson

Energy Technology Data Exchange (ETDEWEB)

Kubis, S.; Kutschera, M.; Stachniewicz, S. [H. Niewodniczanski Institute of Nuclear Physics, Cracow (Poland)

1998-03-01

We study the equation of state (EOS) of {beta}-stable dense matter and models of neutron stars in the relativistic mean field (RMF) theory with the isovector scalar mean field corresponding to the {delta}-meson (a{sub 0}(980)). A range of values of the {delta}-meson coupling compatible with the Bonn potentials is explored. Parameters of the model in the isovector sector are constrained to fit the nuclear symmetry energy, E{sub s}{approx}30 MeV. We find that the quantity most sensitive to the {delta}-meson coupling is the proton fraction of neutron star matter. It increases significantly in the presence of the {delta}-field. The energy per baryon also increases but the effect is smaller. The EOS becomes slightly stiffer and the maximum neutron star mass increases for stronger {delta}-meson coupling. (author) 8 refs, 6 figs, 2 tabs

Neutron scattering cross sections for 204,206Pb and neutron and proton amplitudes of E2 and E3 excitations

International Nuclear Information System (INIS)

Hicks, S.F.; Hanly, J.M.; Hicks, S.E.; Shen, G.R.; McEllistrem, M.T.

1994-01-01

Differential elastic and inelastic scattering cross sections have been measured for neutrons incident on 204 Pb and 206 Pb at energies of 2.5, 4.6, and 8.0 MeV and total cross sections in 100-keV steps from 250 keV to 4.0 MeV. Both spherical and coupled-channels analyses have been used to interpret this large set of data, together with other cross sections extending to 8 MeV. Several purposes motivate this work. The first is to establish the dispersion-corrected mean field appropriate for these nuclei. A consistent description of the energy dependent neutron scattering potential includes a dispersion relation connecting the real and imaginary parts of the potential; the resultant potential relates the energy dependent scattering field to one representing bound single particle levels. Dispersion relations using both the single channel and coupled-channels models have been examined; both give very similar results. The second motivation is to deduce neutron and proton excitation strengths of the lowest-energy quadrupole and octupole excitations seen via neutron scattering, and to compare those strengths with similar values derived from electromagnetic exciton, heavy-ion and pion scattering. The role of target neutrons in both collective excitations was found to be enhanced compared to the proton role

Average cross sections calculated in various neutron fields

International Nuclear Information System (INIS)

Shibata, Keiichi

2002-01-01

Average cross sections have been calculated for the reactions contained in the dosimetry files, JENDL/D-99, IRDF-90V2, and RRDF-98 in order to select the best data for the new library IRDF-2002. The neutron spectra used in the calculations are as follows: 1) 252 Cf spontaneous fission spectrum (NBS evaluation), 2) 235 U thermal fission spectrum (NBS evaluation), 3) Intermediate-energy Standard Neutron Field (ISNF), 4) Coupled Fast Reactivity Measurement Facility (CFRMF), 5) Coupled thermal/fast uranium and boron carbide spherical assembly (ΣΣ), 6) Fast neutron source reactor (YAYOI), 7) Experimental fast reactor (JOYO), 8) Japan Material Testing Reactor (JMTR), 9) d-Li neutron spectrum with a 2-MeV deuteron beam. The items 3)-7) represent fast neutron spectra, while JMTR is a light water reactor. The Q-value for the d-Li reaction mentioned above is 15.02 MeV. Therefore, neutrons with energies up to 17 MeV can be produced in the d-Li reaction. The calculated average cross sections were compared with the measurements. Figures 1-9 show the ratios of the calculations to the experimental data which are given. It is found from these figures that the 58 Fe(n, γ) cross section in JENDL/D-99 reproduces the measurements in the thermal and fast reactor spectra better than that in IRDF-90V2. (author)

Importance estimation in Monte Carlo modelling of neutron and photon transport

International Nuclear Information System (INIS)

Mickael, M.W.

1992-01-01

The estimation of neutron and photon importance in a three-dimensional geometry is achieved using a coupled Monte Carlo and diffusion theory calculation. The parameters required for the solution of the multigroup adjoint diffusion equation are estimated from an analog Monte Carlo simulation of the system under investigation. The solution of the adjoint diffusion equation is then used as an estimate of the particle importance in the actual simulation. This approach provides an automated and efficient variance reduction method for Monte Carlo simulations. The technique has been successfully applied to Monte Carlo simulation of neutron and coupled neutron-photon transport in the nuclear well-logging field. The results show that the importance maps obtained in a few minutes of computer time using this technique are in good agreement with Monte Carlo generated importance maps that require prohibitive computing times. The application of this method to Monte Carlo modelling of the response of neutron porosity and pulsed neutron instruments has resulted in major reductions in computation time. (Author)

Neutron scattering studies of the heavy Fermion superconductors

International Nuclear Information System (INIS)

Goldman, A.I.

1985-01-01

Recent neutron scattering measurements of the heavy Fermion superconductors are described. Those materials offer an exciting opportunity for neutron scattering since the f-electrons, which couple directly to magnetic scattering measurements, seem to be the same electrons which form the superconducting state below T/sub c/. In addition, studies of the magnetic fluctuations in these, and other heavy Fermion systems, by inelastic magnetic neutron scattering can provide information about the nature of the low temperature Fermi liquid character of these novel compounds

Evolution of Neutron Stars and Observational Constraints

Directory of Open Access Journals (Sweden)

Lattimer J.

2010-10-01

Full Text Available The structure and evolution of neutron stars is discussed with a view towards constraining the properties of high density matter through observations. The structure of neutron stars is illuminated through the use of several analytical solutions of Einstein’s equations which, together with the maximally compact equation of state, establish extreme limits for neutron stars and approximations for binding energies, moments of inertia and crustal properties as a function of compactness. The role of the nuclear symmetry energy is highlighted and constraints from laboratory experiments such as nuclear masses and heavy ion collisions are presented. Observed neutron star masses and radius limits from several techniques, such as thermal emissions, X-ray bursts, gammaray flares, pulsar spins and glitches, spin-orbit coupling in binary pulsars, and neutron star cooling, are discussed. The lectures conclude with a discusson of proto-neutron stars and their neutrino signatures.

RESONANCE CONTROL FOR THE COUPLED CAVITY LINAC AND DRIFT TUBE LINAC STRUCTURES OF THE SPALLATION NEUTRON SOURCE LINAC USING A CLOSED-LOOP WATER COOLING SYSTEM

International Nuclear Information System (INIS)

Bernardin, J.D.; Brown, R.L.

2001-01-01

The Spallation Neutron Source (SNS) is a facility being designed for scientific and industrial research and development. SNS will generate and use neutrons as a diagnostic tool for medical purposes, material science, etc. The neutrons will be produced by bombarding a heavy metal target with a high-energy beam of protons, generated and accelerated with a linear particle accelerator, or linac. The low energy end of the linac consists of two room temperature copper structures, the drift tube linac (DTL), and the coupled cavity linac (CCL). Both of these accelerating structures use large amounts of electrical energy to accelerate the protons to an energy of 185 MeV. Approximately 60-80% of the electrical energy is dissipated in the copper structure and must be removed. This is done using specifically designed water cooling passages within the linac's copper structure. Cooling water is supplied to these cooling passages by specially designed resonance control and water cooling systems

3D pin-by-pin power density profiles with high spatial resolution in the vicinity of a BWR control blade tip simulated with coupled neutronics/burn-up calculations

International Nuclear Information System (INIS)

Li, J.; Nünighoff, K.; Allelein, H.-J.

2011-01-01

Highlights: ► High spatial resolution neutronic and burn-up calculations of quarter BWR fuel element section. ► Coupled MCNP(X)–ORIGEN2.2 simulation using VESTA. ► Control blade history effect was taken into account. ► Determining local power excursion after instantaneous control rod movement. ► Correlation between control blade geometry and occurrence of local power excursions. - Abstract: Pellet cladding interaction (PCI) as well as pellet cladding mechanical interaction (PCMI) are well-known fuel failures in light water reactors, especially in boiling water reactors (BWR). Whereas the thermo-mechanical processes of PCI effects have been intensively investigated in the last decades, only rare information is available on the role of neutron physics. However, each power transient is primary due to neutron physics effects and thus knowledge of the neutron physical background is mandatory to better understand the occurrence of PCI effects in BWRs. This paper will focus on a study of local power excursions in a typical BWR fuel assembly during control rod movements. Burn-up and energy deposition were simulated with high spatial granularity, especially in the vicinity of the control blade tip. It could be shown, that the design of the control blade plays a dominant role for the occurrence of local power peaks while instantaneously moving down the control rod. The main result is, that the largest power peak occurs at the interface between steel handle and absorber rods. A full width half maximum (FWHM) of ±2.5 cm was observed. This means, the local power excursion due to neutron physics phenomena involve approximately five pellets. With the VESTA code coupled MCNP(X)/ORIGEN2.2 calculations were performed with more than 3400 burn-up zones in order to take history effects into account.

Topics in the theory of neutron star cooling

International Nuclear Information System (INIS)

Duncan, R.C. Jr.

1986-01-01

The author calculates the neutrino emissivity of interacting, degenerate quark matter, which may make up the dense cores of neutron stars. QCD interactions between quarks are included to first order. The author shows that when massive s-quarks are present in cold quark matter, electrons are not present in equilibrium at densities above a threshold electron extinction density n/sub ex/. This results in a much lower neutrino emissivity epsilon/sub nu/ at high densities than has been previously calculated. Dependences of epsilon/sub nu/ on the strange quark mass m/sub s/ and the QCD coupling constant a/sub c/ are determined for a quark liquid in β-equilibrium. Implications of these calculations for neutron-star cooling are briefly discussed. Eventually, it is shown that neutrino momentum effects may be ignored in neutron star cooling calculations without significant error, even when high-density quark-matter cores are present. Finally considered is the very early cooling epoch, lasting up to ∼1 minutes after formation, when a neutron star is optically thick to neutrinos. It is shown that the coupled equations of neutrino and photon transport in the atmosphere of a sufficiently hot, nascent neutron star do not admit hydrostatic solutions

Non-equilibrium thermodynamics and energy distribution function of neutron gas in constant power reactor under coupling of neutrons and medium

International Nuclear Information System (INIS)

Hayasaka, Hideo

1983-01-01

The thermodynamics and the energy distribution function of the neutron gas in a constant power reactor are considered, taking into account the burn-up of fuel. To separate the secular motion of neutrons owing to fuel burn-up and the microscopic fluctuations of neutrons around this motion, a long time of the order of several months is divided into m equal intervals, and the respective states corresponding to m small time intervals are treated as quasi-stationary states. The local energy distribution function of the neutron gas in the quasi-stationary state is given by a generalized Boltzmann distribution specified by the respective generalized activity coefficient for each subsystem. The effects of fuel burn-up on the respective distribution functions for successive small time intervals are taken into account through various quantities relating to reactor physics, depending upon the fuel burn-up, by successive approximation. (author)

Status of the low energy neutron source at Indiana University

International Nuclear Information System (INIS)

Baxter, D.V.; Cameron, J.M.; Derenchuk, V.P.; Lavelle, C.M.; Leuschner, M.B.; Lone, M.A.; Meyer, H.O.; Rinckel, T.; Snow, W.M.

2005-01-01

The National Science Foundation has recently approved funding for LENS (the low energy neutron source) at Indiana University and construction of this facility has begun. LENS represents a new paradigm for economically introducing neutron scattering into a university or industrial setting. In this design, neutrons are produced in a long-pulse (1 ms) mode through (p,n) reactions on a water-cooled Be target and the target is tightly coupled to a cryogenic moderator with a water reflector. This design gives a facility suitable for materials research, the development of new neutron instrumentation, and the education of new neutron scientists

One-neutron knockout from Ne24-28 isotopes

CERN Document Server

Rodriguez-Tajes, C; Caamano, M; Faestermann, T; Cortina-Gil, D; Zhukov, M; Simon, H; Nilsson, T; Borge, M J G; Alvarez-Pol, H; Winkler, M; Prochazka, A; Nociforo, C; Weick, H; Kanungo, R; Perez-Loureiro, D; Kurtukian, T; Suemmerer, K; Eppinger, K; Perea, A; Chatillon, A; Maierbeck, P; Benlliure, J; Pascual-Izarra, C; Gernhaeuser, R; Geissel, H; Aumann, T; Kruecken, R; Larsson, K; Tengblad, O; Benjamim, E; Jonson, B; Casarejos, E

2010-01-01

One-neutron knockout reactions of Ne24-28 in a beryllium target have been studied in the Fragment Separator (FRS), at GSI. The results include inclusive one-neutron knockout cross-sections as well as longitudinal-momentum distributions of the knockout fragments. The ground-state structure of the neutron-rich neon isotopes was obtained from an analysis of the measured momentum distributions. The results indicate that the two heaviest isotopes, Ne-27 and Ne-28, are dominated by a configuration in which a s(1/2) neutron is coupled to an excited state of the Ne-26 and Ne-27 core, respectively. (C) 2010 Elsevier B.V. All rights reserved.

Specification of the Advanced Burner Test Reactor Multi-Physics Coupling Demonstration Problem

Energy Technology Data Exchange (ETDEWEB)

Shemon, E. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Grudzinski, J. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Lee, C. H. [Argonne National Lab. (ANL), Argonne, IL (United States); Thomas, J. W. [Argonne National Lab. (ANL), Argonne, IL (United States); Yu, Y. Q. [Argonne National Lab. (ANL), Argonne, IL (United States)

2015-12-21

This document specifies the multi-physics nuclear reactor demonstration problem using the SHARP software package developed by NEAMS. The SHARP toolset simulates the key coupled physics phenomena inside a nuclear reactor. The PROTEUS neutronics code models the neutron transport within the system, the Nek5000 computational fluid dynamics code models the fluid flow and heat transfer, and the DIABLO structural mechanics code models structural and mechanical deformation. The three codes are coupled to the MOAB mesh framework which allows feedback from neutronics, fluid mechanics, and mechanical deformation in a compatible format.

Coupled Model of channels in parallel and neutron kinetics in two dimensions; Modelo acoplado de canales en paralelo y cinetica neutronica en dos dimensiones

Energy Technology Data Exchange (ETDEWEB)

Cecenas F, M.; Campos G, R.M. [Instituto de Investigaciones Electricas, Av. Reforma 113, Col. Palmira, 62490 Cuernavaca, Morelos (Mexico)]. E-mail: mcf@iie.org.mx; Valle G, E. del [IPN, ESFM, 07738 Mexico D.F. (Mexico)

2004-07-01

In this work an arrangement of thermohydraulic channels is presented that represent those four quadrants of a nucleus of reactor type BWR. The channels are coupled to a model of neutronic in two dimensions that allow to generate the radial profile of power of the reactor. Nevertheless that the neutronic pattern is of two dimensions, it is supplemented with axial additional information when considering the axial profiles of power for each thermo hydraulic channel. The stationary state is obtained the one it imposes as frontier condition the same pressure drop for all the channels. This condition is satisfied to iterating on the flow of coolant in each channel to equal the pressure drop in all the channels. This stationary state is perturbed later on when modifying the values for the effective sections corresponding to an it assembles. The calculation in parallel of the neutronic and the thermo hydraulic is carried out with Vpm (Virtual parallel machine) by means of an outline teacher-slave in a local net of computers. (Author)

Development of a high efficient conventional type cold neutron source using a non-explosive material

International Nuclear Information System (INIS)

Kiyanagi, Y.; Satoh, S.

1999-01-01

An efficient cold moderator that can be used easily at a small neutron source would be useful for neutron radiography, prompt gamma ray analysis and so on. Non-explosive materials are chosen for a cold moderator since explosive materials such as hydrogen and methane require a safety system. Neutronic performances of coupled moderators of various non-explosive materials are studied so as to develop such a cold moderator since the coupled moderator system is the best to obtain high intensity of cold neutrons. Effect of premoderator is studied and neutron spectra from methanol, ethanol, benzene, mesitylene and benzene methanol are measured around 20 K. The premoderator increased the cold neutron intensity by about 50∼70%. Methanol and mesitylene gave the highest cold neutron intensity. Effect of Be filter-reflector is also studied and a intensity gain of about 20% was obtained below about 5 MeV. (author)

Delayed neutron emission near the shell-closures

Directory of Open Access Journals (Sweden)

Borzov Ivan

2016-01-01

Full Text Available The self-consistent Density Functional + Continuum QRPA approach (DF+CQRPA provides a good description of the recent experimental beta-decay half-lives and delayed neutron emission branchings for the nuclei approaching to (and beyond the neutron closed shells N = 28; 50; 82. Predictions of beta-decay properties are more reliable than the ones of standard global approaches traditionally used for the r-process modelling. An impact of the quasi-particle phonon coupling on the delayed multi-neutron emission rates P2n, P3n,… near the closed shells is also discussed.

Development of precision elliptic neutron-focusing supermirror.

Science.gov (United States)

Hosobata, Takuya; Yamada, Norifumi L; Hino, Masahiro; Yamagata, Yutaka; Kawai, Toshihide; Yoshinaga, Hisao; Hori, Koichiro; Takeda, Masahiro; Takeda, Shin; Morita, Shin-Ya

2017-08-21

This paper details methods for the precision design and fabrication of neutron-focusing supermirrors, based on electroless nickel plating. We fabricated an elliptic mirror for neutron reflectometry, which is our second mirror improved from the first. The mirror is a 550-millimeter-long segmented mirror assembled using kinematic couplings, with each segment figured by diamond cutting, polished using colloidal silica, and supermirror coated through ion-beam sputtering. The mirror was evaluated with neutron beams, and the reflectivity was found to be 68-90% at a critical angle. The focusing width was 0.17 mm at the full width at half maximum.

Quantum physics with neutrons

International Nuclear Information System (INIS)

Durstberger, K.; Hasegawa, Y.; Klepp, J.; Sulyok, G.; Rauch, H.

2008-01-01

Full text: Fundamental quantum properties like quantum coherence and entanglement are among the most interesting features of quantum mechanics. The physical system of interest is the (massive) neutron subjected to interferometric and polarimetric measurements. Neutrons are proper objects for a study of quantum mechanical behavior: they allow for rather easy experimental control and the neutron spin is the simplest two-level system with easy manipulation by magnetic fields. In combination with interferometric measurements the system has enough intrinsic richness to show interesting quantum features such as entanglement. The coupling of the neutron to an external magnetic field allows for selective manipulations of the spinor quantum states. This can be used, on the one hand, to create entangled states where the entanglement occurs between different degrees of freedom (e.g. spin and path) and, on the other hand, one can introduce dephasing and decoherence by varying magnetic fields. We investigate different kinds of entanglement for the neutron system and mechanisms for decoherence and dephasing. We discuss weak measurements and their realization for neutrons where information about the system can be revealed without disturbing the system too much. Beyond the theoretical work we develop experimental strategies to check the results directly in suitably designed experiments. The experimental work is done at the Institute Laue-Langvine (ILL) in Grenoble, France. (author)

Characteristics of the WNR: a pulsed spallation neutron source

International Nuclear Information System (INIS)

Russell, G.J.; Lisowski, P.W.; Howe, S.D.; King, N.S.P.; Meier, M.M.

1982-01-01

The Weapons Neutron Research facility (WNR) is a pulsed spallation neutron source in operation at the Los Alamos National Laboratory. The WNR uses part of the 800-MeV proton beam from the Clinton P. Anderson Meson Physics Facility accelerator. By choosing different target and moderator configurations and varying the proton pulse structure, the WNR can provide a white neutron source spanning the energy range from a few MeV to 800 MeV. The neutron spectrum from a bare target has been measured and is compared with predictions using an Intranuclear Cascade model coupled to a Monte Carlo transport code. Calculations and measurements of the neutronics of WNR target-moderator assemblies are presented

Reflector optimization for coupled liquid hydrogen moderator

International Nuclear Information System (INIS)

Kiyanagi, Y.; Iwasa, H.; Watanabe, N.; Furusaka, M.

1991-01-01

As a part of optimization studies on a coupled liquid hydrogen moderator system, the optimal thickness of the reflector, the effects of neutron absorbing liners and other beam hole/moderator on the cold-neutron-beam intensity were studied experimentally. It turns out that the optimal thickness is rather thick in this system and the existence of Cd liners around the beam extraction hole considerably reduces the cold neutron beam intensity, while the existence of other beam hole and moderator does not give an important intensity reduction. (author)

Fast-neutron total and scattering cross sections of sup 58 Ni and nuclear models

Energy Technology Data Exchange (ETDEWEB)

Smith, A.B.; Guenther, P.T.; Whalen, J.F. (Argonne National Lab., IL (United States)); Chiba, S. (Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment)

1991-07-01

The neutron total cross sections of {sup 58}Ni were measured from {approx} 1 to > 10 MeV using white-source techniques. Differential neutron elastic-scattering cross sections were measured from {approx} 4.5 to 10 MeV at {approx} 0.5 MeV intervals with {ge} 75 differential values per distribution. Differential neutron inelastic-scattering cross sections were measured, corresponding to fourteen levels with excitations up to 4.8 MeV. The measured results, combined with relevant values available in the literature, were interpreted in terms of optical-statistical and coupled-channels model using both vibrational and rotational coupling schemes. The physical implications of the experimental results nd their interpretation are discussed in the contexts of optical-statistical, dispersive-optical, and coupled-channels models. 61 refs.

Dispersion and decay of collective modes in neutron star cores

Science.gov (United States)

Kobyakov, D. N.; Pethick, C. J.; Reddy, S.; Schwenk, A.

2017-08-01

We calculate the frequencies of collective modes of neutrons, protons, and electrons in the outer core of neutron stars. The neutrons and protons are treated in a hydrodynamic approximation and the electrons are regarded as collisionless. The coupling of the nucleons to the electrons leads to Landau damping of the collective modes and to significant dispersion of the low-lying modes. We investigate the sensitivity of the mode frequencies to the strength of entrainment between neutrons and protons, which is not well characterized. The contribution of collective modes to the thermal conductivity is evaluated.

Single- and two-phase flow modeling for coupled neutronics / thermal-hydraulics transient analysis of advanced sodium-cooled fast reactors

International Nuclear Information System (INIS)

Chenu, A.

2011-10-01

Nuclear power is nowadays in the front rank as regards helping to meet the growing worldwide energy demand while avoiding an excessive increase in greenhouse gas emissions. However, the operating nuclear power plants are mainly thermal-neutron reactors and, as such, can not be maintained on the basis of the currently identified uranium resources beyond one century at the present consumption rate. Sustainability of nuclear power thus involves closure of the fuel cycle through breeding. With a uranium-based fuel, breeding can only be achieved using a fast-neutron reactor. Sodium-cooled fast reactor (SFR) technology benefits from 400 reactor-years of accumulated experience and is thus a prime candidate for the implementation of so-called Generation-IV nuclear energy systems. In this context, the safety demonstration of SFRs remains a major Research and Development related issue. The current research aims at the development of a computational tool for the in-depth understanding of SFR core behaviour during accidental transients, particularly those including boiling of the coolant. An accurate modelling of the core physics during such transients requires the coupling between 3D neutron kinetics and thermal-hydraulics in the core, to account for the strong interactions between the two-phase coolant flow and power variations caused by the sodium void effect. The present study is specifically focused upon models for the representation of sodium two-phase flow. The extension of the thermal-hydraulics TRACE code, previously limited to the simulation of single-phase sodium flow, has been carried out through the implementation of equations-of-state and closure relations specific to sodium. The different correlations have then been implemented as options. From the validation study carried out, it has been possible to recommend a set of models which provide satisfactory results, while considering annular flow as the dominant regime up to dryout and a smooth breakdown of the

Fast-neutrons incident on rotors: Tantalum

International Nuclear Information System (INIS)

Smith, Alan B.

2005-01-01

Mono-energetic neutrons are elastically and inelastically scattered from elemental tantalum at incident energies of ∼0.3-10.0 MeV. These experimental results are augmented with neutron total-cross-section and additional neutron-scattering data from the literature to form a composite experimental database. The latter is interpreted in the context of optical-statistical and coupled-channels models, including consideration of collective rotations, dispersion effects and other physical properties. The results are compared with those of similar processes in this region of collective nuclei. A regional model is proposed for the interpretation and prediction of such interactions. The model and the experimental results are compared with corresponding values given in ENDF/B-VI

Optimal power and distribution control for weakly-coupled-core reactor

International Nuclear Information System (INIS)

Oohori, Takahumi; Kaji, Ikuo

1977-01-01

A numerical procedure has been devised for obtaining the optimal power and distribution control for a weakly-coupled-core reactor. Several difficulties were encountered in solving this optimization problem: (1) nonlinearity of the reactor kinetics equations; (2) neutron-leakage interaction between the cores; (3) localized power changes occurring in addition to the total power changes; (4) constraints imposed on the states - e.g. reactivity, reactor period. To obviate these difficulties, use is made of the generalized Newton method to convert the problem into an iterative sequence of linear programming problems, after approximating the differential equations and the integral performance criterion by a set of discrete algebraic equations. In this procedure, a heuristic but effective method is used for deriving an initial approximation, which is then made to converge toward the optimal solution. Delayed-neutron one-group point reactor models embodying transient temperature feed-back to the reactivity are used in obtaining the kinetics equations for the weakly-coupled-core reactor. The criterion adopted for determining the optimality is a norm relevant to the deviations of neutron density from the desired trajectories or else to the time derivatives of the neutron density; uniform control intervals are prescribed. Examples are given of two coupled-core reactors with typical parameters to illustrate the results obtained with this procedure. A comparison is also made between the coupled-core reactor and the one-point reactor. (auth.)

The Standard Model and the neutron beta-decay

CERN Document Server

Abele, H

2000-01-01

This article reviews the relationship between the observables in neutron beta-decay and the accepted modern theory of particle physics known as the Standard Model. Recent neutron-decay measurements of various mixed American-British-French-German-Russian collaborations try to shed light on the following topics: the coupling strength of charged weak currents, the universality of the electroweak interaction and the origin of parity violation.

The impact of the tensor interaction on the β-delayed neutron emission of the neutron-rich Ni isotopes

Directory of Open Access Journals (Sweden)

Sushenok E.O.

2018-01-01

Full Text Available The neutron emission of the β-decay of 74;76;78;80Ni are studied within the quasiparticle random phase approximation with the Skyrme interaction. The coupling between one- and two-phonon terms in the wave functions of the low-energy 1+ states of the daughter nuclei is taken into account. It is shown that the strength decrease of the neutronproton tensor interaction leads to the increase of the half-life and the neutron-emission probability.

Development of sub-channel code SACoS and its application in coupled neutronics/thermal hydraulics system for SCWR

International Nuclear Information System (INIS)

Chaudri, Khurrum Saleem; Su Yali; Chen Ronghua; Tian Wenxi; Su Guanghui; Qiu Suizheng

2012-01-01

Highlights: ► A tool is developed for coupled neutronics/thermal-hydraulic analysis for SCWR. ► For thermal hydraulic analysis, a sub-channel code SACoS is developed and verified. ► Coupled analysis agree quite well with the reference calculations. ► Different choice of important parameters makes huge difference in design calculations. - Abstract: Supercritical Water Reactor (SCWR) is one of the promising reactors from the list of fourth generation of nuclear reactors. High thermal efficiency and low cost of electricity make it an attractive option in the era of growing energy demand. An almost seven fold density variation for coolant/moderator along the active height does not allow the use of constant density assumption for design calculations, as used for previous generations of reactors. The advancement in computer technology gives us the superior option of performing coupled analysis. Thermal hydraulics calculations of supercritical water systems present extra challenges as not many computational tools are available to perform that job. This paper introduces a new sub-channel code called Sub-channel Analysis Code of SCWR (SACoS) and its application in coupled analyses of High Performance Light Water Reactor (HPLWR). SACoS can compute the basic thermal hydraulic parameters needed for design studies of a supercritical water reactor. Multiple heat transfer and pressure drop correlations are incorporated in the code according to the flow regime. It has the additional capability of calculating the thermal hydraulic parameters of moderator flowing in water box and between fuel assemblies under co-current or counter current flow conditions. Using MCNP4c and SACoS, a coupled system has been developed for SCWR design analyses. The developed coupled system is verified by performing and comparing HPLWR calculations. The results were found to be in very good agreement. Significant difference between the results was seen when Doppler feedback effect was included in

MCNP4C2, Coupled Neutron, Electron Gamma 3-D Time-Dependent Monte Carlo Transport Calculations

International Nuclear Information System (INIS)

2002-01-01

1 - Description of program or function: MCNP is a general-purpose, continuous-energy, generalized geometry, time-dependent, coupled neutron-photon-electron Monte Carlo transport code system. MCNP4C2 is an interim release of MCNP4C with distribution restricted to the Criticality Safety community and attendees of the LANL MCNP workshops. The major new features of MCNP4C2 include: - Photonuclear physics; - Interactive plotting; - Plot superimposed weight window mesh; - Implement remaining macro-body surfaces; - Upgrade macro-bodies to surface sources and other capabilities; - Revised summary tables; - Weight window improvements. See the MCNP home page more information http://www-xdiv.lanl.gov/XCI/PROJECTS/MCNP with a link to the MCNP Forum. See the Electronic Notebook at http://www-rsicc.ornl.gov/rsic.html for information on user experiences with MCNP. 2 - Methods:MCNP treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and some special fourth-degree surfaces. Pointwise continuous-energy cross section data are used, although multigroup data may also be used. Fixed-source adjoint calculations may be made with the multigroup data option. For neutrons, all reactions in a particular cross-section evaluation are accounted for. Both free gas and S(alpha, beta) thermal treatments are used. Criticality sources as well as fixed and surface sources are available. For photons, the code takes account of incoherent and coherent scattering with and without electron binding effects, the possibility of fluorescent emission following photoelectric absorption, and absorption in pair production with local emission of annihilation radiation. A very general source and tally structure is available. The tallies have extensive statistical analysis of convergence. Rapid convergence is enabled by a wide variety of variance reduction methods. Energy ranges are 0-60 MeV for neutrons (data generally only available up to

Transport calculations for a 14.8 MeV neutron beam in a water phantom

International Nuclear Information System (INIS)

Goetsch, S.J.

1981-01-01

A coupled neutron/photon Monte Carlo radiation transport code (MORSE-CG) has been used to calculate neutron and photon doses in a water phantom irradiated by 14.8 MeV neutrons from the Gas Target Neutron Source. The source-collimator-phantom geometry was carefully simulated. Results of calculations utilizing two different statistical estimators (next-collision and track-length) are presented

Developing an interface between MCNP and McStas for simulation of neutron moderators

DEFF Research Database (Denmark)

Klinkby, Esben Bryndt; Lauritzen, Bent; Nonbøl, Erik

2012-01-01

Simulations of target-moderator-reflector system at spallation sources are conventionally carried out using MCNP/X whereas simulations of neutron transport and instrument performance are carried out by neutron ray tracing codes such as McStas. The coupling between the two simulations suites...... typically consists of providing analytical fits from MCNP/X neutron spectra to McStas. This method is generally successful, but as will be discussed in the this paper, there are limitations and a more direct coupling between MCNP/X andMcStas could allow for more accurate simulations of e.g. complex...... moderator geometries, interference between beamlines as well as shielding requirements along the neutron guides. In this paper different possible interfaces between McStas and MCNP/X are discussed and first preliminary performance results are shown....

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)

D-D neutron energy-spectra measurements in Alcator C

International Nuclear Information System (INIS)

Pappas, D.S.; Wysocki, F.J.; Furnstahl, R.J.

1982-08-01

Measurements of energy spectra of neutrons produced during high density (anti n/sub e/ > 2 x 10 14 cm -3 ) deuterium discharges have been performed using a proton-recoil (NE 213) spectrometer. A two foot section of light pipe (coupling the scintillator and photomultiplier) was used to extend the scintillator into a diagnostic viewing port to maximize the neutron detection efficiency while not imposing excessive magnetic shielding requirements. A derivative unfolding technique was used to deduce the energy spectra. The results showed a well defined peak at 2.5 MeV which was consistent with earlier neutron flux measurements on Alcator C that indicated the neutrons were of thermonuclear origin

One-neutron knockout from {sup 24-28}Ne isotopes

Energy Technology Data Exchange (ETDEWEB)

Rodriguez-Tajes, C., E-mail: carme.rodriguez@usc.e [Departamento de Fisica de Particulas, Universidade de Santiago de Compostela, 15782 Santiago de Compostela (Spain); Cortina-Gil, D.; Alvarez-Pol, H. [Departamento de Fisica de Particulas, Universidade de Santiago de Compostela, 15782 Santiago de Compostela (Spain); Aumann, T. [GSI Helmholtzzentrum fuer Schwerionenforschung, 64291 Darmstadt (Germany); Benjamim, E.; Benlliure, J. [Departamento de Fisica de Particulas, Universidade de Santiago de Compostela, 15782 Santiago de Compostela (Spain); Borge, M.J.G. [Instituto de Estructura de la Materia, CSIC, 28006 Madrid (Spain); Caamano, M.; Casarejos, E. [Departamento de Fisica de Particulas, Universidade de Santiago de Compostela, 15782 Santiago de Compostela (Spain); Chatillon, A. [GSI Helmholtzzentrum fuer Schwerionenforschung, 64291 Darmstadt (Germany); Eppinger, K.; Faestermann, T. [Physik Department E12, Technische Universitaet Muenchen, 85748 Garching (Germany); Gascon, M. [Departamento de Fisica de Particulas, Universidade de Santiago de Compostela, 15782 Santiago de Compostela (Spain); Geissel, H. [GSI Helmholtzzentrum fuer Schwerionenforschung, 64291 Darmstadt (Germany); Gernhaeuser, R. [Physik Department E12, Technische Universitaet Muenchen, 85748 Garching (Germany); Jonson, B. [Fundamental Fysik, Chalmers Tekniska Hoegskola, 412 96 Goeteborg (Sweden); PH Department, CERN, 1211 Geneve 23 (Switzerland); Kanungo, R. [Astronomy and Physics Department, Saint Mary' s University, Halifax, NS B3H 3C3 (Canada); Kruecken, R. [Physik Department E12, Technische Universitaet Muenchen, 85748 Garching (Germany); Kurtukian, T. [Departamento de Fisica de Particulas, Universidade de Santiago de Compostela, 15782 Santiago de Compostela (Spain); Larsson, K. [Fundamental Fysik, Chalmers Tekniska Hoegskola, 412 96 Goeteborg (Sweden)

2010-04-05

One-neutron knockout reactions of {sup 24-28}Ne in a beryllium target have been studied in the Fragment Separator (FRS), at GSI. The results include inclusive one-neutron knockout cross-sections as well as longitudinal-momentum distributions of the knockout fragments. The ground-state structure of the neutron-rich neon isotopes was obtained from an analysis of the measured momentum distributions. The results indicate that the two heaviest isotopes, {sup 27}Ne and {sup 28}Ne, are dominated by a configuration in which a s{sub 1/2} neutron is coupled to an excited state of the {sup 26}Ne and {sup 27}Ne core, respectively.

Development of criticality accident detector measuring neutrons and gamma-rays

International Nuclear Information System (INIS)

Tsujimura, Norio; Yoshida, Tadayoshi; Ishii, Masato

2005-01-01

The authors developed a new criticality accident detector measuring neutrons and gamma-rays. The detector is a cylindrical plastic scintillator coupled to a current-mode operated photomultiplier, and is covered by an inner cadmium shell, acting as a neutron to gamma-ray converter, and a 5cm thick outer polyethylene moderator in order to respond to the same threshold triggering dose regardless of whether it was exposed to neutrons, gamma-rays or a mixture of the two radiations. (author)

BPS Skyrmions as neutron stars

Energy Technology Data Exchange (ETDEWEB)

Adam, C., E-mail: adam@fpaxp1.usc.es [Departamento de Física de Partículas, Universidad de Santiago de Compostela and Instituto Galego de Física de Altas Enerxias (IGFAE), E-15782 Santiago de Compostela (Spain); Naya, C.; Sanchez-Guillen, J.; Vazquez, R. [Departamento de Física de Partículas, Universidad de Santiago de Compostela and Instituto Galego de Física de Altas Enerxias (IGFAE), E-15782 Santiago de Compostela (Spain); Wereszczynski, A. [Institute of Physics, Jagiellonian University, Reymonta 4, Kraków (Poland)

2015-03-06

The BPS Skyrme model has been demonstrated already to provide a physically intriguing and quantitatively reliable description of nuclear matter. Indeed, the model has both the symmetries and the energy–momentum tensor of a perfect fluid, and thus represents a field theoretic realization of the “liquid droplet” model of nuclear matter. In addition, the classical soliton solutions together with some obvious corrections (spin–isospin quantization, Coulomb energy, proton-neutron mass difference) provide an accurate modeling of nuclear binding energies for heavier nuclei. These results lead to the rather natural proposal to try to describe also neutron stars by the BPS Skyrme model coupled to gravity. We find that the resulting self-gravitating BPS Skyrmions provide excellent results as well as some new perspectives for the description of bulk properties of neutron stars when the parameter values of the model are extracted from nuclear physics. Specifically, the maximum possible mass of a neutron star before black-hole formation sets in is a few solar masses, the precise value of which depends on the precise values of the model parameters, and the resulting neutron star radius is of the order of 10 km.

Conceptual design of the early implementation of the NEutron Detector Array (NEDA) with AGATA

Science.gov (United States)

Hüyük, Tayfun; Di Nitto, Antonio; Jaworski, Grzegorz; Gadea, Andrés; Javier Valiente-Dobón, José; Nyberg, Johan; Palacz, Marcin; Söderström, Pär-Anders; Jose Aliaga-Varea, Ramon; de Angelis, Giacomo; Ataç, Ayşe; Collado, Javier; Domingo-Pardo, Cesar; Egea, Francisco Javier; Erduran, Nizamettin; Ertürk, Sefa; de France, Gilles; Gadea, Rafael; González, Vicente; Herrero-Bosch, Vicente; Kaşkaş, Ayşe; Modamio, Victor; Moszynski, Marek; Sanchis, Enrique; Triossi, Andrea; Wadsworth, Robert

2016-03-01

The NEutron Detector Array (NEDA) project aims at the construction of a new high-efficiency compact neutron detector array to be coupled with large γ-ray arrays such as AGATA. The application of NEDA ranges from its use as selective neutron multiplicity filter for fusion-evaporation reaction to a large solid angle neutron tagging device. In the present work, possible configurations for the NEDA coupled with the Neutron Wall for the early implementation with AGATA has been simulated, using Monte Carlo techniques, in order to evaluate their performance figures. The goal of this early NEDA implementation is to improve, with respect to previous instruments, efficiency and capability to select multiplicity for fusion-evaporation reaction channels in which 1, 2 or 3 neutrons are emitted. Each NEDA detector unit has the shape of a regular hexagonal prism with a volume of about 3.23l and it is filled with the EJ301 liquid scintillator, that presents good neutron- γ discrimination properties. The simulations have been performed using a fusion-evaporation event generator that has been validated with a set of experimental data obtained in the 58Ni + 56Fe reaction measured with the Neutron Wall detector array.

Modelling of the Rod Control System in the coupled code RELAP5-QUABOX/CUBBOX

International Nuclear Information System (INIS)

Bencik, V.; Feretic, D.; Grgic, D.

1999-01-01

There is a general agreement that for many light water reactor transient calculations, it is necessary to use a multidimensional neutron kinetics model coupled to sophisticated thermal-hydraulic models in order to obtain satisfactory results. These calculations are needed for a variety of applications for licensing safety analyses, probabilistic risk assessment, operational support, and training. At FER, Zagreb, a coupling of 3D neutronics code QUABOX/CUBBOX and system code RELAP5 was performed. In the paper the Rod Control System model in the RELAP5 part of the coupled code is presented. A first testing of the model was performed by calculation of reactor trip from full power for NPP Krsko. Results of 3D neutronics calculation obtained by coupled code QUABOX/CUBBOX were compared with point kinetics calculation performed with RELAP5 stand alone code.(author)

New statistical model of inelastic fast neutron scattering

International Nuclear Information System (INIS)

Stancicj, V.

1975-07-01

A new statistical model for treating the fast neutron inelastic scattering has been proposed by using the general expressions of the double differential cross section in impuls approximation. The use of the Fermi-Dirac distribution of nucleons makes it possible to derive an analytical expression of the fast neutron inelastic scattering kernel including the angular momenta coupling. The obtained values of the inelastic fast neutron cross section calculated from the derived expression of the scattering kernel are in a good agreement with the experiments. A main advantage of the derived expressions is in their simplicity for the practical calculations

Three-body ΛNN force due to Λ-Σ coupling

International Nuclear Information System (INIS)

Myint, Khin Swe; Akaishi, Yoshinori

2003-01-01

The ΛNN three - body force due to coherent Λ - Σ Coupling effect was derived from realistic Nijmegen model D potential. Repulsive and attractive three - body ΛNN forces were reconcilably accounted. For 5 He, within one - channel description, ΛNN force is largely repulsive and its origin comes from Pauli forbidden terms. Within two - channel description, attractive Pauli allowed terms exist and resulting three - body force is always attractive. Large attractive ΛNN force effect due to coherent Λ - Σ coupling effect is predicted in neutron - rich nuclei. The attractive coherent Λ - Σ coupling effect is largely enhanced at high density neutron matter. The attractive three - body ΛNN force effect is essential dynamics of Λ - Σ coupling while the repulsive Nogami three - body effect arises from Pauli forbidden diagrams. (Y. Kazumata)

Criticality calculations in reactor accelerator coupling experiment (Race)

International Nuclear Information System (INIS)

Reda, M.A.; Spaulding, R.; Hunt, A.; Harmon, J.F.; Beller, D.E.

2005-01-01

A Reactor Accelerator Coupling Experiment (RACE) is to be performed at the Idaho State University Idaho Accelerator Center (IAC). The electron accelerator is used to generate neutrons by inducing Bremsstrahlung photon-neutron reactions in a Tungsten- Copper target. This accelerator/target system produces a source of ∼1012 n/s, which can initiate fission reactions in the subcritical system. This coupling experiment between a 40-MeV electron accelerator and a subcritical system will allow us to predict and measure coupling efficiency, reactivity, and multiplication. In this paper, the results of the criticality and multiplication calculations, which were carried out using the Monte Carlo radiation transport code MCNPX, for different coupling design options are presented. The fuel plate arrangements and the surrounding tank dimensions have been optimized. Criticality using graphite instead of water for reflector/moderator outside of the core region has been studied. The RACE configuration at the IAC will have a criticality (k-effective) of about 0,92 and a multiplication of about 10. (authors)

Comparison of inductively coupled plasma mass spectrometry with inductively coupled plasma atomic emission spectrometry and instrumental neutron activation analysis for the determination of rare earth elements in Greek bauxites

International Nuclear Information System (INIS)

Ochsenkuehn-Petropoulou, Maria; Luck, Joachim

1991-01-01

Fore the determination of rare earth elements (REE) in bauxitic materials the techniques of inductively coupled plasma mass spectrometry (ICP-MS), inductively coupled plasma atomic emission spectrometry (ICP-AES) and instrumental neutron activation analysis (INAA) were compared. In the NIST (National Institute of Standards and Technology) bauxites SRM 697 Dominican, and SRM 69 b Arkansas, the concentration of some REEs were determined. With the reference bauxite BX-N of the ARNT (Association Nationale de la Recherche Technique) the precision and accuracy of ICP-AES for the determination of REEs in bauxites was tested. Furthermore, Greek bauxites of the Parnassos-Giona area were investigated. In a comparison of the three methods it was possible to calculate from the data series the precision of each method, which showed that the tendency found in the deviations for the different REEs is in accordance with published values. Also the limits of detection for REEs in bauxites were calculated and found to be in the same range as those in the literature. (author)

Simulation of LOCA power transients of CANDU6 by SCAN/RELAP-CANDU coupled code system

International Nuclear Information System (INIS)

Hong, In Seob; Kim, Chang Hyo; Hwang, Su Hyun; Kim, Man Woong; Chung, Bub Dong

2004-01-01

As can be seen in the standalone application of RELAP-CANDU for LOCA analysis of CANDU-PHWR, the system thermal-hydraulic code alone cannot predict the transient behavior accurately. Therefore, best estimate neutronics and system thermal-hydraulic coupled code system is necessary to describe the transient behavior with higher accuracy and reliability. The purpose of this research is to develop and test a coupled neutronics and thermal-hydraulics analysis code, SCAN (SNU CANDU-PHWR Neutronics) and RELAP-CANDU, for transient analysis of CANDU-PHWR's. For this purpose, a spatial kinetics calculation module of SCAN, a 3-D CANDU-PHWR neutronics design and analysis code, is dynamically coupled with RELAP-CANDU, the system thermal-hydraulic code for CANDU-PHWR. The performance of the coupled code system is examined by simulation of reactor power transients caused by a hypothetical Loss Of Coolant Accident (LOCA) in Wolsong units, which involves the insertion of positive void reactivity into the core in the course of transients. Specifically, a 40% Reactor Inlet Header (RIH) break LOCA was assumed for the test of the SCAN/RELAP-CANDU coupled code system analysis

Interfacing MCNPX and McStas for simulation of neutron transport

DEFF Research Database (Denmark)

Klinkby, Esben Bryndt; Lauritzen, Bent; Nonbøl, Erik

2013-01-01

Stas[4, 5, 6, 7]. The coupling between the two simulation suites typically consists of providing analytical fits of MCNPX neutron spectra to McStas. This method is generally successful but has limitations, as it e.g. does not allow for re-entry of neutrons into the MCNPX regime. Previous work to resolve......Simulations of target-moderator-reflector system at spallation sources are conventionally carried out using Monte Carlo codes such as MCNPX[1] or FLUKA[2, 3] whereas simulations of neutron transport from the moderator and the instrument response are performed by neutron ray tracing codes such as Mc...... geometries, backgrounds, interference between beam-lines as well as shielding requirements along the neutron guides....

Fast-neutron dosimetry. Progress report, 1 July 1982-30 June 1983

International Nuclear Information System (INIS)

DeLuca, P.M. Jr.; Attix, F.H.; Gould, M.N.

1983-01-01

Several aspects of neutron and related photon radiological physics are being actively investigated. These research topics relate to measurement techniques, basic data values and theoretical discussions. In addition, a modest radiobiological effort is pursued concurrently. The unique coupled neutron/photon source provides an excellent tool for this latter work

Fiber optic neutron imaging system: calibration

International Nuclear Information System (INIS)

Malone, R.M.; Gow, C.E.; Thayer, D.R.

1981-01-01

Two neutron imaging experiments using fiber optics have been performed at the Nevada Test Site. In each experiment, an array of scintillator fluor tubes is exposed to neutrons. Light is coupled out through radiation resistant PCS fibers (8-m long) into high-bandwidth, graded index fibers. For image reconstruction to be accurate, common timing differences and transmission variations between fiber optic channels are needed. The calibration system featured a scanning pulsed dye laser, a specially designed fiber optic star coupler, a tektronix 7912AD transient digitizer, and a DEC PDP 11/34 computing system

Low temperature and neutron physics studies

International Nuclear Information System (INIS)

Shull, C.G.

1989-01-01

A search for a novel coupling interaction between the Pendelloesung periodicity which is formed in a diffracting crystal and the Larmor precession of neutrons in a magnetic field has been carried out. This interaction is expected to exhibit a resonant behavior when the two spatial periodicities become matched upon scanning the magnetic field being applied to the crystal. Observations on a diffracting, perfect crystal of silicon with neutrons of wavelength 1 Angstrom show the expected resonant action but some discrepancy between the observed magnitude of the resonance effects remains for interpretation. 16 refs

Neutron stars in relativistic mean field theory with isovector scalar meson

International Nuclear Information System (INIS)

Kubis, S.; Kutschera, M.; Stachniewicz, S.

1996-12-01

We study the equation of state (EOS) of neutron star matter in a relativistic mean field (RMF) theory with the isovector scalar mean field corresponding to the δ-meson [a 0 (980)]. A range of values of the δ-meson coupling compatible with the Bonn potentials is explored. Parameters of the model in the isovector sector are constrained to fit the nuclear symmetry energy, E s ∼ 30 MeV. We find that proton fraction of neutron star matter is higher in the presence of the δ-field whereas the energy per particle is lower. The EOS becomes slightly stiffer and the maximum mass of the neutron star increased with increasing δmeson coupling. The effect is stronger for soft EOS. (author). 7 refs, 6 figs, 1 tab

Neutron stars in relativistic mean field theory with isovector scalar meson

Energy Technology Data Exchange (ETDEWEB)

Kubis, S.; Kutschera, M.; Stachniewicz, S. [Institute of Nuclear Physics, Cracow (Poland)

1996-12-01

We study the equation of state (EOS) of neutron star matter in a relativistic mean field (RMF) theory with the isovector scalar mean field corresponding to the {delta}-meson [a{sub 0}(980)]. A range of values of the {delta}-meson coupling compatible with the Bonn potentials is explored. Parameters of the model in the isovector sector are constrained to fit the nuclear symmetry energy, E{sub s} {approx} 30 MeV. We find that proton fraction of neutron star matter is higher in the presence of the {delta}-field whereas the energy per particle is lower. The EOS becomes slightly stiffer and the maximum mass of the neutron star increased with increasing {delta}meson coupling. The effect is stronger for soft EOS. (author). 7 refs, 6 figs, 1 tab.

Neutronics and thermal-hydraulics coupling: some contributions toward an improved methodology to simulate the initiating phase of a severe accident in a sodium fast reactor

International Nuclear Information System (INIS)

Guyot, Maxime

2014-01-01

This project is dedicated to the analysis and the quantification of bias corresponding to the computational methodology for simulating the initiating phase of severe accidents on Sodium Fast Reactors. A deterministic approach is carried out to assess the consequences of a severe accident by adopting best estimate design evaluations. An objective of this deterministic approach is to provide guidance to mitigate severe accident developments and re-criticalities through the implementation of adequate design measures. These studies are generally based on modern simulation techniques to test and verify a given design. The new approach developed in this project aims to improve the safety assessment of Sodium Fast Reactors by decreasing the bias related to the deterministic analysis of severe accident scenarios. During the initiating phase, the subassembly wrapper tubes keep their mechanical integrity. Material disruption and dispersal is primarily one-dimensional. For this reason, evaluation methodology for the initiating phase relies on a multiple-channel approach. Typically a channel represents an average pin in a subassembly or a group of similar subassemblies. In the multiple-channel approach, the core thermal-hydraulics model is composed of 1 or 2 D channels. The thermal-hydraulics model is coupled to a neutronics module to provide an estimate of the reactor power level. In this project, a new computational model has been developed to extend the initiating phase modeling. This new model is based on a multi-physics coupling. This model has been applied to obtain information unavailable up to now in regards to neutronics and thermal-hydraulics models and their coupling. (author) [fr

Review of microscopic integral cross section data in fundamental reactor dosimetry benchmark neutron fields

International Nuclear Information System (INIS)

Fabry, A.; McElroy, W.N.; Kellogg, L.S.; Lippincott, E.P.; Grundl, J.A.; Gilliam, D.M.; Hansen, G.E.

1976-01-01

This paper is intended to review and critically discuss microscopic integral cross section measurement and calculation data for fundamental reactor dosimetry benchmark neutron fields. Specifically the review covers the following fundamental benchmarks: the spontaneous californium-252 fission neutron spectrum standard field; the thermal-neutron induced uranium-235 fission neutron spectrum standard field; the (secondary) intermediate-energy standard neutron field at the center of the Mol-ΣΣ, NISUS, and ITN-ΣΣ facilities; the reference neutron field at the center of the Coupled Fast Reactor Measurement Facility; the reference neutron field at the center of the 10% enriched uranium metal, cylindrical, fast critical; the (primary) Intermediate-Energy Standard Neutron Field

Thermal neutron detection using a silicon pad detector and {sup 6}LiF removable converters

Energy Technology Data Exchange (ETDEWEB)

Barbagallo, Massimo [Istituto Nazionale di Fisica Nucleare, Sezione di Bari (Italy); Cosentino, Luigi; Marchetta, Carmelo; Pappalardo, Alfio; Scire, Carlotta; Scire, Sergio; Schillaci, Maria; Vecchio, Gianfranco; Finocchiaro, Paolo [Istituto Nazionale di Fisica Nucleare, Laboratori Nazionali del Sud, Catania (Italy); Forcina, Vittorio; Peerani, Paolo [European Commission, Joint Research Centre, Institute of Transuranium Elements, Ispra (Italy); Vaccaro, Stefano [European Commission, Directorate-General for Energy (Luxembourg)

2013-03-15

A semiconductor detector coupled with a neutron converter is a good candidate for neutron detection, especially for its compactness and reliability if compared with other devices, such as {sup 3}He tubes, even though its intrinsic efficiency is rather lower. In this paper we show a neutron detector design consisting of a 3 cm Multiplication-Sign 3 cm silicon pad detector coupled with one or two external {sup 6}LiF layers, enriched in {sup 6}Li at 95%, placed in contact with the Si active surfaces. This prototype, first characterized and tested at INFN Laboratori Nazionali del Sud and then at JRC Ispra, was successfully shown to detect thermal neutrons with the expected efficiency and an outstanding gamma rejection capability.

Transport of accelerator produced high energy neutrons though concrete

International Nuclear Information System (INIS)

Prabhakar Rao, G.; Sarkar, P.K.

1996-01-01

Development of a computational system for estimating the production and transport of high energy neutrons in particle accelerators is reported. The energy-angle distribution of neutrons from accelerated ions bombarding thick targets is calculated by a hybrid nuclear reaction model code, ALICE-91, modified to suit the purpose. Subsequent transmission of these neutrons through concrete slabs is treated using the anisotropic source-flux iteration technique (ASFIT) in the framework of a coupled neutron-gamma transport. Several parameters of both the codes have been optimized to obtain the transmitted dose through concrete. The calculations are found to be accurate and at the same time faster compared to the detailed Monte Carlo calculations. (author). 8 refs., 2 figs

Concept on coupled spectrum B/T (burning and/or transmutation) reactor for treatment of minor actinides by thermal and fast neutrons

International Nuclear Information System (INIS)

Aziz, Ferhat; Kitamoto, Asashi

1996-01-01

A conceptual design of B/T (burning and/or transmutation) reactor based on a modified conventional 1150 MWe-PWR system, with core consisted of two concentric regions for thermal and fast neutrons, was proposed herein for B/T treatment of MA (minor actinides). The B/T fuel considered was supposed such that MA discharged from 1 GWe-LWR was blended homogeneously with the composition of LWR fuel. In the outer region 23- Np, 241 Am and 243 Am were loaded and burned by thermal neutron, while in the inner region 244 Cm was loaded and burned mainly by fast neutron. The geometry of B/T fuel and the fuel assembly in the outer region was left in the same condition to those of standard PWR while in the inner region the B/T fuel was arranged in the hexagonal geometry, allowed high fuel to coolant volume ratio (V m /V f ), to keep the harder neutron spectrum. Two cases of the Coupled Spectrum B/T Reactor (CSR) with different (V m 1 f ) ratio in the inner region were studied, and the results for the tight lattice with (V m /V f ) = 0.5 showed that those isotopes approached the equilibrium composition after about 5 recycle period, when the CSR was operated under the reactivity swing of 2.8 % dk/k. The evaluations on the void coefficient of reactivity, the Doppler effect and the reactivity swing showed that the CSR concept has the inherent safety and can burn and/or transmute all kind of MA in a single reactor. This CSR can burn about 808 kg of MA in one recycle period of 3 years, which is equivalent to the discharged fuel from about 12 units of LWR in a year. (author)

Entrainment in the inner crust of a neutron star

International Nuclear Information System (INIS)

Chamel, N.

2004-01-01

The inner crust of a neutron star, which is composed of a solid Coulomb lattice of nuclei immersed in a neutron super-fluid, is studied from both a macroscopic and a microscopic level. In the first part, we develop a non-relativistic but 4-dimensionally covariant formulation of the hydrodynamics of a perfect fluid mixture based on a variational principle. This formalism is applied to the description of neutron star crust as 2-fluid model, a neutron super-fluid and a plasma of nuclei and electrons coupled via non dissipative entrainment effects, whose microscopic evaluation is studied in a second part. Applying mean field methods beyond the Wigner-Seitz approximation, the Bragg scattering of dripped neutrons upon crustal nuclei lead to a 'mesoscopic' effective neutron mass, which unlike the 'microscopic' effective mass, takes very large values compared to the bare mass in the middle layers of the crust. (author)

Impact of cross-section generation procedures on the simulation of the VVER 1000 pump startup experiment in the OECD/DOE/CEA V1000CT benchmark by coupled 3-D thermal hydraulics/ neutron kinetics models

International Nuclear Information System (INIS)

Boyan D Ivanov; Kostadin N Ivanov; Sylvie Aniel; Eric Royer

2005-01-01

Full text of publication follows: In the framework of joint effort between the Nuclear Energy Agency (NEA) of OECD, the United States Department of Energy (US DOE), and the Commissariat a l'Energie Atomique (CEA), France a coupled 3-D thermal hydraulics/neutron kinetics benchmark was defined. The overall objective OECD/NEA V1000CT benchmark is to assess computer codes used in analysis of VVER-1000 reactivity transients where mixing phenomena (mass flow and temperature) in the reactor pressure vessel are complex. Original data from the Kozloduy-6 Nuclear Power Plant are available for the validation of computer codes: one experiment of pump start-up (V1000CT-1) and one experiment of steam generator isolation (V1000CT-2). Additional scenarios are defined for code-to-code comparison. As a 3D core model is necessary for a best-estimate computation of all the scenarios of the V1000CT benchmark, all participants were asked to develop their own core coupled 3-D thermal hydraulics/ neutron kinetics models based on the data available in the benchmark specifications. The first code to code comparisons based on the V1000CT-1 Exercise 2 specifications exhibited unacceptable discrepancies between 2 sets of results, one of them being close to experimental results. The present paper focuses first on the analysis of the observed discrepancies. The VVER 1000 3-D thermal hydraulics/neutron kinetics models are based on thermal-hydraulic and neutronic data homogenized at the assembly scale. The neutronic data, provided as part of the benchmark specifications, consist thus in a set of parametrized 2 group cross sections libraries representing the different assemblies and the reflectors. The origin of the high observed discrepancies was found to lie in the use of these neutronic libraries. The concern was then to find a way to provide neutronic data, compatible with all the benchmark participants neutronic models, that enable also comparisons with experimental results. An analysis of the

Mode coupling analysis of coherent quasi-elastic neutron scattering from fluorite-type materials approaching the superionic transition

International Nuclear Information System (INIS)

Chaturvedi, D.K.; Tosi, M.P.

1987-08-01

Neutron scattering experiments on SrCl 2 , CaF 2 and PbF 2 have shown that intensity and width of the coherent diffuse quasi-elastic spectrum increase rapidly with temperature into the fast-ion conducting phase, the main feature in the integrated quasi-elastic intensity being a peak just beyond the (200) point along the (100) direction in scattering wave vector space. The Zwanzig-Mori memory function formalism is used in this work to analyze the quasi-elastic scattering cross section from charge density fluctuations in terms of anharmonic couplings between the vibrational modes of the crystal. The two- and three-mode channels are examined for compatibility with the quasi-elastic neutron scattering evidence, on the basis of (i) energy and momentum conservation and van Hove singularity arguments and (ii) measured phonon dispersion curves along the main symmetry directions in SrCl 2 , CaF 2 , SrF 2 and BaF 2 . The analysis identifies a specific microscopic role for the Raman-active optic branches. The eigenvectors of the relevant Raman-active and partner modes in the three-mode channel describe relative displacements of the two halogens in the unit cell superposed on relative displacements of the halogen and alkaline earth components. This microscopic picture is thus consistent with the superionic transition being associated with the onset of dynamic disorder in the anionic component of the crystal. (author). 13 refs, 2 tabs

Effects of strangeness on the mass-radius of neutron stars in MQMC

International Nuclear Information System (INIS)

Sahoo, H.S.; Mishra, R.N.; Panda, P.K.; Barik, N.

2017-01-01

With the increase of baryon density towards centers of neutron stars, chemical potentials of neutrons become high so that neutrons at Fermi surfaces are changed to hyperons via strangeness non-conserving weak interactions overcoming rest masses of hyperons. In the present attempt we incorporate an additional pair of hidden strange mesons σ∗ and ϕ which couple only to the strange quark and the hyperons of the nuclear matter

Conceptual design of the early implementation of the NEutron Detector Array (NEDA) with AGATA

Energy Technology Data Exchange (ETDEWEB)

Hueyuek, Tayfun; Gadea, Andres; Domingo-Pardo, Cesar [Universidad de Valencia, Instituto de Fisica Corpuscular, CSIC, Paterna (Valencia) (Spain); Di Nitto, Antonio [Istituto Nazionale di Fisica Nucleare, Napoli (Italy); Johannes Gutenberg-Universitaet Mainz, Mainz (Germany); Jaworski, Grzegorz; Javier Valiente-Dobon, Jose; De Angelis, Giacomo; Modamio, Victor; Triossi, Andrea [Laboratori Nazionali di Legnaro, Istituto Nazionale di Fisica Nucleare, Legnaro (Italy); Nyberg, Johan [Uppsala University, Department of Physics and Astronomy, Uppsala (Sweden); Palacz, Marcin [University of Warsaw, Heavy Ion Laboratory, Warsaw (Poland); Soederstroem, Paer-Anders [RIKEN Nishina Center, Saitama (Japan); Aliaga-Varea, Ramon Jose [Universidad de Valencia, Instituto de Fisica Corpuscular, CSIC, Paterna (Valencia) (Spain); Universidad Politecnica de Valencia, I3M, Valencia (Spain); Atac, Ayse [Ankara University, Department of Physics, Faculty of Sciences, Ankara (Turkey); The Royal Institute of Technology, Stockholm (Sweden); Collado, Javier; Egea, Francisco Javier; Gonzalez, Vicente; Sanchis, Enrique [University of Valencia, Department of Electronic Engineering, Burjassot (Valencia) (Spain); Erduran, Nizamettin [Istanbul Sabahattin Zaim University, Faculty of Engineering and Natural Sciences, Istanbul (Turkey); Ertuerk, Sefa [University of Nigde, Department of Physics, Faculty of Science and Arts, Nigde (Turkey); France, Gilles de [CNRS/IN2P3, GANIL, CEA/DSAM, Caen (France); Gadea, Rafael; Herrero-Bosch, Vicente [Universidad Politecnica de Valencia, I3M, Valencia (Spain); Kaskas, Ayse [Ankara University, Department of Physics, Faculty of Sciences, Ankara (Turkey); Moszynski, Marek [National Centre for Nuclear Research, Otwock-Swierk (Poland); Wadsworth, Robert [University of York, Department of Physics, York (United Kingdom)

2016-03-15

The NEutron Detector Array (NEDA) project aims at the construction of a new high-efficiency compact neutron detector array to be coupled with large γ -ray arrays such as AGATA. The application of NEDA ranges from its use as selective neutron multiplicity filter for fusion-evaporation reaction to a large solid angle neutron tagging device. In the present work, possible configurations for the NEDA coupled with the Neutron Wall for the early implementation with AGATA has been simulated, using Monte Carlo techniques, in order to evaluate their performance figures. The goal of this early NEDA implementation is to improve, with respect to previous instruments, efficiency and capability to select multiplicity for fusion-evaporation reaction channels in which 1, 2 or 3 neutrons are emitted. Each NEDA detector unit has the shape of a regular hexagonal prism with a volume of about 3.23l and it is filled with the EJ301 liquid scintillator, that presents good neutron- γ discrimination properties. The simulations have been performed using a fusion-evaporation event generator that has been validated with a set of experimental data obtained in the {sup 58}Ni + {sup 56}Fe reaction measured with the Neutron Wall detector array. (orig.)

TART 2000: A Coupled Neutron-Photon, 3-D, Combinatorial Geometry, Time Dependent, Monte Carlo Transport Code

International Nuclear Information System (INIS)

Cullen, D.E

2000-01-01

TART2000 is a coupled neutron-photon, 3 Dimensional, combinatorial geometry, time dependent Monte Carlo radiation transport code. This code can run on any modern computer. It is a complete system to assist you with input Preparation, running Monte Carlo calculations, and analysis of output results. TART2000 is also incredibly FAST; if you have used similar codes, you will be amazed at how fast this code is compared to other similar codes. Use of the entire system can save you a great deal of time and energy. TART2000 is distributed on CD. This CD contains on-line documentation for all codes included in the system, the codes configured to run on a variety of computers, and many example problems that you can use to familiarize yourself with the system. TART2000 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART2000 and its data files

The OECD/NEA/NSC PBMR coupled neutronics/thermal hydraulics transient benchmark: The PBMR-400 core design

International Nuclear Information System (INIS)

Reitsma, F.; Ivanov, K.; Downar, T.; De Haas, H.; Gougar, H. D.

2006-01-01

The Pebble Bed Modular Reactor (PBMR) is a High-Temperature Gas-cooled Reactor (HTGR) concept to be built in South Africa. As part of the verification and validation program the definition and execution of code-to-code benchmark exercises are important. The Nuclear Energy Agency (NEA) of the Organisation for Economic Cooperation and Development (OECD) has accepted, through the Nuclear Science Committee (NSC), the inclusion of the Pebble-Bed Modular Reactor (PBMR) coupled neutronics/thermal hydraulics transient benchmark problem in its program. The OECD benchmark defines steady-state and transients cases, including reactivity insertion transients. It makes use of a common set of cross sections (to eliminate uncertainties between different codes) and includes specific simplifications to the design to limit the need for participants to introduce approximations in their models. In this paper the detailed specification is explained, including the test cases to be calculated and the results required from participants. (authors)

TART 2000 A Coupled Neutron-Photon, 3-D, Combinatorial Geometry, Time Dependent, Monte Carlo Transport Code

CERN Document Server

Cullen, D

2000-01-01

TART2000 is a coupled neutron-photon, 3 Dimensional, combinatorial geometry, time dependent Monte Carlo radiation transport code. This code can run on any modern computer. It is a complete system to assist you with input Preparation, running Monte Carlo calculations, and analysis of output results. TART2000 is also incredibly FAST; if you have used similar codes, you will be amazed at how fast this code is compared to other similar codes. Use of the entire system can save you a great deal of time and energy. TART2000 is distributed on CD. This CD contains on-line documentation for all codes included in the system, the codes configured to run on a variety of computers, and many example problems that you can use to familiarize yourself with the system. TART2000 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART2000 and its data files.

Measurement of Angular Correlations in the Decay of Polarized Neutrons

DEFF Research Database (Denmark)

Christensen, Carl Jørgen; Krohn, V.E.; Ringo, G.R.

1970-01-01

The electron-momentum-neutron-spin correlation coefficient was found to be A=-0.115±0.008, and the antineutrino-momentum-neutron-spin correlation coefficient was found to be B=1.00±0.05. The value of A leads to |GA/GV|=1.26±0.02 for the ratio of Gamow-Teller-to-Fermi coupling constants in β decay...

Study of proton-induced reactions and correlation with fast-neutron scattering

International Nuclear Information System (INIS)

Hansen, L.F.

1982-01-01

The generation of cross sections for fast neutron-nucleon interactions obtained from elastic and charge-exchange proton data is discussed in terms of the Lane model formalism. A general description of the interaction of nucleons with nuclei is presented in terms of the optical model and the extended (or coupled-channel) optical model, together with the relation of these models to microscopic calculations of the nucleon-nucleon interaction. Comparisons between neutron elastic data and calculations carried out with optical model potentials obtained from (p,p) and (p,n) data are presented for a large number of nuclei. The validity of the Lane model and the importance of coupled effects in the actinide region are shown in a detailed comparison of calculations for elastic and inelastic neutron differential cross sections and measurements for 232 Th and 238 U

Coupled fast-thermal system at the RB, masters thesis; Spregnuti brzo-termicki sistem na reaktoru RB, magistarski rad

Energy Technology Data Exchange (ETDEWEB)

Pesic, M [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

1984-05-15

Coupled fast-thermal system at the RB reactor was formed owing to availability of highly enriched fuel. This paper deals with reactor parameters calculations and measurements of coupled core taking into account safety constraints. Validity of applied calculation methods was confirmed. The following parameters were analyzed: critical height of the core; reactivity dependent on heavy water level in the core; fast neutron spectrum in the fast region channel; spatial distribution of thermal. epithermal and fat neutrons in the fast region channel; reactivity of safety rods; neutron and gamma absorption doses in the center of the coupled core.

Polyethylene-reflected plutonium metal sphere : subcritical neutron and gamma measurements.

Energy Technology Data Exchange (ETDEWEB)

Mattingly, John K.

2009-11-01

Numerous benchmark measurements have been performed to enable developers of neutron transport models and codes to evaluate the accuracy of their calculations. In particular, for criticality safety applications, the International Criticality Safety Benchmark Experiment Program (ICSBEP) annually publishes a handbook of critical and subcritical benchmarks. Relatively fewer benchmark measurements have been performed to validate photon transport models and codes, and unlike the ICSBEP, there is no program dedicated to the evaluation and publication of photon benchmarks. Even fewer coupled neutron-photon benchmarks have been performed. This report documents a coupled neutron-photon benchmark for plutonium metal reflected by polyethylene. A 4.5-kg sphere of ?-phase, weapons-grade plutonium metal was measured in six reflected configurations: (1) Bare; (2) Reflected by 0.5 inch of high density polyethylene (HDPE); (3) Reflected by 1.0 inch of HDPE; (4) Reflected by 1.5 inches of HDPE; (5) Reflected by 3.0 inches of HDPE; and (6) Reflected by 6.0 inches of HDPE. Neutron and photon emissions from the plutonium sphere were measured using three instruments: (1) A gross neutron counter; (2) A neutron multiplicity counter; and (3) A high-resolution gamma spectrometer. This report documents the experimental conditions and results in detail sufficient to permit developers of radiation transport models and codes to construct models of the experiments and to compare their calculations to the measurements. All of the data acquired during this series of experiments are available upon request.

Polyethylene-reflected plutonium metal sphere: subcritical neutron and gamma measurements

International Nuclear Information System (INIS)

Mattingly, John K.

2009-01-01

Numerous benchmark measurements have been performed to enable developers of neutron transport models and codes to evaluate the accuracy of their calculations. In particular, for criticality safety applications, the International Criticality Safety Benchmark Experiment Program (ICSBEP) annually publishes a handbook of critical and subcritical benchmarks. Relatively fewer benchmark measurements have been performed to validate photon transport models and codes, and unlike the ICSBEP, there is no program dedicated to the evaluation and publication of photon benchmarks. Even fewer coupled neutron-photon benchmarks have been performed. This report documents a coupled neutron-photon benchmark for plutonium metal reflected by polyethylene. A 4.5-kg sphere of ?-phase, weapons-grade plutonium metal was measured in six reflected configurations: (1) Bare; (2) Reflected by 0.5 inch of high density polyethylene (HDPE); (3) Reflected by 1.0 inch of HDPE; (4) Reflected by 1.5 inches of HDPE; (5) Reflected by 3.0 inches of HDPE; and (6) Reflected by 6.0 inches of HDPE. Neutron and photon emissions from the plutonium sphere were measured using three instruments: (1) A gross neutron counter; (2) A neutron multiplicity counter; and (3) A high-resolution gamma spectrometer. This report documents the experimental conditions and results in detail sufficient to permit developers of radiation transport models and codes to construct models of the experiments and to compare their calculations to the measurements. All of the data acquired during this series of experiments are available upon request.

A two-dimensional simulator of the neutronic behaviour of low power fast reactors

International Nuclear Information System (INIS)

Penha, M.A.V.R. da.

1984-01-01

A model to simulate the temporal neutronic behaviour of fast breeder reactors was developed. The effective cross-sections are corrected, whenever the reactor state change; by using linear correlations and interpolation schemes with data contained in a library previously compiled. This methodology was coupled with a simplified spatial neutronic calculation to investigate the temporal behaviour of neutronic parameters such as breeding gain, flux and power. (Author) [pt

Study of multi-neutron emission in the $\\beta$-decay of $^{11}$Li

CERN Multimedia

A new investigation of neutron emission in the $\\beta$-decay of $^{11}$Li is proposed. The principal goal of this study will be to directly measure, for the first time for any system, two $\\beta$-delayed neutrons in coincidence and determine the energy and angular correlations. This will be possible using liquid scintillator detectors, capable of distinguishing between neutrons and ambient $\\gamma$ and cosmic-rays, coupled to a new digital electronics and acquisition system. In parallel, a considerably more refined picture of the single-neutron emission will be obtained.

Model of superdense matter and its application to neutron stars

International Nuclear Information System (INIS)

Pedico, R.D.

1976-01-01

A phenomenological model of superdense baryonic matter at zero temperature is developed and the resulting equation of state is employed in the calculation of neutron star masses and moments of inertia. The strong interactions between the baryons are described by couplings to one scalar and one vector field. These fields are not identified with observed mesons. Only a particular class of diagrams, constructed from tadpole terms, is retained in this investigation. It is argued that these terms contain the leading order density dependence of any set of diagrams that can be built up from fundamental two baryon-one meson vertices. The two parameters in the model, the coupling strengths, are fixed by the requirement that the accepted binding energy of infinite nuclear matter be reproduced at nuclear density. These couplings are used to calculate a forward proton-neutron cross section, which is found to agree with experimental data over a limited energy range. A pressure-energy density equation of state is generated for an electrically neutral system of electrons, muons, and the lowest mass baryon octet. The constituents are held in chemical equilibrium by the weak interactions. The equation of state exhibits a broad phase transition encompassing nuclear density, which leads to neutron stars containing a nearly incompressible core surrounded by a significantly less dense shell. The masses and moments of inertia of these model neutron stars are in good agreement with observational data for pulsars

Prominent Role of Spin-Orbit Coupling in FeSe Revealed by Inelastic Neutron Scattering

Directory of Open Access Journals (Sweden)

Mingwei Ma

2017-05-01

Full Text Available In most existing theories for iron-based superconductors, spin-orbit coupling (SOC has been assumed to be insignificant. Here, we use spin-polarized inelastic neutron scattering to show that collective low-energy spin excitations in the orthorhombic (or “nematic” phase of FeSe possess nearly no in-plane component. Such spin-space anisotropy is present over an energy range greater than the superconducting gap 2Δ_{sc} and gets fully inherited in the superconducting state, resulting in a c-axis polarized “spin resonance” without any noticeable isotropic spectral-weight rearrangement related to the superconductivity, which is distinct from observations in the superconducting iron pnictides. The contrast between the strong suppression of long-range magnetic order in FeSe and the persisting large spin-space anisotropy, which cannot be explained microscopically by introducing single-ion anisotropy into local-moment spin models, demonstrates the importance of SOC in an itinerant-electron description of the low-energy spin excitations. Our result helps to elucidate the nearby magnetic instabilities and the debated interplay between spin and orbital degrees of freedom in FeSe. The prominent role of SOC also implies a possible unusual nature of the superconducting state.

Review of microscopic integral cross section data in fundamental reactor dosimetry benchmark neutron fields

International Nuclear Information System (INIS)

Fabry, A.; McElroy, W.N.; Kellogg, L.S.; Lippincott, E.P.; Grundl, J.A.; Gilliam, D.M.; Hansen, G.E.

1976-10-01

The paper is intended to review and critically discuss microscopic integral cross section measurement and calculation data for fundamental reactor dosimetry benchmark neutron fields. Specifically the review covers the following fundamental benchmarks: (1) the spontaneous californium-252 fission neutron spectrum standard field; (2) the thermal-neutron induced uranium-235 fission neutron spectrum standard field; (3) the (secondary) intermediate-energy standard neutron field at the center of the Mol-ΣΣ, NISUS, and ITN--ΣΣ facilities; (4) the reference neutron field at the center of the Coupled Fast Reactor Measurement Facility (CFRMF); (5) the reference neutron field at the center of the 10 percent enriched uranium metal, cylindrical, fast critical; and (6) the (primary) Intermediate-Energy Standard Neutron Field

MODICO, 1-D Time-Dependent 1 Group, 2 Group Neutron Diffusion with Delayed Neutron Precursors

International Nuclear Information System (INIS)

Camiciola, P.; Cundari, D.; Montagnini, B.

1992-01-01

1 - Description of program or function: The program solves the 1-D time-dependent one and two group coarse-mesh neutron diffusion equations, coupled with the equations for the delayed-neutron precursor, in plane geometry. 2 - Method of solution: The program is based on a simple coarse-mesh cubic approximation formula for the spatial behaviour of the flux inside each interval. An implicit scheme (the time-integrated method) is used for the advancement of the solution. The resulting (block three-diagonal) matrix is inverted at each time step by Thomas' method. 3 - Restrictions on the complexity of the problem: Number of coarse- mesh intervals LE 80; number of material regions LE 10; number of delayed-neutron precursor groups LE 10. Typical mesh sizes range from 5 cm to 20 cm; typical step length (non-prompt critical transients) ranges from 0.005 to 0.1 seconds

Neutron noise measurement technique in a coupled reactor

International Nuclear Information System (INIS)

Genoud, J.P.

1976-01-01

Describes work carried out on the swimming pool reactor at the Physikalisch-Technische Bundesanstalt at Braunschweig. The reactor has two multiplying zones, is light water moderated, with 90% enriched 235 U fuel. There is a D 2 0 reservoir between the two parts of the reactor. Signal/noise ratio obtained by means of ionisation chamber type neutron detectors of 10 -13 amp/u.f. sensitivity is of the order of 40 dB and band frequency 1.5 kHz. Spectral density of the interzone interaction energy was obtained by use of Fourier transforms, previously corrected by a Hanning window. (S.W.)

TITAN: an advanced three-dimensional coupled neutronic/thermal-hydraulics code for light water nuclear reactor core analysis

International Nuclear Information System (INIS)

Griggs, D.P.; Kazimi, M.S.; Henry, A.F.

1984-06-01

The three-dimensional nodal neutronics code QUANDRY and the three-dimensional two-fluid thermal-hydraulics code THERMIT are combined into TITAN. Steady-state and transient coupling methodologies based upon a tandem structure were devised and implemented. Additional models for nuclear feedback, equilibrium xenon and direct moderator heating were added. TITAN was tested using a boiling water two channel problem and the coupling methodologies were shown to be effective. Simulated turbine trip transients and several control rod withdrawal transients were analyzed with good results. Sensitivity studies indicated that the time-step size can affect transient results significantly. TITAN was also applied to a quarter core PWR problem based on a real reactor geometry. The steady-state results were compared to a solution produced by MEKIN-B and poor agreement between the horizontal power shapes was found. Calculations with various mesh spacings showed that the mesh spacings in the MEKIN-B analysis were too large to produce accurate results with a finite difference method. The TITAN results were shown to be reasonable. A pair of control rod ejection accidents were also analyzed with TITAN. A comparison of the TITAN PWR control rod ejection results with results from coupled point kinetics/thermal-hydraulics analyses showed that the point kinetics method used (adiabatic method for control rod reactivities, steady-state flux shape for core-averaged reactivity feedback) underpredicted the power excursion in one case and overpredicted it in the other. It was therefore concluded that point kinetics methods should be used with caution and that three-dimensional codes like TITAN are superior for analyzing PWR control rod ejection transients

Two-component Superfluid Hydrodynamics of Neutron Star Cores

Energy Technology Data Exchange (ETDEWEB)

Kobyakov, D. N. [Institute of Applied Physics of the Russian Academy of Sciences, 603950 Nizhny Novgorod (Russian Federation); Pethick, C. J., E-mail: dmitry.kobyakov@appl.sci-nnov.ru, E-mail: pethick@nbi.dk [The Niels Bohr International Academy, The Niels Bohr Institute, University of Copenhagen, Blegdamsvej 17, DK-2100 Copenhagen Ø (Denmark)

2017-02-20

We consider the hydrodynamics of the outer core of a neutron star under conditions when both neutrons and protons are superfluid. Starting from the equation of motion for the phases of the wave functions of the condensates of neutron pairs and proton pairs, we derive the generalization of the Euler equation for a one-component fluid. These equations are supplemented by the conditions for conservation of neutron number and proton number. Of particular interest is the effect of entrainment, the fact that the current of one nucleon species depends on the momenta per nucleon of both condensates. We find that the nonlinear terms in the Euler-like equation contain contributions that have not always been taken into account in previous applications of superfluid hydrodynamics. We apply the formalism to determine the frequency of oscillations about a state with stationary condensates and states with a spatially uniform counterflow of neutrons and protons. The velocities of the coupled sound-like modes of neutrons and protons are calculated from properties of uniform neutron star matter evaluated on the basis of chiral effective field theory. We also derive the condition for the two-stream instability to occur.

A fully coupled Monte Carlo/discrete ordinates solution to the neutron transport equation. Final report

Energy Technology Data Exchange (ETDEWEB)

Baker, Randal Scott [Univ. of Arizona, Tucson, AZ (United States)

1990-01-01

The neutron transport equation is solved by a hybrid method that iteratively couples regions where deterministic (S_N) and stochastic (Monte Carlo) methods are applied. Unlike previous hybrid methods, the Monte Carlo and S_N regions are fully coupled in the sense that no assumption is made about geometrical separation or decoupling. The hybrid method provides a new means of solving problems involving both optically thick and optically thin regions that neither Monte Carlo nor S_N is well suited for by themselves. The fully coupled Monte Carlo/S_N technique consists of defining spatial and/or energy regions of a problem in which either a Monte Carlo calculation or an S_N calculation is to be performed. The Monte Carlo region may comprise the entire spatial region for selected energy groups, or may consist of a rectangular area that is either completely or partially embedded in an arbitrary S_N region. The Monte Carlo and S_N regions are then connected through the common angular boundary fluxes, which are determined iteratively using the response matrix technique, and volumetric sources. The hybrid method has been implemented in the S_N code TWODANT by adding special-purpose Monte Carlo subroutines to calculate the response matrices and volumetric sources, and linkage subrountines to carry out the interface flux iterations. The common angular boundary fluxes are included in the S_N code as interior boundary sources, leaving the logic for the solution of the transport flux unchanged, while, with minor modifications, the diffusion synthetic accelerator remains effective in accelerating S_N calculations. The special-purpose Monte Carlo routines used are essentially analog, with few variance reduction techniques employed. However, the routines have been successfully vectorized, with approximately a factor of five increase in speed over the non-vectorized version.

Non-Destructive Spent Fuel Characterization with Semiconducting Gallium Arsinde Neutron Imaging Arrays

International Nuclear Information System (INIS)

McGregor, Douglas S.; Gersch, Holly K.; Sanders, Jeffrey D.; Lee, John C.; Hammig, Mark D.; Hartman, Michael R.; Yong Hong Yang; Klann, Raymond T.; Elzen, Brian Van Der; Lindsay, John T.; Simpson, Philip A.

2002-01-01

High resistivity bulk grown GaAs has been used to produce thermal neutron imaging devices for use in neutron radiography and characterizing burnup in spent fuel. The basic scheme utilizes a portable Sb/Be source for monoenergetic (24 keV) neutron radiation source coupled to an Fe filter with a radiation hard B-coated pixellated GaAs detector array as the primary neutron detector. The coated neutron detectors have been tested for efficiency and radiation hardness in order to determine their fitness for the harsh environments imposed by spent fuel. Theoretical and experimental results are presented, showing detector radiation hardness, expected detection efficiency and the spatial resolution from such a scheme. A variety of advanced neutron detector designs have been explored, with experimental results achieving 13% thermal neutron detection efficiency while projecting the possibility of over 30% thermal neutron detection efficiency

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 spin-spin effect in the total neutron cross section of polarized neutrons on polarized 165Ho

International Nuclear Information System (INIS)

Fasoli, U.; Galeazzi, G.; Pavan, P.; Toniolo, D.; Zago, G.; Zannoni, R.

1978-01-01

The spin-spin effect in the total neutron cross section of polarized neutrons on polarized 165 Ho has been measured in the energy interval 0.4 to 2.5 MeV, in perpendicular geometry. The results are consistent with zero effect. The spin-spin cross section sigmasub(ss) has been theoretically evaluated by a non-adiabatic coupled-channel calculation. From the comparison between the experimental and theoretical results a value Vsub(ss) = 9+-77 keV for the strength of the spin-spin potential has been obtained. Compound-nucleus effects do not seem to be relevant. (Auth.)

Measurement of epithermal neutrons by a coherent demodulation technique

CERN Document Server

Horiuchi, N; Takahashi, H; Kobayashi, H; Harasawa, S

2000-01-01

Epithermal neutrons have been measured using a neutron dosimeter via a coherent demodulation technique. This dosimeter consists of CsI(Tl)-photodiode scintillation detectors, four of which are coupled to neutron-gamma converting foils of various sizes. Neutron-gamma converting foils of In, Au and Co materials were used, each of which has a large capture cross section which peaks in the epithermal neutron energy region. The type of foil was selected according to the material properties that best correspond to the energy of the epithermal neutrons to be measured. In addition, the proposed technique was applied using Au-foils in order to measure the Cd ratio. The validity of the proposed technique was examined using an sup 2 sup 4 sup 1 Am-Be source placed in a testing stack of polyethylene blocks, and the results were compared with the theoretical values calculated by the Monte Carlo calculation. Finally, the dosimeter was applied for measuring epithermal neutrons and the Cd ratio in an experimental beam-tube o...

Coupling of THALES and FROST using MPI Method

International Nuclear Information System (INIS)

Park, Jin Woo; Ryu, Seok Hee; Jung, Chan Do; Jung, Jee Hoon; Um, Kil Sup; Lee, Jae Il

2013-01-01

This paper presents the coupling method between THALES and FROST and the simulation results with the coupled code system. In this study, subchannel analysis code THALES and transient fuel performance code FROST were coupled using MPI method as the first stage of the development of the multi-dimensional safety analysis methodology. As a part of the validation, the CEA ejection accident was simulated using the coupled THALES-FROST code and the results were compared with the ShinKori 3 and 4 FSAR. Comparison results revealed that CHASER using MPI method predicts fuel temperatures and heat flux quantitatively well. Thus it was confirmed that the THALES and FROST are properly coupled. In near future, ASTRA, multi-dimensional core neutron kinetics code, will be linked to THALESFROST code for the detailed three-dimensional CEA ejection analysis. The current safety analysis methodology for a CEA ejection accident based on numerous conservative assumptions with the point kinetics model results in quite adverse consequences. Thus, KNF is developing the multi-dimensional safety analysis methodology to enhance the consequences of the CEA ejection accident. For this purpose, three-dimensional core neutron kinetics code ASTRA, subchannel analysis code THALES, and transient fuel performance analysis code FROST are being coupled using message passing interface(MPI). For the first step, THALES and FROST are coupled and tested

Comparative study of the Peach Bottom turbine trip experiment using two different coupled codes approaches

International Nuclear Information System (INIS)

Bambara, M.; Bousbia-Salah, A.; D'Auria, F.

2005-01-01

Full text of publication follows: In the last years a great concern about the neutron-3D/thermal-hydraulic codes coupling took place. Owing to the improved computational technology, 'best estimate' analyses are today a common tool to assess safety features, and they are necessary if an asymmetric behaviour in the core region exists, or if strong interactions between the core neutronics and reactor thermal-hydraulic occur. In order to validate the coupled codes performances, several international programmes were issued. Among these activities, the OECD/NEA BWR Turbine Trip (TT) was chosen for further sensitivity analyses. It consists of a turbine trip (TT) experiment carried out at the Peach Bottom 2 BWR. In this paper, the results of two different coupled codes systems are summarized and compared. The BWR TT simulations were carried out coupling the thermal-hydraulic system code RELAP5/mode 3.2 to the 3D neutron kinetics code Parcs/2.3, and also the system code ATHLET to the neutronics code QUABOX-CUBBOX. An exhaustive overview of the main features is given, and those aspects, which need further developments and experiences, are pointed out. (authors)

Description of a stable scheme for steady-state coupled Monte Carlo–thermal–hydraulic calculations

International Nuclear Information System (INIS)

Dufek, Jan; Eduard Hoogenboom, J.

2014-01-01

Highlights: • A stable coupling scheme for steady-state MC–TH calculations is described. • The coupling scheme is based on the stochastic approximation method. • The neutron flux (or power) distribution is relaxed using a variable step-size. - Abstract: We provide a detailed description of a numerically stable and efficient coupling scheme for steady-state Monte Carlo neutronic calculations with thermal–hydraulic feedback. While we have previously derived and published the stochastic approximation based method for coupling the Monte Carlo criticality and thermal–hydraulic calculations, its possible implementation has not been described in a step-by-step manner. As the simple description of the coupling scheme was repeatedly requested from us, we have decided to make it available via this note

A directional fast neutron detector using scintillating fibers and an intensified CCD camera system

International Nuclear Information System (INIS)

Holslin, Daniel; Armstrong, A.W.; Hagan, William; Shreve, David; Smith, Scott

1994-01-01

We have been developing and testing a scintillating fiber detector (SFD) for use as a fast neutron sensor which can discriminate against neutrons entering at angles non-parallel to the fiber axis (''directionality''). The detector/convertor component is a fiber bundle constructed of plastic scintillating fibers each measuring 10 cm long and either 0.3 mm or 0.5 mm in diameter. Extensive Monte Carlo simulations were made to optimize the bundle response to a range of fast neutron energies and to intense fluxes of high energy gamma-rays. The bundle is coupled to a set of gamma-ray insenitive electro-optic intensifiers whose output is viewed by a CCD camera directly coupled to the intensifiers. Two types of CCD cameras were utilized: 1) a standard, interline RS-170 camera with electronic shuttering and 2) a high-speed (up to 850 frame/s) field-transfer camera. Measurements of the neutron detection efficiency and directionality were made using 14 MeV neutrons, and the response to gamma-rays was performed using intense fluxes from radioisotopic sources (up to 20 R/h). Recently, the detector was constructed and tested using a large 10 cm by 10 cm square fiber bundle coupled to a 10 cm diameter GEN I intensifier tube. We present a description of the various detector systems and report the results of experimental tests. ((orig.))

Temperature measurement with neutrons

International Nuclear Information System (INIS)

Bizard, G.; Durand, D.; Lecolley, J.F.; Lefebvres, F.; Marques, M.; Peter, J.; Tamain, B.

1998-01-01

The results presented in this report were obtained from the information provided by charged products. Another alternative consists in detecting the neutrons abundantly emitted particularly by heavy nuclei. The residue channel was studied in the 40 Ar + 197 Au at 60 MeV/nucleon by means of the neutron multidetector DEMON. The evolution of the multiplicity of neutrons emitted backwards in the framework of the heavy nucleus forwardly detected as a function of the residue velocity by a silicon detector, placed at 8 degrees and at 24.5 cm from target, agrees with the expected results i.e. an increase with the residue velocity hence with the collision violence. For the same detector the first measurements show similarly a linear increase of the apparent temperature of 4.0 to around 6.5 MeV for residue velocities varying from 0.5 to 1.3 cm/ns and masses ranging from 140 to 160 uma. This first results of the analysis show therefore a good behaviour of the assembly and especially of the couple DeMoN-SyReP

Neutron resonance transmission spectroscopy with high spatial and energy resolution at the J-PARC pulsed neutron source

Energy Technology Data Exchange (ETDEWEB)

Tremsin, A.S., E-mail: ast@ssl.berkeley.edu [University of California at Berkeley, 7 Gauss Way, Berkeley, CA 94720 (United States); Shinohara, T.; Kai, T.; Ooi, M. [Japan Atomic Energy Agency, 2–4 Shirakata-shirane, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Kamiyama, T.; Kiyanagi, Y.; Shiota, Y. [Hokkaido University, Kita 13 Nishi 8 Kita-ku, Sapporo-shi, Hokkaido 060-8628 (Japan); McPhate, J.B.; Vallerga, J.V.; Siegmund, O.H.W. [University of California at Berkeley, 7 Gauss Way, Berkeley, CA 94720 (United States); Feller, W.B. [NOVA Scientific, Inc., 10 Picker Rd., Sturbridge, MA 01566 (United States)

2014-05-11

The sharp variation of neutron attenuation at certain energies specific to particular nuclides (the lower range being from ∼1 eV up to ∼1 keV), can be exploited for the remote mapping of element and/or isotope distributions, as well as temperature probing, within relatively thick samples. Intense pulsed neutron beam-lines at spallation sources combined with a high spatial, high-timing resolution neutron counting detector, provide a unique opportunity to measure neutron transmission spectra through the time-of-flight technique. We present the results of experiments where spatially resolved neutron resonances were measured, at energies up to 50 keV. These experiments were performed with the intense flux low background NOBORU neutron beamline at the J-PARC neutron source and the high timing resolution (∼20 ns at epithermal neutron energies) and spatial resolution (∼55 µm) neutron counting detector using microchannel plates coupled to a Timepix electronic readout. Simultaneous element-specific imaging was carried out for several materials, at a spatial resolution of ∼150 µm. The high timing resolution of our detector combined with the low background beamline, also enabled characterization of the neutron pulse itself – specifically its pulse width, which varies with neutron energy. The results of our measurements are in good agreement with the predicted results for the double pulse structure of the J-PARC facility, which provides two 100 ns-wide proton pulses separated by 600 ns, broadened by the neutron energy moderation process. Thermal neutron radiography can be conducted simultaneously with resonance transmission spectroscopy, and can reveal the internal structure of the samples. The transmission spectra measured in our experiments demonstrate the feasibility of mapping elemental distributions using this non-destructive technique, for those elements (and in certain cases, specific isotopes), which have resonance energies below a few keV, and with lower

Study of neutron spectrometers for ITER

Energy Technology Data Exchange (ETDEWEB)

Kaellne, Jan

2005-11-15

A review is presented of the developments in the field of neutron emission spectrometry (NES) which is of relevance for identifying the role of NES diagnostics on ITER and selecting suitable instrumentation. Neutron spectrometers will be part of the ITER neutron diagnostic complement and this study makes a special effort to examine which performance characteristics the spectrometers should possess to provide the best burning plasma diagnostic information together with neutron cameras and neutron yield monitors. The performance of NES diagnostics is coupled to how much interface space can be provided which has lead to an interest to find compact instruments and their NES capabilities. This study assesses all known spectrometer types of potential interest for ITER and makes a ranking of their performance (as demonstrated or projected), which, in turn, are compared with ITER measurement requirements as a reference; the ratio of diagnostic performance to interface cost for different spectrometers is also discussed for different spectrometer types. The overall result of the study is an assessment of which diagnostic functions neutron measurements can provide in burning plasma fusion experiments on ITER and the role that NES can play depending on the category of instrument installed. Of special note is the result that much higher quality diagnostic information can be obtained from neutron measurements with total yield monitors, profile flux cameras and spectrometers when the synergy in the data is considered in the analysis and interpretation.

Application of the coupled Relap5/Panther codes for PWR steam. Line break accident analysis

International Nuclear Information System (INIS)

Guisset, J.-P.; Bosso, S.; Charlier, A.; Delhaye, X.; Ergo, O.; Ouliddren, K.; Schneidesch, C.; Zhang, J.

2001-01-01

A dynamic coupling between the existing 1-dimensional thermal-hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER is applied via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel thermal-hydraulic analysis code COBRA 3C allows direct evaluation of the Departure from Nucleate Boiling Ratio in parallel with the coupled PANTHER/RELAP5 simulation. The coupled codes are applied to develop a Final Safety Analysis Report (FSAR) accident analysis methodology for the major Steam Line Break (SLB) accident at hot zero power in a typical three-loop pressurised water reactor. In this methodology, the uncertainties related to the plant, core thermal-hydraulic and neutronic parameters are combined in a deterministic bounding approach based on sensitivity studies. The results of coupled thermal-hydraulic and neutronic analysis of SLB are presented and discussed. It is shown that there exists an important margin in the traditional FSAR accident analysis for SLB, which can be attributed by the conservatism's introduced by de-coupling the plant sub-systems. (author)

Application of the coupled Relap5/Panther codes for PWR steam. Line break accident analysis

Energy Technology Data Exchange (ETDEWEB)

Guisset, J.-P.; Bosso, S.; Charlier, A.; Delhaye, X.; Ergo, O.; Ouliddren, K.; Schneidesch, C.; Zhang, J. [Tractebel Energy Engineering, Brussels (Belgium)

2001-07-01

A dynamic coupling between the existing 1-dimensional thermal-hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER is applied via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel thermal-hydraulic analysis code COBRA 3C allows direct evaluation of the Departure from Nucleate Boiling Ratio in parallel with the coupled PANTHER/RELAP5 simulation. The coupled codes are applied to develop a Final Safety Analysis Report (FSAR) accident analysis methodology for the major Steam Line Break (SLB) accident at hot zero power in a typical three-loop pressurised water reactor. In this methodology, the uncertainties related to the plant, core thermal-hydraulic and neutronic parameters are combined in a deterministic bounding approach based on sensitivity studies. The results of coupled thermal-hydraulic and neutronic analysis of SLB are presented and discussed. It is shown that there exists an important margin in the traditional FSAR accident analysis for SLB, which can be attributed by the conservatism's introduced by de-coupling the plant sub-systems. (author)

Generalization of the Fourier Convergence Analysis in the Neutron Diffusion Eigenvalue Problem

International Nuclear Information System (INIS)

Lee, Hyun Chul; Noh, Jae Man; Joo, Hyung Kook

2005-01-01

Fourier error analysis has been a standard technique for the stability and convergence analysis of linear and nonlinear iterative methods. Lee et al proposed new 2- D/1-D coupling methods and demonstrated several advantages of the new methods by performing a Fourier convergence analysis of the methods as well as two existing methods for a fixed source problem. We demonstrated the Fourier convergence analysis of one of the 2-D/1-D coupling methods applied to a neutron diffusion eigenvalue problem. However, the technique cannot be used directly to analyze the convergence of the other 2-D/1-D coupling methods since some algorithm-specific features were used in our previous study. In this paper we generalized the Fourier convergence analysis technique proposed and analyzed the convergence of the 2-D/1-D coupling methods applied to a neutron diffusion Eigenvalue problem using the generalized technique

Burn-up measurements coupling gamma spectrometry and neutron measurement

Energy Technology Data Exchange (ETDEWEB)

Toubon, H.; Pin, P. [AREVA/CANBERRA, 1 rue des Herons, 78182 St Quentin-en-Yvelines Cedex (France); Lebrun, A. [IAEA, Wagramer Strasse 5, PO Box 100, Vienna (Austria); Oriol, L.; Saurel, N. [CEA Cadarache, 13108 Saint Paul Lez Durance Cedex (France); Gain, T. [AREVA/COGEMA Reprocessing Business Unit, La Hague, 50444 Beaumont Hague Cedex (France)

2006-07-01

The need to apply for burn-up credit arises with the increase of the initial enrichment of nuclear fuel. When burn-up credit is used in criticality safety studies, it is often necessary to confirm it by measurement. For the last 10 years, CANBERRA has manufactured the PYTHON system for such measurements. However, the method used in the PYTHON itself uses certain reactor data to arrive at burn-up estimates. Based on R and D led by CEA and COGEMA in the framework of burn-up measurement for burn-up credit and safeguards applications, CANBERRA is developing the next generation of burn-up measurement device. This new product, named SMOPY, is able to measure burn-up of any kind of irradiated fuel assembly with a combination of gamma spectrometry and passive neutron measurements. The measurement data is used as input to the CESAR depletion code, which has been developed and qualified by CEA and COGEMA for burn-up credit determinations. In this paper, we explain the complementary nature of the gamma and neutron measurements. In addition, we draw on our previous experience from PYTHON system and from COGEMA La Hague to show what types of evaluations are required to qualify the SMOPY system, to estimate its uncertainties, and to detect discrepancies in the fuel data given by the reactor plant to characterize the irradiated fuel assembly. (authors)

Burn-up measurements coupling gamma spectrometry and neutron measurement

International Nuclear Information System (INIS)

Toubon, H.; Pin, P.; Lebrun, A.; Oriol, L.; Saurel, N.; Gain, T.

2006-01-01

The need to apply for burn-up credit arises with the increase of the initial enrichment of nuclear fuel. When burn-up credit is used in criticality safety studies, it is often necessary to confirm it by measurement. For the last 10 years, CANBERRA has manufactured the PYTHON system for such measurements. However, the method used in the PYTHON itself uses certain reactor data to arrive at burn-up estimates. Based on R and D led by CEA and COGEMA in the framework of burn-up measurement for burn-up credit and safeguards applications, CANBERRA is developing the next generation of burn-up measurement device. This new product, named SMOPY, is able to measure burn-up of any kind of irradiated fuel assembly with a combination of gamma spectrometry and passive neutron measurements. The measurement data is used as input to the CESAR depletion code, which has been developed and qualified by CEA and COGEMA for burn-up credit determinations. In this paper, we explain the complementary nature of the gamma and neutron measurements. In addition, we draw on our previous experience from PYTHON system and from COGEMA La Hague to show what types of evaluations are required to qualify the SMOPY system, to estimate its uncertainties, and to detect discrepancies in the fuel data given by the reactor plant to characterize the irradiated fuel assembly. (authors)

Design of a system for neutrons dosimetry; Diseno de un sistema para dosimetria de neutrones

Energy Technology Data Exchange (ETDEWEB)

Ceron, P.; Rivera, T. [IPN, Centro de Investigacion en Ciencia Aplicada y Tecnologia Avanzada, Legaria No. 694, Col. Irrigacion, 11500 Mexico D. F. (Mexico); Paredes G, L. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Azorin, J. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico D. F. (Mexico); Sanchez, A. [IPN, Escuela Superior de Fisica y Matematicas, Av. Instituto Politecnico Nacional s/n, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Vega C, H. R., E-mail: victceronr@hotmail.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico)

2014-08-15

At the present time diverse systems of detection of neutrons exist, as proportional counters based on BF{sub 3}, He{sub 3} and spectrometers of Bonner spheres. However, the cost and the complexity of the implementation of these systems put them far from the reach for dosimetric purposes. For these reasons a system of neutrons detection composed by a medium paraffin moderator that forms a 4π (spheres) arrangement and of several couples of thermoluminescent dosimeters TLD 600/TLD 700. The response of the system presents a minor repeatability to 5% in several assays when being irradiated with a {sup 239}PuBe source and a deviation of 13.8% in the Tl readings of four different spheres. The calibration factor of the system with regard to the neutrons source which was of 56.2 p Sv/nc also was calculated. These detectors will be used as passive monitors of photoneutrons in a radiotherapy room with lineal accelerator of high energy. (Author)

Conceptual moderator studies for the Spallation Neutron Source short-pulse second target station

Energy Technology Data Exchange (ETDEWEB)

Gallmeier, F. X., E-mail: gallmeierfz@ornl.gov; Lu, W.; Riemer, B. W.; Zhao, J. K.; Herwig, K. W.; Robertson, J. L. [Instrument and Source Division, Oak Ridge National Laboratory, P.O. Box 2008, MS6466, Oak Ridge, Tennessee 37831 (United States)

2016-06-15

Candidate moderator configurations for a short-pulse second target station (STS) at the Oak Ridge National Laboratory Spallation Neutron Source (SNS) have been identified using a global optimizer framework built around the MCNPX particle transport code. Neutron brightness metrics were selected as the figure-of-merit. We assumed that STS would use one out of six proton pulses produced by an SNS accelerator upgraded to operate at 1.3 GeV proton energy, 2.8 MW power and 60 Hz repetition rate. The simulations indicate that the peak brightness can be increased by a factor of 5 and 2.5 on a per proton pulse basis compared to the SNS first target station for both coupled and decoupled para-hydrogen moderators, respectively. Additional increases by factors of 3 and 2 were demonstrated for coupled and decoupled moderators, respectively, by reducing the area of neutron emission from 100 × 100 mm{sup 2} to 20 × 20 mm{sup 2}. This increase in brightness has the potential to translate to an increase of beam intensity at the instruments’ sample positions even though the total neutron emission of the smaller moderator is less than that of the larger. This is especially true for instruments with small samples (beam dimensions). The increased fluxes in the STS moderators come at accelerated poison and de-coupler burnout and higher radiation-induced material damage rates per unit power, which overall translate into lower moderator lifetimes. A first effort was undertaken to group decoupled moderators into a cluster collectively positioning them at the peak neutron production zone in the target and having a three-port neutron emission scheme that complements that of a cylindrical coupled moderator.

Microstructured boron foil scintillating G-GEM detector for neutron imaging

Energy Technology Data Exchange (ETDEWEB)

Fujiwara, Takeshi, E-mail: fujiwara-t@aist.go.jp [Research Institute for Measurement and Analytical Instrumentation, Advanced Industrial Science and Technology (AIST), Tsukuba, Ibaraki (Japan); Center for Advanced Photonics, Neutron Beam Technology Team, RIKEN, Saitama (Japan); Bautista, Unico [Department of Nuclear Engineering and Management, The University of Tokyo, Tokyo (Japan); Philippine Nuclear Research Institute-Department of Science and Technology (PNRI-DOST), Commonwealth Avenue, Diliman, Quezon City (Philippines); Mitsuya, Yuki [Nuclear Professional School, The University of Tokyo, Tokai-mura, Naka-gun, Ibaraki (Japan); Takahashi, Hiroyuki [Department of Nuclear Engineering and Management, The University of Tokyo, Tokyo (Japan); Yamada, Norifumi L. [Neutron Science Laboratory, Institute of Material Structure Science, High Energy Accelerator Research Organization (KEK) (Japan); Otake, Yoshie; Taketani, Atsushi [Center for Advanced Photonics, Neutron Beam Technology Team, RIKEN, Saitama (Japan); Uesaka, Mitsuru [Nuclear Professional School, The University of Tokyo, Tokai-mura, Naka-gun, Ibaraki (Japan); Toyokawa, Hiroyuki [Research Institute for Measurement and Analytical Instrumentation, Advanced Industrial Science and Technology (AIST), Tsukuba, Ibaraki (Japan)

2016-12-01

In this study, a new simple neutron imaging gaseous detector was successfully developed by combining a micro-structured {sup 10}B foil, a glass gas electron multiplier (G-GEM), and a mirror–lens–charge-coupled device (CCD)–camera system. The neutron imaging system consists of a chamber filled with Ar/CF{sub 4} scintillating gas mixture. Inside this system, the G-GEM is mounted for gas multiplication. The neutron detection in this system is based on the reaction between {sup 10}B and neutrons. A micro-structured {sup 10}B is developed to overcome the issue of low detection efficiency. Secondary electrons excite Ar/CF{sub 4} gas molecules, and high-yield visible photons are emitted from those excited gas molecules during the gas electron multiplication process in the G-GEM holes. These photons are easily detected by a mirror–lens–CCD–camera system. A neutron radiograph is then simply formed. We obtain the neutron images of different materials with a compact accelerator-driven neutron source. We confirm that the new scintillating G-GEM-based neutron imager works properly with low gamma ray sensitivity and exhibits a good performance as a new simple digital neutron imaging device.

Electroweak Measurements of Neutron Densities in CREX and PREX at JLab, USA

Energy Technology Data Exchange (ETDEWEB)

Horowitz, Charles J. [Indiana U.; Kumar, Krishna S. [UMass; Michaels, Robert W. [JLAB

2014-02-01

Measurement of the parity-violating electron scattering asymmetry is an established technique at Jefferson Lab and provides a new opportunity to measure the weak charge distribution and hence pin down the neutron radius in nuclei in a relatively clean and model-independent way. This is because the Z boson of the weak interaction couples primarily to neutrons. We will describe the PREX and CREX experiments on ${}^{208}$Pb and ${}^{48}$Ca respectively; these are both doubly-magic nuclei whose first excited state can be discriminated by the high resolution spectrometers at JLab. The heavier lead nucleus, with a neutron excess, provides an interpretation of the neutron skin thickness in terms of properties of bulk neutron matter. For the lighter ${}^{48}$Ca nucleus, which is also rich in neutrons, microscopic nuclear theory calculations are feasible and are sensitive to poorly constrained 3-neutron forces.

Calculations of neutron spectra after neutron-neutron scattering

Energy Technology Data Exchange (ETDEWEB)

Crawford, B E [Gettysburg College, Box 405, Gettysburg, PA 17325 (United States); Stephenson, S L [Gettysburg College, Box 405, Gettysburg, PA 17325 (United States); 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); Tornow, W [Duke University and Triangle Universities Nuclear Laboratory, Durham, NC 27708-0308 (United States); Furman, W I [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Lychagin, E V [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Muzichka, A Yu [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Nekhaev, G V [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Strelkov, A 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)

2004-09-01

A direct neutron-neutron scattering length, a{sub nn}, measurement with the goal of 3% accuracy (0.5 fm) is under preparation at the aperiodic pulsed reactor YAGUAR. A direct measurement of a{sub nn} will not only help resolve conflicting results of a{sub nn} by indirect means, but also in comparison to the proton-proton scattering length, a{sub pp}, shed light on the charge-symmetry of the nuclear force. We discuss in detail the analysis of the nn-scattering data in terms of a simple analytical expression. We also discuss calibration measurements using the time-of-flight spectra of neutrons scattered on He and Ar gases and the neutron activation technique. In particular, we calculate the neutron velocity and time-of-flight spectra after scattering neutrons on neutrons and after scattering neutrons on He and Ar atoms for the proposed experimental geometry, using a realistic neutron flux spectrum-Maxwellian plus epithermal tail. The shape of the neutron spectrum after scattering is appreciably different from the initial spectrum, due to collisions between thermal-thermal and thermal-epithermal neutrons. At the same time, the integral over the Maxwellian part of the realistic scattering spectrum differs by only about 6 per cent from that of a pure Maxwellian nn-scattering spectrum.

Steady-State Core Temperature Prediction Based on GAMMA+/CAPP Coupling

International Nuclear Information System (INIS)

Tak, Nam-il; Lee, Hyun-Chul; Lim, Hong-Sik

2015-01-01

In spite of sizable applications of the GAMMA+ code for the thermo-fluid analysis and design of a prismatic VHTR, the existing works are limited to stand-alone calculations. In the stand-alone calculations, information from the neutronic analysis (e.g., reactor power density profile) was considered only once i.e., when the calculations get started. For the neutronic analysis and design of a VHTR, the CAPP code, which is also under development at KAERI, is used. The main objective of this paper is to investigate the capability of GAMMA+ and CAPP coupling and to examine the results of the coupled analysis. Based on the coupling of GAMMA+ and CAPP, the steady-state core temperature was investigated in this work. It is found that the communication of data was successful. And the results of the GAMMA+ and CAPP coupling are found to be reasonable. The design modification of PMR200 is required to satisfy the design limit for the hot spot fuel temperature

Neutron detection using a planar array of superheated superconductors

International Nuclear Information System (INIS)

Meagher, G.

1996-01-01

A new thermal neutron detector consisting of an indium/mylar PASS array with a filler of boron powder has been tested successfully. A boron nucleus captures a neutron and the α-particle emitted in this reaction nucleates the transition to the normal state in a neighbouring grain. A PASS with grain radius r=12 μm showed very low response to irradiation by 835 keV γ-rays. The same PASS responded to α-particles with high efficiency, and thermal neutrons were detected with good efficiency and very low background. Thermal neutrons were also counted with an R=25 μm PASS. The use of large granules will allow spatial resolution to a single grain to be obtained and a read-out based on conventional inductive coupling and fast electronics to be utilized. (orig.)

SPIN-PRECESSION: BREAKING THE BLACK HOLE-NEUTRON STAR DEGENERACY

Energy Technology Data Exchange (ETDEWEB)

Chatziioannou, Katerina; Cornish, Neil; Klein, Antoine; Yunes, Nicolás [Department of Physics, Montana State University, Bozeman, MT 59717 (United States)

2015-01-01

Mergers of compact stellar remnants are prime targets for the LIGO/Virgo gravitational wave detectors. The gravitational wave signals from these merger events can be used to study the mass and spin distribution of stellar remnants, and provide information about black hole horizons and the material properties of neutron stars. However, it has been suggested that degeneracies in the way that the star's mass and spin are imprinted in the waveforms may make it impossible to distinguish between black holes and neutron stars. Here we show that the precession of the orbital plane due to spin-orbit coupling breaks the mass-spin degeneracy, and allows us to distinguish between standard neutron stars and alternative possibilities, such as black holes or exotic neutron stars with large masses and spins.

A possible approach to 14MeV neutron moderation: A preliminary study case.

Science.gov (United States)

Flammini, D; Pilotti, R; Pietropaolo, A

2017-07-01

Deuterium-Tritium (D-T) interactions produce almost monochromatic neutrons with about 14MeV energy. These neutrons are used in benchmark experiments as well as for neutron cross sections assessment in fusion reactors technology. The possibility to moderate 14MeV neutrons for purposes beyond fusion is worth to be studied in relation to projects of intense D-T sources. In this preliminary study, carried out using the MCNP Monte Carlo code, the moderation of 14MeV neutrons is approached foreseeing the use of combination of metallic materials as pre-moderator and reflectors coupled to standard water moderators. Copyright © 2017 Elsevier Ltd. All rights reserved.

Analysis of PBMR transients using a coupled neutron transport/thermal-hydraulics code DORT-TD/thermix

International Nuclear Information System (INIS)

Tyobeka, B.; Ivanov, K.; Pautz, A.

2007-01-01

In the advent of increased demand for safety and economics of nuclear power plants, nuclear engineers and designers are called upon to develop advanced computation tools. In these developments, space-time effects in the dynamics of nuclear reactors must be considered within the framework of a full 3-dimensional treatment of both neutron kinetics and thermal hydraulics. In a recent effort at the Pennsylvania State University, a time-dependent version of the discrete ordinates transport code DORT, DORT-TD was coupled to a 2-dimensional core thermal hydraulics code THERMIX-DIREKT. In the coupling process, a feedback model was developed to account for the feedback effects and was implemented into DORT-TD. During the calculation process for each spatial node of the DORT-TD core model, feedback parameters representative of this node are passed to the feedback module. Using these values, cross section tables are then interpolated for the appropriate macroscopic cross section values. The updated macroscopic cross sections are passed back to DORT-TD to perform transport core calculations, and the power distribution is transferred to THERMIX-DIREKT to obtain the relevant thermal-hydraulics data in turn, and this calculation loop continues. In this paper, DORT-TD/THERMIX is used to simulate transients of interest in the PBMR (Pebble Bed Modular Reactor) safety using established benchmark problems: load change from 100% to 40% power and fast control rod ejection (PBMR-268 benchmark problem). The results obtained are compared with those obtained using the diffusion-based module of the code. The results are only preliminary and so far show that diffusion theory is not such a bad approximation for PBMR for the prediction of integral parameters

Finite element analysis and frequency shift studies for the bridge coupler of the coupled cavity linear accelerator of the spallation neutron source.

Energy Technology Data Exchange (ETDEWEB)

Chen, Z. (Zukun)

2001-01-01

The Spallation Neutron Source (SNS) is an accelerator-based neutron scattering research facility. The linear accelerator (linac) is the principal accelerating structure and divided into a room-temperature linac and a superconducting linac. The normal conducting linac system that consists of a Drift Tube Linac (DTL) and a Coupled Cavity Linac (CCL) is to be built by Los Alamos National Laboratory. The CCL structure is 55.36-meters long. It accelerates H- beam from 86.8 Mev to 185.6 Mev at operating frequency of 805 MHz. This side coupled cavity structure has 8 cells per segment, 12 segments and 11 bridge couplers per module, and 4 modules total. A 5-MW klystron powers each module. The number 3 and number 9 bridge coupler of each module are connected to the 5-MW RF power supply. The bridge coupler with length of 2.5 {beta}{gamma} is a three-cell structure and located between the segments and allows power flow through the module. The center cell of each bridge coupler is excited during normal operation. To obtain a uniform electromagnetic filed and meet the resonant frequency shift, the RF induced heat must be removed. Thus, the thermal deformation and frequency shift studies are performed via numerical simulations in order to have an appropriate cooling design and predict the frequency shift under operation. The center cell of the bridge coupler also contains a large 4-inch slug tuner and a tuning post that used to provide bulk frequency adjustment and field intensity adjustment, so that produce the proper total field distribution in the module assembly.

Cellular neural network to the spherical harmonics approximation of neutron transport equation in x–y geometry

International Nuclear Information System (INIS)

Pirouzmand, Ahmad; Hadad, Kamal

2012-01-01

Highlights: ► This paper describes the solution of time-dependent neutron transport equation. ► We use a novel method based on cellular neural networks (CNNs) coupled with the spherical harmonics method. ► We apply the CNN model to simulate step and ramp perturbation transients in a core. ► The accuracy and capabilities of the CNN model are examined for x–y geometry. - Abstract: In an earlier paper we utilized a novel method using cellular neural networks (CNNs) coupled with spherical harmonics method to solve the steady state neutron transport equation in x–y geometry. Here, the previous work is extended to the study of time-dependent neutron transport equation. To achieve this goal, an equivalent electrical circuit based on a second-order form of time-dependent neutron transport equation and one equivalent group of neutron precursor density is obtained by the CNN method. The CNN model is used to simulate step and ramp perturbation transients in a typical 2D core.

Four-wave neutron-resonance spin echo

International Nuclear Information System (INIS)

Grigoriev, S.V.; Kraan, W.H.; Rekveldt, M.Th.

2004-01-01

We develop a technique of scattering from many-body systems. It is based on the principle of the neutron spin echo (SE), where a neutron wave in the magnetic field splits into two waves, which are separated in space or in time after propagation in this field. The neutron thus prepared as a probe passes through the sample to test its properties on a space R or time t scale. This separation in space or in time can be measured using coherence of these two waves as a phase shift φ between them. These two waves are collected or focused and compensated by the SE technique in order to compare their phases after interaction with the sample. In this way one studies interference between these waves and thus can directly measure the pair-correlation function in space or in time. Instead of two-wave SE we propose to realize the four-wave neutron-resonance spin-echo (NRSE). In our experiments, spin precession produced by a couple of the neutron-resonance coils in one arm is compensated by an identical couple of other NR coils in a second arm of a spin-echo machine. The neutron spin-flip probability ρ in the resonance coils is a key parameter of the NRSE arm. The limiting cases, ρ=0 and ρ=1, provide, in quantum terms, a two-level-two-wave k splitting of the neutron and result in the separation of the split waves into two different lengths in space (R 1 ,R 2 ) or in time (t 1 ,t 2 ). These two cases correspond to Larmor precession with phase φ 1 in the static magnetic fields of the NR flippers or to NRSE precession with φ 2 , respectively. The intermediate case, 0 1 ,R 2 ,R 3 ) or in time (t 1 ,t 2 ,t 3 ). The interference of each pair of waves after compensation results in three different echos with phases φ 1 , φ 2 , and φ 3 =(φ 1 +φ 2 )/2. Focusing or compensating all four waves into a single point of the phase-of-waves diagram produces quantum interference of all newly created waves. This task of focusing is experimentally performed. Different options for the

PAD: a one-dimensional, coupled neutronic-thermodynamic-hydrodynamic computer code

International Nuclear Information System (INIS)

Peterson, D.M.; Stratton, W.R.; McLaughlin, T.P.

1976-12-01

Theoretical and numerical foundations, utilization guide, sample problems, and program listing and glossary are given for the PAD computer code which describes dynamic systems with interactive neutronics, thermodynamics, and hydrodynamics in one-dimensional spherical, cylindrical, and planar geometries. The code has been applied to prompt critical excursions in various fissioning systems (solution, metal, LMFBR, etc.) as well as to nonfissioning systems

Crystallographic structures of absorbates and neutron diffraction

International Nuclear Information System (INIS)

Marti, C.; Thorel, P.

1975-01-01

The advantage of neutron diffraction is that it is possible to work at any pressure and therefore to study an adsorbant-adsorbate couple within a wide pressure and temperature range and at thermodynamic equilibrium. Nitrogen adsorbed on graphite and CF 4 adsorbed on graphite were measured [fr

Development of a new electronic neutron imaging system

Energy Technology Data Exchange (ETDEWEB)

Brenizer, J.S. [Department of Mechanical, Aerospace and Nuclear Engineering, Thornton Hall, University of Virginia, Charlottesville, VA 22903-2442 (United States); Berger, H. [Industrial Quality, Inc., Gaithersburg, MD (United States); Gibbs, K.M. [Industrial Quality, Inc., Gaithersburg, MD (United States); Mengers, P. [Paultek Systems, Inc., Nevada City, CA (United States); Stebbings, C.T. [Department of Mechanical, Aerospace and Nuclear Engineering, Thornton Hall, University of Virginia, Charlottesville, VA 22903-2442 (United States); Polansky, D. [Industrial Quality, Inc., Gaithersburg, MD (United States); Rogerson, D.J. [Naval Air Warfare Center, China Lake, CA (United States)

1999-11-03

An electronic neutron imaging camera system was developed for use with thermal, epithermal, and fast neutrons in applications that include nondestructive inspection of explosives, corrosion, turbine blades, electronics, low Z components, etc. The neutron images are expected to provide information to supplement that available from X-ray tests. The primary camera image area was a 30x30 cm field-of-view with a spatial resolution approaching 1.6 line pairs/mm (lp/mm). The camera had a remotely changeable second lens to limit the field-of-view to 7.6x7.6 cm for high spatial resolution (at least 4 lp/mm) thermal neutron imaging, but neutron and light scatter will limit resolution for fast neutrons to about 0.5 lp/mm. Remote focus capability enhanced camera set-up for optimum operation. The 75 dB dynamic range camera system included {sup 6}Li-based screens for imaging of thermal and epithermal neutrons and ZnS(Ag)-based screens for fast neutron imaging. The fast optics was input to a Super S-25 Gen II image intensifier, fiber optically coupled to a 1134 (h)x486 (v) frame transfer CCD camera. The camera system was designed to be compatible with a Navy-sponsored accelerator neutron source. The planned neutron source is an RF quadrupole accelerator that will provide a fast neutron flux of 10{sup 7} n/cm{sup 2}-s (at a source distance of 1 m) at an energy of about 2.2 MeV and a thermal neutron flux of 10{sup 6} n/cm{sup 2}-s at a source L/D ratio of 30. The electronic camera produced good quality real-time images at these neutron levels. On-chip integration could be used to improve image quality for low flux situations. The camera and accelerator combination provided a useful non-reactor neutron inspection system.

Analysis of the Temporal Response of Coupled Asymmetrical Zero-Power Subcritical Bare Metal Reactor Systems

Energy Technology Data Exchange (ETDEWEB)

Klain, Kimberly L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-06-21

The behavior of symmetrical coupled-core systems has been extensively studied, yet there is a dearth of research on asymmetrical systems due to the increased complexity of the analysis of such systems. In this research, the multipoint kinetics method is applied to asymmetrical zeropower, subcritical, bare metal reactor systems. Existing research on asymmetrical reactor systems assumes symmetry in the neutronic coupling; however, it will be shown that this cannot always be assumed. Deep subcriticality adds another layer of complexity and requires modification of the multipoint kinetics equations to account for the effect of the external neutron source. A modified set of multipoint kinetics equations is derived with this in mind. Subsequently, the Rossi-alpha equations are derived for a two-region asymmetrical reactor system. The predictive capabilities of the radiation transport code MCNP6 for neutron noise experiments are shown in a comparison to the results of a series of Rossi-alpha measurements performed by J. Mihalczo utilizing a coupled set of symmetrical bare highly-enriched uranium (HEU) cylinders. The ptrac option within MCNP6 can generate time-tagged counts in a cell (list-mode data). The list-mode data can then be processed similarly to measured data to obtain values for system parameters such as the dual prompt neutron decay constants observable in a coupled system. The results from the ptrac simulations agree well with the historical measured values. A series of case studies are conducted to study the effects of geometrical asymmetry in the coupling between two bare metal HEU cylinders. While the coupling behavior of symmetrical systems has been reported on extensively, that of asymmetrical systems remains sparse. In particular, it appears that there has been no previous research in obtaining the coupling time constants for asymmetrically-coupled systems. The difficulty in observing such systems is due in part to the inability to determine the

High frame-rate neutron radiography of dynamic events

International Nuclear Information System (INIS)

Bossi, R.H.; Robinson, A.H.; Barton, J.P.

1981-01-01

A system has been developed to perform neutron radiographic analysis of dynamic events having a duration of several milliseconds. The system has been operated in the range of 2000 to 10,000 frames/second. Synchronization has provided high-speed-motion neutron radiographs for evaluation of the firing cycle of 7.62 mm munition rounds within a steel rifle barrel. The system has also been used to demonstrate the ability to produce neutron radiographic movies of two-phase flow. The equipment uses the Oregon State University TRIGA reactor capable of pulsing to 3000 MW peak power, a neutron beam collimator, a scintillator neutron conversion screen coupled to an image intensifier, and a 16 mm high speed movie camera. The peak neutron flux incident at the object position is approximately 4 x 10 11 n/cm 2 s with a pulse, full width at half maximum, of 9 ms. Special studies have been performed on the scintillator conversion screens and on the effects of statistical limitations on the image quality. Modulation transfer function analysis has been used to assist in the evaluation of the system performance

High frame-rate neutron radiography of dynamic events

International Nuclear Information System (INIS)

Bossi, R.H.; Robinson, A.H.; Barton, J.P.

1983-01-01

A system has been developed to perform neutron radiographic analysis of dynamic events having a duration of several milliseconds. The system has been operated in the range of 2000 to 10,000 frames/second. Synchronization has provided high-speed-motion neutron radiographs for evaluation of the firing cycle of 7.62 mm munition rounds within a steel rifle barrel. The system has also been used to demonstrate the ability to produce neutron radiographic movies of two phase flow. The equipment uses the Oregon State University TRIGA reactor capable of pulsing to 3000 MW peak power, a neutron beam collimator, a scintillator neutron conversion screen coupled to an image intensifier, and a 16 mm high speed movie camera. The peak neutron flux incident at the object position is approximately 4 x 10 11 n/cm 2 s with a pulse, full width at half maximum, of 9 ms. Special studies have been performed on the scintillator conversion screens and on the effects of statistical limitations on the image quality. Modulation transfer function analysis has been used to assist in the evaluation of the system performance. (Auth.)

Pulsed White Spectrum Neutron Generator for Explosive Detection

International Nuclear Information System (INIS)

King, Michael J.; Miller, Gill T.; Reijonen, Jani; Ji, Qing; Andresen, Nord; Gicquel, Frederic; Kavlas, Taneli; Leung, Ka-Ngo; Kwan, Joe

2008-01-01

Successful explosive material detection in luggage and similar sized containers is a critical issue in securing the safety of all airline passengers. Tensor Technology Inc. has recently developed a methodology that will detect explosive compounds with pulsed fast neutron transmission spectroscopy. In this scheme, tritium beams will be used to generate neutrons with a broad energy spectrum as governed by the T(t,2n)4He fission reaction that produces 0-9 MeV neutrons. Lawrence Berkeley National Laboratory (LBNL), in collaboration with Tensor Technology Inc., has designed and fabricated a pulsed white-spectrum neutron source for this application. The specifications of the neutron source are demanding and stringent due to the requirements of high yield and fast pulsing neutron emission, and sealed tube, tritium operation. In a unique co-axial geometry, the ion source uses ten parallel rf induction antennas to externally couple power into a toroidal discharge chamber. There are 20 ion beam extraction slits and 3 concentric electrode rings to shape and accelerate the ion beam into a titanium cone target. Fast neutron pulses are created by using a set of parallel-plate deflectors switching between +-1500 volts and deflecting the ion beams across a narrow slit. The generator is expected to achieve 5 ns neutron pulses at tritium ion beam energies between 80-120 kV. First experiments demonstrated ion source operation and successful beam pulsing

Neutron scattering research at JAERI reactors - past, present and future -

International Nuclear Information System (INIS)

Funahashi, Satoru; Morii, Yukio; Minakawa, Nobuaki

1992-01-01

It was in 1961 that the first neutron scattering experiment was performed in Japan at JRR-2. The start of JRR-3 in 1964 accelerated the neutron scattering activities in Japan. The research in this field in Japan grew up by using these two research reactors. Among them JRR-2 has played an important role because its neutron flux was about seven times higher than that of the old JRR-3. The completion of the new JRR-3M in 1990 made an epoch to the neutron scattering activities in Japan. The long-waited JRR-3M came up to the expectations of the scientists of Japan. It is a realization of the ideal reactor with tangential beam holes, cold source and neutron guides in a large guide hall. The flux at the neutron scattering instruments is about five times higher than that of JRR-2. Utilization of JRR-3M has just started. Twelve neutron scattering machines are running there. The number will increase up to close twenty in a couple of years. (author)

Measuring neutron fluences and gamma/x-ray fluxes with CCD cameras

International Nuclear Information System (INIS)

Yates, G.J.; Smith, G.W.; Zagarino, P.; Thomas, M.C.

1991-01-01

The capability to measure bursts of neutron fluences and gamma/x-ray fluxes directly with charge coupled device (CCD) cameras while being able to distinguish between the video signals produced by these two types of radiation, even when they occur simultaneously, has been demonstrated. Volume and area measurements of transient radiation-induced pixel charge in English Electric Valve (EEV) Frame Transfer (FT) charge coupled devices (CCDs) from irradiation with pulsed neutrons (14 MeV) and Bremsstrahlung photons (4--12 MeV endpoint) are utilized to calibrate the devices as radiometric imaging sensors capable of distinguishing between the two types of ionizing radiation. Measurements indicate ∼.05 V/rad responsivity with ≥1 rad required for saturation from photon irradiation. Neutron-generated localized charge centers or ''peaks'' binned by area and amplitude as functions of fluence in the 10 5 to 10 7 n/cm 2 range indicate smearing over ∼1 to 10% of CCD array with charge per pixel ranging between noise and saturation levels

Gamma-Free Neutron Detector Based upon Lithium Phosphate Nanoparticles

International Nuclear Information System (INIS)

Steven Wallace

2007-01-01

A gamma-free neutron-sensitive scintillator is needed to enhance radiation sensing and detection for nonproliferation applications. Such a scintillator would allow very large detectors to be placed at the perimeter of spent-fuel storage facilities at commercial nuclear power plants, so that any movement of spontaneously emitted neutrons from spent nuclear fuel or weapons grade plutonium would be noted in real-time. This task is to demonstrate that the technology for manufacturing large panels of fluor-doped plastic containing lithium-6 phosphate nanoparticles can be achieved. In order to detect neutrons, the nanoparticles must be sufficiently small so that the plastic remains transparent. In this way, the triton and alpha particles generated by the capture of the neutron will result in a photon burst that can be coupled to a wavelength shifting fiber (WLS) producing an optical signal of about ten nanoseconds duration signaling the presence of a neutron emitting source

Neutron beam imaging with GEM detectors

International Nuclear Information System (INIS)

Albani, G.; Cazzaniga, C.; Rebai, M.; Gorini, G.; Croci, G.; Muraro, A.; Cippo, E. Perelli; Tardocchi, M.; Cavenago, M.; Murtas, F.; Claps, G.; Pasqualotto, R.

2015-01-01

Neutron GEM-based detectors represent a new frontier of devices in neutron physics applications where a very high neutron flux must be measured such as future fusion experiments (e.g. ITER Neutral beam Injector) and spallation sources (e.g. the European Spallation source). This kind of detectors can be properly adapted to be used both as beam monitors but also as neutron diffraction detectors that could represent a valid alternative for the 3 He detectors replacement. Fast neutron GEM detectors (nGEM) feature a cathode composed by one layer of polyethylene and one of aluminium (neutron scattering on hydrogen generates protons that are detected in the gas) while thermal neutron GEM detectors (bGEM) are equipped with a borated aluminium cathode (charged particles are generated through the 10 B(n,α) 7 Li reaction). GEM detectors can be realized in large area (1 m 2 ) and their readout can be pixelated. Three different prototypes of nGEM and one prototype of bGEM detectors of different areas and equipped with different types of readout have been built and tested. All the detectors have been used to measure the fast and thermal neutron 2D beam image at the ISIS-VESUVIO beamline. The different kinds of readout patterns (different areas of the pixels) have been compared in similar conditions. All the detectors measured a width of the beam profile consitent with the expected one. The imaging property of each detector was then tested by inserting samples of different material and shape in the beam. All the samples were correctly reconstructed and the definition of the reconstruction depends on the type of readout anode. The fast neutron beam profile reconstruction was then compared to the one obtained by diamond detectors positioned on the same beamline while the thermal neutron one was compared to the imaged obtained by cadmium-coupled x-rays films. Also efficiency and the gamma background rejection have been determined. These prototypes represent the first step towards

Neutron-Induced Charged Particle Studies at LANSCE

Science.gov (United States)

Lee, Hye Young; Haight, Robert C.

2014-09-01

Direct measurements on neutron-induced charged particle reactions are of interest for nuclear astrophysics and applied nuclear energy. LANSCE (Los Alamos Neutron Science Center) produces neutrons in energy of thermal to several hundreds MeV. There has been an effort at LANSCE to upgrade neutron-induced charged particle detection technique, which follows on (n,z) measurements made previously here and will have improved capabilities including larger solid angles, higher efficiency, and better signal to background ratios. For studying cross sections of low-energy neutron induced alpha reactions, Frisch-gridded ionization chamber is designed with segmented anodes for improving signal-to-noise ratio near reaction thresholds. Since double-differential cross sections on (n,p) and (n,a) reactions up to tens of MeV provide important information on deducing nuclear level density, the ionization chamber will be coupled with silicon strip detectors (DSSD) in order to stop energetic charged particles. In this paper, we will present the status of this development including the progress on detector design, calibrations and Monte Carlo simulations. This work is funded by the US Department of Energy - Los Alamos National Security, LLC under Contract DE-AC52-06NA25396.

ARTEMIS: The core simulator of AREVA NP's next generation coupled neutronics/thermal-hydraulics code system ARCADIAR

International Nuclear Information System (INIS)

Hobson, Greg; Merk, Stephan; Bolloni, Hans-Wilhelm; Breith, Karl-Albert; Curca-Tivig, Florin; Van Geemert, Rene; Heinecke, Jochen; Hartmann, Bettina; Porsch, Dieter; Tiles, Viatcheslav; Dall'Osso, Aldo; Pothet, Baptiste

2008-01-01

AREVA NP has developed a next-generation coupled neutronics/thermal-hydraulics code system, ARCADIA R , to fulfil customer's current demands and even anticipate their future demands in terms of accuracy and performance. The new code system will be implemented world-wide and will replace several code systems currently used in various global regions. An extensive phase of verification and validation of the new code system is currently in progress. One of the principal components of this new system is the core simulator, ARTEMIS. Besides the stand-alone tests on the individual computational modules, integrated tests on the overall code are being performed in order to check for non-regression as well as for verification of the code. Several benchmark problems have been successfully calculated. Full-core depletion cycles of different plant types from AREVA's French, American and German regions (e.g. N4 and KONVOI types) have been performed with ARTEMIS (using APOLLO2-A cross sections) and compared directly with current production codes, e.g. with SCIENCE and CASCADE-3D, and additionally with measurements. (authors)

Advanced Neutron Source Cross Section Libraries (ANSL-V): ENDF/B-V based multigroup cross-section libraries for advanced neutron source (ANS) reactor studies

International Nuclear Information System (INIS)

Ford, W.E. III; Arwood, J.W.; Greene, N.M.; Moses, D.L.; Petrie, L.M.; Primm, R.T. III; Slater, C.O.; Westfall, R.M.; Wright, R.Q.

1990-09-01

Pseudo-problem-independent, multigroup cross-section libraries were generated to support Advanced Neutron Source (ANS) Reactor design studies. The ANS is a proposed reactor which would be fueled with highly enriched uranium and cooled with heavy water. The libraries, designated ANSL-V (Advanced Neutron Source Cross Section Libraries based on ENDF/B-V), are data bases in AMPX master format for subsequent generation of problem-dependent cross-sections for use with codes such as KENO, ANISN, XSDRNPM, VENTURE, DOT, DORT, TORT, and MORSE. Included in ANSL-V are 99-group and 39-group neutron, 39-neutron-group 44-gamma-ray-group secondary gamma-ray production (SGRP), 44-group gamma-ray interaction (GRI), and coupled, 39-neutron group 44-gamma-ray group (CNG) cross-section libraries. The neutron and SGRP libraries were generated primarily from ENDF/B-V data; the GRI library was generated from DLC-99/HUGO data, which is recognized as the ENDF/B-V photon interaction data. Modules from the AMPX and NJOY systems were used to process the multigroup data. Validity of selected data from the fine- and broad-group neutron libraries was satisfactorily tested in performance parameter calculations

A model for steady-state and transient determination of subcooled boiling for calculations coupling a thermohydraulic and a neutron physics calculation program for reactor core calculation

International Nuclear Information System (INIS)

Mueller, R.G.

1987-06-01

Due to the strong influence of vapour bubbles on the nuclear chain reaction, an exact calculation of neutron physics and thermal hydraulics in light water reactors requires consideration of subcooled boiling. To this purpose, in the present study a dynamic model is derived from the time-dependent conservation equations. It contains new methods for the time-dependent determination of evaporation and condensation heat flow and for the heat transfer coefficient in subcooled boiling. Furthermore, it enables the complete two-phase flow region to be treated in a consistent manner. The calculation model was verified using measured data of experiments covering a wide range of thermodynamic boundary conditions. In all cases very good agreement was reached. The results from the coupling of the new calculation model with a neutron kinetics program proved its suitability for the steady-state and transient calculation of reactor cores. (orig.) [de

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

Transient coupled calculations of the Molten Salt Fast Reactor using the Transient Fission Matrix approach

Energy Technology Data Exchange (ETDEWEB)

Laureau, A., E-mail: laureau.axel@gmail.com; Heuer, D.; Merle-Lucotte, E.; Rubiolo, P.R.; Allibert, M.; Aufiero, M.

2017-05-15

Highlights: • Neutronic ‘Transient Fission Matrix’ approach coupled to the CFD OpenFOAM code. • Fission Matrix interpolation model for fast spectrum homogeneous reactors. • Application for coupled calculations of the Molten Salt Fast Reactor. • Load following, over-cooling and reactivity insertion transient studies. • Validation of the reactor intrinsic stability for normal and accidental transients. - Abstract: In this paper we present transient studies of the Molten Salt Fast Reactor (MSFR). This generation IV reactor is characterized by a liquid fuel circulating in the core cavity, requiring specific simulation tools. An innovative neutronic approach called “Transient Fission Matrix” is used to perform spatial kinetic calculations with a reduced computational cost through a pre-calculation of the Monte Carlo spatial and temporal response of the system. Coupled to this neutronic approach, the Computational Fluid Dynamics code OpenFOAM is used to model the complex flow pattern in the core. An accurate interpolation model developed to take into account the thermal hydraulics feedback on the neutronics including reactivity and neutron flux variation is presented. Finally different transient studies of the reactor in normal and accidental operating conditions are detailed such as reactivity insertion and load following capacities. The results of these studies illustrate the excellent behavior of the MSFR during such transients.

Transient coupled calculations of the Molten Salt Fast Reactor using the Transient Fission Matrix approach

International Nuclear Information System (INIS)

Laureau, A.; Heuer, D.; Merle-Lucotte, E.; Rubiolo, P.R.; Allibert, M.; Aufiero, M.

2017-01-01

Highlights: • Neutronic ‘Transient Fission Matrix’ approach coupled to the CFD OpenFOAM code. • Fission Matrix interpolation model for fast spectrum homogeneous reactors. • Application for coupled calculations of the Molten Salt Fast Reactor. • Load following, over-cooling and reactivity insertion transient studies. • Validation of the reactor intrinsic stability for normal and accidental transients. - Abstract: In this paper we present transient studies of the Molten Salt Fast Reactor (MSFR). This generation IV reactor is characterized by a liquid fuel circulating in the core cavity, requiring specific simulation tools. An innovative neutronic approach called “Transient Fission Matrix” is used to perform spatial kinetic calculations with a reduced computational cost through a pre-calculation of the Monte Carlo spatial and temporal response of the system. Coupled to this neutronic approach, the Computational Fluid Dynamics code OpenFOAM is used to model the complex flow pattern in the core. An accurate interpolation model developed to take into account the thermal hydraulics feedback on the neutronics including reactivity and neutron flux variation is presented. Finally different transient studies of the reactor in normal and accidental operating conditions are detailed such as reactivity insertion and load following capacities. The results of these studies illustrate the excellent behavior of the MSFR during such transients.

Research on out-phase oscillation in a nuclear-coupled parallel double-channel boiling system

International Nuclear Information System (INIS)

Zhou Linglan; Zhang Hong; Liu Yu; Zang Xi'nian

2011-01-01

In this paper, the RELAP5 thermal-hydraulic system code is coupled with the TDOT-T 3D neutron kinetic code by PVM (Parallel Virtual Machine). A parallel double-channel boiling system is built by the coupled code and the instability boundary of out-of-phase oscillation in the system is obtained. The effects of axis power distribution and neutron feedback on the out-of-phase oscillation are analyzed in details. It is found that there are type-Ⅰ and type-Ⅱ density wave oscillation regions when the axial power peak is located at upstream of the heating section. At relatively lower values of fuel time constant, the neutron feedback always delays both types of density wave oscillations. (authors)

Array detector for neutron pre-emission investigations

International Nuclear Information System (INIS)

Petrascu, M.; Cruceru, I.; Bordeanu, C.

1999-01-01

It was predicted that in a fusion experiment induced by 11 Li halo nuclei on light targets, due to the very large dimension of 11 Li, one may expect that the valence neutrons will not be absorbed together with the 9 Li core, but will be emitted in the early stage of the fusion process. The experiment aiming at checking this expectation was performed at the RIKEN-RIPS facility. It was found from neutron energy spectra measurements, that an important number of fusions, more than 30%, are preceded by the pre-emission of one or two neutrons. In the position spectra measurements a very narrow neutron component has been found. This component is much narrower than that calculated by using the Cluster Shell Model Approximation (COSMA). The recent results of time- position coincidence measurements show that within the narrow component the neutrons are pre-emitted predominantly as neutron pairs. The Program Advisory Committee of RIKEN has approved a new measurement at RIKEN Ring Cyclotron aiming at investigation of neutron-neutron coincidences by using a new neutron array detector. This detector has been recently accomplished within the collaboration existing between IFIN-HH, Romania and RIKEN, Japan. The array system consists of 81 4 x 4 x 12 cm 3 BC400 plastic scintillators each coupled to XP2972 Phototubes. The mounting and the testing of the new neutron array detector will be done at RIKEN. The components of one of the 81 elements of the array detector are shown in a photo. The Monte Carlo calculated neutron detection efficiencies as a function of energy are shown. This detector will be used for the investigation of neutron-neutron coincidences in the case of Si( 11 Li, fusion) reaction. The cross- talk between adjacent and non adjacent detectors will be determined by using a 9 Li beam. As it is known in the case of Si( 9 Li, fusion) the neutrons are of evaporation origin, and since these neutrons are emitted in 4 π the chance for detecting 2 coincident neutrons in the

Determination of hafnium at the 10−4% level (relative to zirconium content) using neutron activation analysis, inductively coupled plasma mass spectrometry and inductively coupled plasma atomic emission spectrometry

International Nuclear Information System (INIS)

Smolik, Marek; Polkowska-Motrenko, Halina; Hubicki, Zbigniew; Jakóbik-Kolon, Agata; Danko, Bożena

2014-01-01

Graphical abstract: -- Highlights: •We worked out ICP-MS method of Hf determination in Zr and Zr compounds. •We used NAA method as reference one. •We obtained pure zirconium matrix by ion exchange (Diphonix ® resin). •These permit to determine ≥1 × 10 −4 % Hf in Zr sample by ICP MS with good precision and accuracy. -- Abstract: Hafnium at the very low level of 1–8 ppm (in relation to zirconium) was determined in zirconium sulfate solutions (originating from investigations of the separation of ca. 44 ppm Hf from zirconium by means of the ion exchange method) by using three independent methods: inductively coupled plasma mass spectrometry (ICP MS), neutron activation analysis (NAA) and inductively coupled plasma atomic emission spectrometry (ICP-AES). The results of NAA and ICP MS determinations were consistent with each other across the entire investigated range (the RSD of both methods did not exceed 38%). The results of ICP-AES determination were more diverse, particularly at less than 5 ppm Hf (RSD was significantly higher: 29–253%). The ion exchange method exploiting Diphonix ® resin proved sufficient efficiency in Zr–Hf separation when the initial concentration ratio of the elements ([Zr] 0 /[Hf] 0 ) ranged from 1200 to ca. 143,000

Neutron-photon multigroup cross sections for neutron energies up to 400 MeV: HILO86R

International Nuclear Information System (INIS)

Kotegawa, Hiroshi; Nakane, Yoshihiro; Hasegawa, Akira; Tanaka, Shun-ichi

1993-02-01

A macroscopic multigroup cross section library of 66 neutron and 22 photon groups for neutron energies up to 400 MeV: HILO86R is prepared for 10 typical shielding materials; water, concrete, iron, air, graphite, polyethylene, heavy concrete, lead, aluminum and soil. The library is a revision of the DLC-119/HILO86, in which only the cross sections below 19.6 MeV have been exchanged with a group cross section processed from the JENDL-3 microscopic cross section library. In the HILO86R library, self shielding factors are used to produce effective cross sections for neutrons less than 19.6 MeV considering rather coarse energy meshes. Energy spectra and dose attenuation in water, concrete and iron have been compared among the HILO, HILO86 and HILO86R libraries for different energy neutron sources. Significant discrepancy has been observed in the energy spectra less than a couple of MeV energy in iron among the libraries, resulting large difference in the dose attenuation. The difference was attributed to the effect of self-shielding factor, namely to the difference between infinite dilution and effective cross sections. Even for 400 MeV neutron source the influence of the self-shielding factor is significant, nevertheless only the cross sections below 19.6 MeV are exchanged. (author)

Short pulse neutron generator

Science.gov (United States)

Elizondo-Decanini, Juan M.

2016-08-02

Short pulse neutron generators are described herein. In a general embodiment, the short pulse neutron generator includes a Blumlein structure. The Blumlein structure includes a first conductive plate, a second conductive plate, a third conductive plate, at least one of an inductor or a resistor, a switch, and a dielectric material. The first conductive plate is positioned relative to the second conductive plate such that a gap separates these plates. A vacuum chamber is positioned in the gap, and an ion source is positioned to emit ions in the vacuum chamber. The third conductive plate is electrically grounded, and the switch is operable to electrically connect and disconnect the second conductive plate and the third conductive plate. The at least one of the resistor or the inductor is coupled to the first conductive plate and the second conductive plate.

A FIFO based neutron arrival time collection technique for assay of plutonium

International Nuclear Information System (INIS)

Parthasarathy, R.; Saisubalakshmi, D.; Venkatasubramani, C.R.

2004-01-01

The system assays plutonium by counting the time correlated neutrons emitted by the spontaneous fissions of the even-even Pu isotopes in the presence of random neutron background, originating principally from (a,n) reactions in the material. The correlation technique discussed in this paper utilizes twofold neutron coincidence counting but the system is proposed to be enhanced for neutron multiplicity counting. A microcontroller based data acquisition system has been developed using a couple of fast FIFO 2kX9 bit memory ICs and a 16 bit counter for identifying time-correlated neutrons. Since the neutron pulses are arriving at a rapid rate, the incoming pulses are buffered in the FIFO and then transferred to PC by the microcontroller through the parallel port. The correlation analysis based on this time arrival information is done in the PC off-line. (author)

Coupled hydrogen moderator optimization with ortho/para hydrogen ratio

International Nuclear Information System (INIS)

Kai, Tetsuya; Harada, Masahide; Teshigawara, Makoto; Watanabe, Noboru; Ikeda, Yujiro

2004-01-01

Neutronic performance of a coupled hydrogen moderator was studied as a function of para hydrogen concentration, moderator thickness and height, premoderator thickness, etc. for the J-PARC spallation neutron source. It was found that a thick (140 mm) moderator with 100% para-hydrogen was optimal to provide a high time- and energy-integrated neutron intensity below 15 meV and high pulse-peak intensities at lower energies. Distribution of the cold neutrons on a moderator viewed surface was studied and found to exhibit an intensity-enhanced region at the fringe part near the premoderator. This rather peculiar distribution suggested that the moderator and the viewed surface must be designed so as to take the full advantage of the brighter region near the premoderator

Effect of isovector coupling channel on the macroscopic part of the nuclear binding energy

International Nuclear Information System (INIS)

Haddad, S.

2011-04-01

The effect of the isovector coupling channel on the macroscopic part of the nuclear binding energy is determined utilizing the relativistic density dependent Thomas-Fermi approach for the calculation of the macroscopic part of the nuclear binding energy, and the dependency of this effect on the numbers of neutrons and protons is studied. The isovector coupling channel leads to increased nuclear binding energy, and this effect sharpens with growing excess of the number of neutrons on the number of protons. (author)

Progress report on neutron scattering at JAERI

Energy Technology Data Exchange (ETDEWEB)

Morii, Yukio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-10-01

In the first half of fiscal year 1997, JRR-3M was operated for 97 days followed by a long term shut down for its annual maintenance. Three days were lost out of 100 scheduled operation days, due to a trouble in irradiation facility. Neutron scattering research activities at the JRR-3M have been extended from that of fiscal year 1996. In the Research Group for Quantum Condensed Matter System, experimental study under high pressures, low temperatures and high fields as well as coupling of these conditions were planned to find new quantum condensed matter systems. And, obtained experimental results were immediately provided to theorists for their investigations. In cooperation with new group, Research Group for Neutron Scattering of Strongly Correlated Electron Systems and Research Group for Neutron Scattering at Ultralow Temperatures were carrying neutron scattering experiments at JRR-3M. Research Group for Neutron Crystallography in Biology had opened a way for investigating biomatter neutron diffraction research with high experimental accuracy by growing a millimeter-class large single crystal. In fiscal year 1997, 39 research projects were conducted by these four groups and other staffs in JAERI, 27 projects collaborated with university researchers and 3 projects collaborated with private enterprises were also conducted as complementary researches. 2117 days of machine times were requested to use 8 neutron scattering instruments this year, which corresponded to 1.51 times larger than those planned at its beginning. (G.K.)

Two-axis Neutron and X-ray Reflectivity

DEFF Research Database (Denmark)

Bouwman, W.G.; Vigild, M.E.; Findeisen, E.

1997-01-01

Sample alignment for neutron (and in some cases x-ray) reflectometry can be complicated due to a coupling between angle and position which occurs when slits are used to define the path of the beam. Misalignments in sample position or sample rotation angle give rise to systematic errors in the exp...

Preliminary Coupling of MATRA Code for Multi-physics Analysis

International Nuclear Information System (INIS)

Kim, Seongjin; Choi, Jinyoung; Yang, Yongsik; Kwon, Hyouk; Hwang, Daehyun

2014-01-01

The boundary conditions such as the inlet temperature, mass flux, averaged heat flux, power distributions of the rods, and core geometry is given by constant values or functions of time. These conditions are separately calculated and provided by other codes, such as a neutronics or a system codes, into the MATRA code. In addition, the coupling of several codes in the different physics field is focused and embodied. In this study, multiphysics coupling methods were developed for a subchannel code (MATRA) with neutronics codes (MASTER, DeCART) and a fuel performance code (FRAPCON-3). Preliminary evaluation results for representative sample cases are presented. The MASTER and DeCART codes provide the power distribution of the rods in the core to the MATRA code. In case of the FRAPCON-3 code, the variation of the rod diameter induced by the thermal expansion is yielded and provided. The MATRA code transfers the thermal-hydraulic conditions that each code needs. Moreover, the coupling method with each code is described

Methodology for coupling computational fluid dynamics and integral transport neutronics

International Nuclear Information System (INIS)

Thomas, J. W.; Zhong, Z.; Sofu, T.; Downar, T. J.

2004-01-01

The CFD code STAR-CD was coupled to the integral transport code DeCART in order to provide high-fidelity, full physics reactor simulations. An interface program was developed to perform the tasks of mapping the STAR-CD mesh to the DeCART mesh, managing all communication between STAR-CD and DeCART, and monitoring the convergence of the coupled calculations. The interface software was validated by comparing coupled calculation results with those obtained using an independently developed interface program. An investigation into the convergence characteristics of coupled calculations was performed using several test models on a multiprocessor LINUX cluster. The results indicate that the optimal convergence of the coupled field calculation depends on several factors, to include the tolerance of the STAR-CD solution and the number of DeCART transport sweeps performed before exchanging data between codes. Results for a 3D, multi-assembly PWR problem on 12 PEs of the LINUX cluster indicate the best performance is achieved when the STAR-CD tolerance and number of DeCART transport sweeps are chosen such that the two fields converge at approximately the same rate. (authors)

Role of neutrons in the fusion of nuclei

International Nuclear Information System (INIS)

Zagrebaev, V. I.; Samarin, V. V.

2007-01-01

The problem of a quantum-mechanical description of a near-barrier fusion of heavy nuclei under conditions of a strong coupling of their relative motion to neutron degrees of freedom is studied. New mutually complementing efficient models employing the multidimensional time-dependent Schroedinger equation and methods for numerically solving the problems in question within these models are proposed. This makes it possible to study the evolution of wave functions for outer neutrons and to calculate nucleon transfer probabilities and occupation numbers for quasimolecular states over the entire energy range, including that of deep-subbarrier energies. It is shown that the valence-neutron wave functions extend over the volumes of both colliding nuclei before their surfaces come into contact and even before these nuclei overcome the Coulomb barrier (collectivization of outer nucleons). This gives rise to a substantial neutron-transfer effect on the fusion of nuclei and, in particular, to a sizable increase in the probability of subbarrier fusion for specific combinations of nuclei

Elemental analysis of fertilizer by fast neutron activation

International Nuclear Information System (INIS)

Bodart, F.; Deconninck, G.

1977-01-01

A simple and accurate technique has been developed to analyse commercial fertilizers for phosphorus, potassium, chlorine, magnesium and silicon. The method is based on fast-neutron activation using a neutron flux of 2x10 11 neutrons/second. The optimum analytical conditions are tabulated. After irradiation, the sample is measured on a conventional counting system including a Ge(Li) detector (10% efficiency and 2 keV resolution for 60 Co) and a multichannel analyser. Monitor foils radioactivity are measured separately at the same time with a 2''x2''NaI detector coupled with a single channel analyser and a scaler. Fast neutron activation has proved to be a fast, simple, reliable and low cost analytical technique for the determination of phosphorus, silicon, potassium, magnesium and chlorine in fertilizers. Not less than five phosphorus determinations are possible in one hour, while two potassium, magnesium and chlorine determinations are made at the same time. (T.G.)

Nigerian Journal of Chemical Research 22 Vol,15, 2010 Neutron ...

African Journals Online (AJOL)

Hp

techniques based on thermal neutron from a nuclear reactor in combination with high resolution gamma-ray ... calibration and validation of analytical procedure. The techniques have ... 6,7,8. ; Inductively. Coupled Plasma Atomic Emission.

Neutron--neutron logging

International Nuclear Information System (INIS)

Allen, L.S.

1977-01-01

A borehole logging tool includes a steady-state source of fast neutrons, two epithermal neutron detectors, and two thermal neutron detectors. A count rate meter is connected to each neutron detector. A first ratio detector provides an indication of the porosity of the formation surrounding the borehole by determining the ratio of the outputs of the two count rate meters connected to the two epithermal neutron detectors. A second ratio detector provides an indication of both porosity and macroscopic absorption cross section of the formation surrounding the borehole by determining the ratio of the outputs of the two count rate meters connected to the two thermal neutron detectors. By comparing the signals of the two ratio detectors, oil bearing zones and salt water bearing zones within the formation being logged can be distinguished and the amount of oil saturation can be determined. 6 claims, 2 figures

Axial distribution of absorbed doses in fast neutron field at the RB reactor

International Nuclear Information System (INIS)

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

1988-11-01

The coupled fast thermal system CFTS at the RB reactor is created for obtaining fast neutron fields. The axial distribution of fast neutron flux density in its second configuration (CFTS-2) is measured. The axial distribution of absorbed doses is computed on the basis of mentioned experimental results. At the end these experimental and computed results are given. (Author)

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

Neutronics-processing interface analyses for the Accelerator Transmutation of Waste (ATW) aqueous-based blanket system

International Nuclear Information System (INIS)

Davidson, J.W.; Battat, M.E.

1993-01-01

Neutronics-processing interface parameters have large impacts on the neutron economy and transmutation performance of an aqueous-based Accelerator Transmutation of Waste (ATW) system. A detailed assessment of the interdependence of these blanket neutronic and chemical processing parameters has been performed. Neutronic performance analyses require that neutron transport calculations for the ATW blanket systems be fully coupled with the blanket processing and include all neutron absorptions in candidate waste nuclides as well as in fission and transmutation products. The effects of processing rates, flux levels, flux spectra, and external-to-blanket inventories on blanket neutronic performance were determined. In addition, the inventories and isotopics in the various subsystems were also calculated for various actinide and long-lived fission product transmutation strategies

A burn-up module coupling to an AMPX system

International Nuclear Information System (INIS)

Salvatore Duque, M.; Gomez, S.E.; Patino, N.E.; Abbate, M.J.; Sbaffoni, M.M.

1990-01-01

The Reactors and Neutrons Division of the Bariloche Atomic Center uses the AMPX system for the study of high conversion reactors (HCR). Such system allows to make neutronic calculations from the nuclear data library (ENDF/B-IV). The Nuclear Engineering career of the Balseiro Institute developed and implemented a burn-up module at a μ-cell level (BUM: Burn-up Module) which agrees with the requirement to be coupled to the AMPX system. (Author) [es

Enhancement of safety analysis reliability for a CANDU-6 reactor using RELAP-CANDU/SCAN coupled code system

International Nuclear Information System (INIS)

Kim, Man Woong; Choi, Yong Seog; Sin, Chul; Kim, Hyun Koon; Kim, Hho Jung; Hwang, Su Hyun; Hong, In Seob; Kim, Chang Hyo

2005-01-01

In LOCA analysis of the CANDU reactor, the system thermal-hydraulic code, RELAP-CANDU, alone cannot predict the transient behavior accurately. Therefore, the best estimate neutronics and system thermal-hydraulic coupled code system is necessary to describe the transient behavior with higher accuracy and reliability. To perform on-line calculation of safety analysis for CANDU reactor, a coupled thermal hydraulics-neutronics code system was developed in such a way that the best-estimate thermal-hydraulic system code for CANDU reactor, RELAP-CANDU, is coupled with the full three-dimensional reactor core kinetic code

Fission multipliers for D-D/D-T neutron generators

International Nuclear Information System (INIS)

Lou, T.P.; Vujic, J.L.; Koivunoro, H.; Reijonen, J.; Leung, K.-N.

2003-01-01

A compact D-D/D-T fusion based neutron generator is being designed at the Lawrence Berkeley National Laboratory to have a potential yield of 10 12 D-D n/s and 10 14 D-T n/s. Because of its high neutron yield and compact size (∼20 cm in diameter by 4 cm long), this neutron generator design will be suitable for many applications. However, some applications required higher flux available from nuclear reactors and spallation neutron sources operated with GeV proton beams. In this study, a subcritical fission multiplier with k eff of 0.98 is coupled with the compact neutron generators in order to increase the neutron flux output. We have chosen two applications to show the gain in flux due to the use of fission multipliers--in-core irradiation and out-of-core irradiation. For the in-core irradiation, we have shown that a gain of ∼25 can be achieved in a positron production system using D-T generator. For the out-of-core irradiation, a gain of ∼17 times is obtained in Boron Neutron Capture Therapy (BNCT) using a D-D neutron generator. The total number of fission neutrons generated by a source neutron in a fission multiplier with k eff is ∼50. For the out-of-core irradiation, the theoretical maximum net multiplication is ∼30 due to the absorption of neutrons in the fuel. A discussion of the achievable multiplication and the theoretical multiplication will be presented in this paper

Pseudo Nambu–Goldstone modes in neutron stars

Energy Technology Data Exchange (ETDEWEB)

Kojo, Toru, E-mail: torujj@mail.ccnu.edu.cn

2017-06-10

If quarks and gluons are either gapped or confined in neutron stars (NSs), the most relevant light modes are Nambu–Goldstone (NG) modes. We study NG modes within a schematic quark model whose parameters at high density are constrained by the two-solar mass constraint. Our model has the color-flavor-locked phase at high density, with the effective couplings as strong as in hadron physics. We find that strong coupling effects make NG modes more massive than in weak coupling predictions, and would erase several phenomena caused by the stressed pairings in mismatched Fermi surfaces. For instance, we found that charged kaons, which are dominated by diquark and anti-diquark components, are not light enough to condense at strong coupling. Implications for gravitational wave signals for NS–NS mergers are also briefly discussed.

PREFACE: Fundamental Neutron Physics: Introduction and Overview Fundamental Neutron Physics: Introduction and Overview

Science.gov (United States)

Holstein, Barry R.

2009-10-01

In the 77 years since its discovery by Chadwick in 1932, the neutron has come to play an increasingly important role in contemporary physics. As the next to lightest baryon, it is, of course, one of the two primary components of the atomic nucleus and studies of isotopes (nuclei with varying numbers of neutrons but the same proton number) and of the neutron drip line are one of the important focuses of the recently approved radioactive beam machine to be built at Michigan State University. Precise knowledge of its ~900 second lifetime is crucial to determination of the time at which nucleosynthesis occurs in the early universe. Because it is electrically neutral, the neutron can penetrate the atomic cloud and neutron scattering has become a powerful tool in the study of the structure of materials in condensed matter and biophysics. These are all important issues, but will not be addressed in the articles presented below. Rather, in the set of manuscripts published herein, we show various ways in which the neutron has come to probe fundamental questions in physics. We present six such articles: Because of its simple structure, neutron beta decay has served as a laboratory for the study of possible symmetry violations, including search for possible Script T-violation via measurement of the D coefficient, search for second class currents and/or possible CVC violation via examination of recoil terms, search for right-handed currents via examination of correlations, search for S, T couplings via measurement of the b parameter, etc. The study of neutron decay is reviewed in the article by Jeff Nico. The use of the neutron as a probe of possible Script T-violation via the existence of a non-zero electric dipole moment is discussed in the article by Steve Lamoreaux. The neutron is a prime player in the experimental study of hadronic parity violation, via experiments involving radiative capture and spin rotation, as examined in the article by Barry Holstein. Because of its

Numerical and computational aspects of the coupled three-dimensional core/ plant simulations: organization for economic cooperation and development/ U.S. nuclear regulatory commission pressurized water reactor main-steam-line-break benchmark-II. 5. TMI-1 Benchmark Performed by Different Coupled Three-Dimensional Neutronics Thermal- Hydraulic Codes

International Nuclear Information System (INIS)

D'Auria, F.; Galassi, G.M.; Spadoni, A.; Gago, J.L.; Grgic, D.

2001-01-01

A comprehensive analysis of a double-ended main-steam-line-break (MSLB) accident assumed to have occurred in the Babcock and Wilcox Three Mile Island (TMI) Unit 1 nuclear power plant (NPP) has been carried out at the Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione of the University of Pisa, Italy. The research has been carried out in cooperation with the University of Zagreb, Croatia, and with partial financial support from the European Union through a grant to one of the authors. The overall activity has been completed within the framework of the participation in the Organization for Economic Cooperation and Development Committee on the Safety of Nuclear Installations-Nuclear Science Committee PWR MSLB Benchmark. Different code versions have been adopted in the analysis. Results from the following codes (or code versions) are described in this paper: 1. RELAP5/mod 3.2.2, gamma version, coupled with the three-dimensional (3-D) neutron kinetics PARCS code; 2. RELAP5/mod 3.2.2, gamma version, coupled with the 3-D neutron kinetics QUABBOX code; 3. RELAP5/3D code coupled with the 3-D neutron kinetics NESTLE code. Boundary and initial conditions of the system, including those relevant to the fuel status, have been supplied by The Pennsylvania State University in cooperation with GPU Nuclear (the utility, owner of TMI) and the U.S. Nuclear Regulatory Commission (NRC). The main challenge for the calculation was the prediction of the return to power (RTP) following the inlet of cold water into the core and one 'stuck-withdrawn' control rod. Non-realistic assumptions were proposed to augment the core power peak following scram. Zero-dimensional neutronics codes were capable of detecting the RTP after scram. However, the application of 3-D neutronics codes to the same scenario allowed the calculation of a similar value for overall core power peak but showed power increase occurrence in about one-tenth of the core volume. The results achieved in phase 1 of

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)

Dosimetry of fission neutrons in a 1-W reactor, UTR-KINKI

CERN Document Server

Endo, S; Yoshitake, Y

2002-01-01

The energy spectrum of fission neutrons in the biological irradiation field of the Kinki University reactor, UTR-KINKI, has been determined by a multi-foil activation analysis coupled with artificial neural network techniques and a Au-foil activation method. The mean neutron energy was estimated to be 1.26+-0.05 MeV from the experimentally determined spectrum. Based on this energy value and other information, the neutron dose rate was estimated to be 19.7+-1.4 cGy/hr. Since this dose rate agrees with that measured by a pair of ionizing chambers (21.4 cGy/hr), we conclude that the mean neutron energy could be estimated with reasonable accuracy in the irradiation field of UTR-KINKI. (author)

Qualification of the coupled RELAP5/PANTHER/COBRA code package for licensing applications

International Nuclear Information System (INIS)

Schneidesch, C.R.; Zhang Jinzhao

2004-01-01

A coupled thermal hydraulics-neutronics code package has been developed at Tractebel Engineering (TE), in which the best-estimate thermal-hydraulic system code, RELAP5/mod2.5, is coupled with the full three-dimensional reactor core kinetics code, PANTHER, via the dynamic data exchange interface, TALINK. The Departure from Nucleate Boiling Ratio (DNBR) is calculated by the sub-channel thermal-hydraulic analysis code COBRA-3C. The package provides the capability to accurately simulate the key physical phenomena in nuclear power plant accidents with strong asymmetric behaviours and system-core interactions. This paper presents the TE coupled code package and focuses on the methodology followed for qualifying it for licensing applications. The qualification of the coupling demonstrated the robustness achieved by the combined 3-D neutron kinetics/system T-H code package for transient simulations. The coupled TE code package has been qualified and will be used at Tractebel Engineering (TE) for analyzing asymmetric PWR accidents with strong core-system interactions

Experimental subcritical facility driven by D-D/D-T neutron generator at BARC, India

Energy Technology Data Exchange (ETDEWEB)

Sinha, Amar, E-mail: image@barc.gov.in; Roy, Tushar; Kashyap, Yogesh; Ray, Nirmal; Shukla, Mayank; Patel, Tarun; Bajpai, Shefali; Sarkar, P.S.; Bishnoi, Saroj

2015-05-01

Highlights: •Experimental subcritical facility BRAHMMA coupled to D-D/D-T neutron generator. •Preliminary results of PNS experiments reported. •Feynman-alpha noise measurements explored with continuous source. -- Abstract: The paper presents design of an experimental subcritical assembly driven by D-D/D-T neutron and preliminary experimental measurements. The system has been developed for investigating the static and dynamic neutronic properties of accelerator driven sub-critical systems. This system is modular in design and it is first in the series of subcritical assemblies being designed. The subcritical core consists of natural uranium fuel with high density polyethylene as moderator and beryllium oxide as reflector. The fuel is embedded in high density polyethylene moderator matrix. Estimated k{sub eff} of the system is ∼0.89. One of the unique features of subcritical core is the use of Beryllium oxide (BeO) as reflector and HDPE as moderator making the assembly a compact modular system. The subcritical core is coupled to Purnima Neutron Generator which works in D-D and D-T mode with both DC and pulsed operation. It has facility for online source strength monitoring using neutron tagging and programmable source modulation. Preliminary experiments have been carried out for spatial flux measurement and reactivity estimation using pulsed neutron source (PNS) techniques with D-D neutrons. Further experiments are being planned to measure the reactivity and other kinetic parameters using noise methods. This facility would also be used for carrying out studies on effect of source importance and measurement of source multiplication factor k{sub s} and external neutron source efficiency φ{sup ∗} in great details. Experiments with D-T neutrons are also underway.

Neutron detection devices with 6LiF converter layers

Science.gov (United States)

Finocchiaro, Paolo; Cosentino, Luigi; Meo, Sergio Lo; Nolte, Ralf; Radeck, Desiree

2018-01-01

The demand for new thermal neutron detectors as an alternative to 3He tubes in research, industrial, safety and homeland security applications, is growing. These needs have triggered research and development activities about new generations of thermal neutron detectors, characterized by reasonable efficiency and gamma rejection comparable to 3He tubes. In this paper we show the state of art of a promising lowcost technique, based on commercial solid state silicon detectors coupled with thin neutron converter layers of 6LiF deposited onto carbon fiber substrates. Several configurations were studied with the GEANT4 simulation code, and then calibrated at the PTB Thermal Neutron Calibration Facility. The results show that the measured detection efficiency is well reproduced by the simulations, therefore validating the simulation tool in view of new designs. These neutron detectors have also been tested at neutron beam facilities like ISIS (Rutherford Appleton Laboratory, UK) and n_TOF (CERN) where a few samples are already in operation for beam flux and 2D profile measurements. Forthcoming applications are foreseen for the online monitoring of spent nuclear fuel casks in interim storage sites.

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

A 14 MeV neutron irradiation facility with an automated fast cyclic pneumatic

International Nuclear Information System (INIS)

Montgomery, M.T.; Yoho, M.D.; Biegalski, S.R.; Landsberger, S.; Welch, L.

2016-01-01

This work details the design criteria, construction, controls, and optimization of the 14 MeV neutron irradiation facility at the University of Texas, built with the motivation of performing neutron activation analysis on samples with short half-lives. The facility couples a D-T neutron generator with a pneumatic transfer system capable of transit of approximately one second between source and detector, while the cyclic automated nature allows for many irradiation/count trials with any number of samples, translating to significantly improved counting statistics. (author)

Simulation for developing new pulse neutron spectrometers I. Creation of new McStas components of moderators of JSNS

CERN Document Server

Tamura, I; Arai, M; Harada, M; Maekawa, F; Shibata, K; Soyama, K

2003-01-01

Moderators components of the McStas code have been created for the design of JSNS instruments. Three cryogenic moderators are adopted in JSNS, one is coupled H sub 2 moderators for high intensity experiments and other two are decoupled H sub 2 with poisoned or unpoisoned for high resolution moderators. Since the characteristics of neutron beams generated from moderators make influence on the performance of pulse neutron spectrometers, it is important to perform the Monte Carlo simulation with neutron source component written precisely. The neutron spectrum and time structure were calculated using NMTC/JAERI97 and MCNP4a codes. The simulation parameters, which describe the pulse shape over entire spectrum as a function of time, are optimized. In this paper, the creation of neutron source components for port No.16 viewed to coupled H sub 2 moderator and for port No.11 viewed to decoupled H sub 2 moderator of JSNS are reported.

The nonlinear dynamics of a coupled fission system

International Nuclear Information System (INIS)

Bilanovic, Z.; Harms, A.A.

1993-01-01

The dynamic properties of a nonlinear and in situ vibrationally perturbed nuclear-to-thermal coupled neutron multiplying medium are examined. Some unique self-organizational temporal patterns appear in such fission systems and suggest a complex underlying dynamic. (Author)

Superconducting magnetic Wollaston prism for neutron spin encoding

Energy Technology Data Exchange (ETDEWEB)

Li, F., E-mail: fankli@indiana.edu; Parnell, S. R.; Wang, T.; Baxter, D. V. [Center for Exploration of Energy and Matter, Indiana University, Bloomington, Indiana 47408 (United States); Hamilton, W. A. [Neutron Sciences Directorate, Oak Ridge National Laboratory, Oak Ridge, Tennessee 37830 (United States); Maranville, B. B. [National Institute of Standards and Technology, Gaithersburg, Maryland 20899 (United States); Semerad, R. [Ceraco Ceramic Coating GmbH, Ismaning 85737 (Germany); Cremer, J. T. [Adelphi Technology Inc., Redwood City, California 94063 (United States); Pynn, R. [Center for Exploration of Energy and Matter, Indiana University, Bloomington, Indiana 47408 (United States); Neutron Sciences Directorate, Oak Ridge National Laboratory, Oak Ridge, Tennessee 37830 (United States)

2014-05-15

A magnetic Wollaston prism can spatially split a polarized neutron beam into two beams with different neutron spin states, in a manner analogous to an optical Wollaston prism. Such a Wollaston prism can be used to encode the trajectory of neutrons into the Larmor phase associated with their spin degree of freedom. This encoding can be used for neutron phase-contrast radiography and in spin echo scattering angle measurement (SESAME). In this paper, we show that magnetic Wollaston prisms with highly uniform magnetic fields and low Larmor phase aberration can be constructed to preserve neutron polarization using high temperature superconducting (HTS) materials. The Meissner effect of HTS films is used to confine magnetic fields produced electromagnetically by current-carrying HTS tape wound on suitably shaped soft iron pole pieces. The device is cooled to ∼30 K by a closed cycle refrigerator, eliminating the need to replenish liquid cryogens and greatly simplifying operation and maintenance. A HTS film ensures that the magnetic field transition within the prism is sharp, well-defined, and planar due to the Meissner effect. The spin transport efficiency across the device was measured to be ∼98.5% independent of neutron wavelength and energizing current. The position-dependent Larmor phase of neutron spins was measured at the NIST Center for Neutron Research facility and found to agree well with detailed simulations. The phase varies linearly with horizontal position, as required, and the neutron beam shows little depolarization. Consequently, the device has advantages over existing devices with similar functionality and provides the capability for a large neutron beam (20 mm × 30 mm) and an increase in length scales accessible to SESAME to beyond 10 μm. With further improvements of the external coupling guide field in the prototype device, a larger neutron beam could be employed.

Superconducting magnetic Wollaston prism for neutron spin encoding

Science.gov (United States)

Li, F.; Parnell, S. R.; Hamilton, W. A.; Maranville, B. B.; Wang, T.; Semerad, R.; Baxter, D. V.; Cremer, J. T.; Pynn, R.

2014-05-01

A magnetic Wollaston prism can spatially split a polarized neutron beam into two beams with different neutron spin states, in a manner analogous to an optical Wollaston prism. Such a Wollaston prism can be used to encode the trajectory of neutrons into the Larmor phase associated with their spin degree of freedom. This encoding can be used for neutron phase-contrast radiography and in spin echo scattering angle measurement (SESAME). In this paper, we show that magnetic Wollaston prisms with highly uniform magnetic fields and low Larmor phase aberration can be constructed to preserve neutron polarization using high temperature superconducting (HTS) materials. The Meissner effect of HTS films is used to confine magnetic fields produced electromagnetically by current-carrying HTS tape wound on suitably shaped soft iron pole pieces. The device is cooled to ˜30 K by a closed cycle refrigerator, eliminating the need to replenish liquid cryogens and greatly simplifying operation and maintenance. A HTS film ensures that the magnetic field transition within the prism is sharp, well-defined, and planar due to the Meissner effect. The spin transport efficiency across the device was measured to be ˜98.5% independent of neutron wavelength and energizing current. The position-dependent Larmor phase of neutron spins was measured at the NIST Center for Neutron Research facility and found to agree well with detailed simulations. The phase varies linearly with horizontal position, as required, and the neutron beam shows little depolarization. Consequently, the device has advantages over existing devices with similar functionality and provides the capability for a large neutron beam (20 mm × 30 mm) and an increase in length scales accessible to SESAME to beyond 10 μm. With further improvements of the external coupling guide field in the prototype device, a larger neutron beam could be employed.

Neutron data evaluation of {sup 238}U

Energy Technology Data Exchange (ETDEWEB)

Maslov, V.M.; Porodzinskij, Y.V.; Hasegawa, Akira; Shibata, Keiichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-08-01

Cross sections for neutron-induced reactions on {sup 238}U are calculated by using the Hauser-Feshbach-Moldauer theory, the coupled channel model and the double-humped fission barrier model. The direct excitation of ground state band levels is calculated with a rigid-rotator model. The direct excitation of vibrational octupole and K = 2{sup +} quadrupole bands is included using a soft (deformable) rotator model. The competition of inelastic scattering to fission reaction is shown to be sensitive to the target nucleus level density at excitations above the pairing gap. As for fission, (n,2n), (n,3n), and (n,4n) reactions, secondary neutron spectra data are consistently reproduced. Pre-equilibrium emission of first neutron is included. Shell effects in the level densities are shown to be important for estimation of energy dependence of non-emissive fission cross section. (author). 105 refs.

Constructing neutron stars with a gravitational Higgs mechanism

Science.gov (United States)

Franchini, Nicola; Coates, Andrew; Sotiriou, Thomas P.

2018-03-01

In scalar-tensor theories, spontaneous scalarization is a phase transition that can occur in ultradense environments such as neutron stars. The scalar field develops a nontrivial configuration once the stars exceeds a compactness threshold. We recently pointed out that, if the scalar exhibits some additional coupling to matter, it could give rise to significantly different microphysics in these environments. In this work we study, at the nonperturbative level, a toy model in which the photon is given a large mass when spontaneous scalarization occurs. Our results demonstrate clearly the effectiveness of spontaneous scalarization as a Higgs-like mechanism in neutron stars.

Improving the refueling cycle of a WWER-1000 using cuckoo search method and thermal-neutronic coupling of PARCS v2.7, COBRA-EN and WIMSD-5B codes

Energy Technology Data Exchange (ETDEWEB)

Yarizadeh-Beneh, M.; Mazaheri-Beni, H.; Poursalehi, N., E-mail: n_poursalehi@sbu.ac.ir

2016-12-15

Highlights: • The cuckoo search algorithm is applied to the loading pattern optimization of a nuclear reactor core. • Calculations during the cycle show a good agreement between results and reference for the original LP. • Results indicate the efficient performance of cuckoo search approach coupled with thermal-neutronic solvers. • Neutronic parameters of proposed core pattern are improved relative to original core pattern. - Abstract: The fuel loading pattern optimization is an important process in the refueling design of a nuclear reactor core. Also the analysis of reactor core performance during the operation cycle can be a significant step in the core loading pattern optimization (LPO). In this work, for the first time, a new method i.e. cuckoo search algorithm (CS) has been applied to the fuel loading pattern design of Bushehr WWER-1000 core. In this regard, two objectives have been chosen for finding the best configuration including the improvement of operation cycle length associated with flattening the radial power distribution of fuel assemblies. The core pattern optimization has been performed by coupling the CS algorithm to thermal-neutronic codes including PARCS v2.7, COBRA-EN and WIMSD-5B for earning desired parameters along the operation cycle. The calculations have been done for the beginning of cycle (BOC) to the end of cycle (EOC) states. According to numerical results, the longer operation cycle for the semi-optimized loading pattern has been achieved along with less power peaking factor (PPF) in comparison to the original core pattern of Bushehr WWER-1000. Gained results confirm the efficient and suitable performance of the developed program and also the introduced CS method in the LPO of a nuclear WWER type.

Novel design concepts for generating intense accelerator based beams of mono-energetic fast neutrons

International Nuclear Information System (INIS)

Franklyn, C.B.; Govender, K.; Guzek, J.; Beer, A. de; Tapper, U.A.S.

2001-01-01

Full text: Successful application of neutron techniques in research, medicine and industry depends on the availability of suitable neutron sources. This is particularly important for techniques that require mono-energetic fast neutrons with well defined energy spread. There are a limited number of nuclear reactions available for neutron production and often the reaction yield is low, particularly for thin targets required for the production of mono-energetic neutron beams. Moreover, desired target materials are often in a gaseous form, such as the reactions D(d,n) 3 He and T(d,n) 3 He, requiring innovative design of targets, with sufficient target pressure and particle beam handling capability. Additional requirements, particularly important in industrial applications, and for research institutions with limited funds, are the cost effectiveness as well as small size, coupled with reliable and continuous operation of the system. Neutron sources based on high-power, compact radio-frequency quadrupole (RFQ) linacs can satisfy these criteria, if used with a suitable target system. This paper discusses the characteristics of a deuteron RFQ linear accelerator system coupled to a high pressure differentially pumped deuterium target. Such a source, provides in excess of 10 10 mono- energetic neutrons per second with minimal slow neutron and gamma-ray contamination, and is utilised for a variety of applications in the field of mineral identification and materials diagnostics. There is also the possibility of utilising a proposed enhanced system for isotope production. The RFQ linear accelerator consists of: 1) Deuterium 25 keV ion source injector; 2) Two close-coupled RFQ resonators, each powered by an rf amplifier supplying up to 300 kW of peak power at 425 MHz; 3) High energy beam transport system consisting of a beam line, a toroid for beam current monitoring, two steering magnets and a quadrupole triplet for beam focusing. Basic technical specifications of the RFQ linac

Imaging of heterogeneous materials by prompt gamma-ray neutron activation analysis

International Nuclear Information System (INIS)

Staples, P.; Prettyman, T.; Lestone, J.

1998-01-01

The authors have used a tomographic gamma scanner (TGS) to produce tomographic prompt gamma-ray neutron activation analysis imaging (PGNAA) of heterogeneous matrices. The TGS was modified by the addition of graphite reflectors that contain isotopic neutron sources for sample interrogation. The authors are in the process of developing the analysis methodology necessary for a quantitative assay of large containers of heterogeneous material. This nondestructive analysis (NDA) technique can be used for material characterization and the determination of neutron assay correction factors. The most difficult question to be answered is the determination of the source-to-sample coupling term. To assist in the determination of the coupling term, the authors have obtained images for a range of sample that are very well characterized, such as, homogenous pseudo one-dimensional samples to three-dimensional heterogeneous samples. They then compare the measurements to MCNP calculations. For an accurate quantitative measurement, it is also necessary to determine the sample gamma-ray self attenuation at higher gamma-ray energies, namely pair production should be incorporated into the analysis codes

Experimental Methods Related to Coupled Fast-Thermal Systems at the RB Reactor

International Nuclear Information System (INIS)

Pesic, M.

2002-01-01

In addition to the review of RB reactor characteristics this presentation is focused on the coupled fast-thermal systems achieved at the reactor. The following experimental methods are presented: neutron spectra measurements; steady state experiments and kinetic measurements ( β eff ) related to the coupled fast-thermal cores

With Iterative and Bosonized Coupling towards Fundamental Particle Properties

CERN Document Server

Binder, B

2003-01-01

Previous results have shown that the linear topological potential-to-phase relationship (well known from Josephson junctions) is the key to iterative coupling and non-perturbative bosonization of the 2 two-spinor Dirac equation. In this paper those results are combined to approach the nature of proton, neutron, and electron via extrapolations from Planck units to the System of Units (SI). The electron acts as a bosonizing bridge between opposite parity topological currents. The resulting potentials and masses are based on a fundamental soliton mass limit and two iteratively obtained coupling constants, where one is the fine structure constant. The simple non-perturbative and relativistic results are within measurement uncertainty and show a very high significance. The deviation for the proton and electron masses are approximately 1 ppb ($10^{-9}$), for the neutron 4 ppb.

Neutron calculation scheme for coupled reactors

International Nuclear Information System (INIS)

Porta, Jacques.

1980-11-01

The CABRI and PHEBUS cores are of the low enrichment rod type in which the fuel is made up of uranium oxide pellets encased in tubular cladding but the SCARABEE core has high enrichment plates, the fuel, an aluminium-uranium alloy (UAl) is metal, rolled into plate form. These three cores in well described rectangular geometry, receive in their centres the very heterogeneous cylindrical test loops (numerous containments of different kinds, large void spaces acting as lagging). After a detailed study of these three reactors, it is found that the search for a calculation scheme (common to the three projects) leads to the elimination of the scattering approximation in our calculations. It is therefore necessary to review the various existing models from a theoretical angle and then to select a particular method, according to the available data processing tools, a choice that will be dictated by the optimization of the parameters: cost in calculation time, difficulties (or ease) of use and accuracy achieved. A problem of experiment interpretation by calculation is dealt with in Chapter 3. The determination of the coupling by calculation is closely linked to the geometrical and energy modelization chosen. But from the experimental angle the determination of the coupling also gives rise to problems with respect to the interpretation of the experimental values obtained by thermal balance determinations, counting of the gamma emission of the fission products of fissile detectors and counting of lanthane 140 in the fuel fission products. The method of calculation is discussed as is the use made of detectors and the counting procedures. In chapter 4, it is not a local modelization that is discussed but an overall one in an original three dimensional calculation [fr

Real time neutron reflectometry using neutron optical imaging

International Nuclear Information System (INIS)

Smith, Gregory S.; Majewski, Jaroslaw

2001-01-01

We will describe recent improvements to the SPEAR reflectometer at the Manuel Lujan Jr. Neutron Scattering Center at Los Alamos. One of the changes consists of wider convergent, incident-beam, collimation to take advantage of optical imaging for specular scattering. In addition, the instrument now views a partially coupled liquid hydrogen moderator as opposed to the decoupled moderator that was previous in-place. While the wavelength distribution is poorer, it matches the time (wavelength) resolution of the reflectometer more closely with the angular resolution. Since the integrated intensity of the partially coupled moderator is higher than the decoupled moderator, we show a similar gain in incident beam flux on the sample without loss of the ability to separate fringes. The increases in intensity from the moderator gain and the improved collimation combine to allow us to measure reflectivities with good statistics down to 10 -4 in a matter of minutes and reflectivities of 10 -6 in an hour. Examples of measurements showing the gain in data accumulation rates are presented. (author)

TRIPOLI-4.3.3 and 4.4, Coupled Neutron, Photon, Electron, Positron 3-D, Time Dependent Monte-Carlo, Transport Calculation

International Nuclear Information System (INIS)

Both, J.P.; Mazzolo, A.; Petit, O.; Peneliau, Y.; Roesslinger, B.

2008-01-01

1 - Description of program or function: TRIPOLI-4 is a general purpose radiation transport code. It uses the Monte Carlo method to simulate neutron and photon behaviour in three-dimensional geometries. The main areas of applications include but are not restricted to: radiation protection and shielding, nuclear criticality safety, fission and fusion reactor design, nuclear instrumentation. In addition, it can simulate electron-photon cascade showers. It computes particle fluxes and currents and several related physical quantities such as, reaction rates, dose rates, heating, energy deposition, effective multiplication factor, perturbation effects due to density, concentration or partial cross-section variations. The summary precises the types of particles, the nuclear data format and cross sections, the energy ranges, the geometry, the sources, the calculated physical quantities and estimators, the biasing, the time-dependant transport for neutrons, the perturbation, the coupled particle transport and the qualification benchmarks. Data libraries distributed with the TRIPOLI-4: ENDFB6R4, ENDL, JEF2, Mott-Rutherford and Qfission. NEA-1716/04: TRIPOLI-4.4 does not contain the source programs. New features available in TRIPOLI-4 version 4 concern the following points: New biasing features, neutron collision in multigroup homogenized mode, display of the collision sites, ENDF format evaluations, computation of the gamma source produced by neutrons, output format for all results, Verbose level for output warnings, photons reactions rates, XML format output, ENDF format evaluations, combinatorial geometry checks, Green's functions files, and neutronics-shielding coupling. 2 - Methods: The geometry package allows the user to describe a three dimensional configuration by means of surfaces (as in the MCNP code) and also through predefined shapes combine with operators (union, intersection, subtraction...). It is also possible to repeat a pattern to built a network of networks

Neutron radiography with ultracold neutrons

International Nuclear Information System (INIS)

Bates, J.C.

1981-01-01

The neutron transmission factor of very thin films may be low if the neutron energy is comparable to the pseudo-potential of the film material. Surprisingly, perhaps, it is relatively easy to obtain neutrons with such low energies in sufficient numbers to produce neutron radiographs. (orig.)

Scattering of 14.2 MeV polarized neutrons from 12C

International Nuclear Information System (INIS)

Casparis, R.; Leemann, B.Th.; Preiswerk, M.; Rudin, H.; Wagner, R.; Zupranski, P.

1976-01-01

Polarized 14.2 MeV neutrons with a polarization of approximately 50% were produced in the 3 H(d(pol),n(pol)) 4 He reaction using vector polarized deuterons from an 'atomic beam' source of polarized ions. The angular distributions of the analyzing power in the elastic and inelastic (Q = -4.43 MeV) scattering of neutrons from carbon have been measured at ten angles in the range from 22 0 to 152 0 c.m. A time-of-flight technique was used to separate elastically and inelastically scattered neutrons. The results have been compared with theoretical calculations obtained with the DWBA and the coupled channels method. (Auth.)

Double perovskites with strong spin-orbit coupling

Science.gov (United States)

Cook, Ashley M.

We first present theoretical analysis of powder inelastic neutron scattering experiments in Ba2FeReO6 performed by our experimental collaborators. Ba2FeReO6, a member of the double perovskite family of materials, exhibits half-metallic behavior and high Curie temperatures Tc, making it of interest for spintronics applications. To interpret the experimental data, we develop a local moment model, which incorporates the interaction of Fe spins with spin-orbital locked magnetic moments on Re, and show that it captures the experimental observations. We then develop a tight-binding model of the double perovskite Ba 2FeReO6, a room temperature ferrimagnet with correlated and spin-orbit coupled Re t2g electrons moving in the background of Fe moments stabilized by Hund's coupling. We show that for such 3d/5d double perovskites, strong correlations on the 5d-element (Re) are essential in driving a half-metallic ground state. Incorporating both strong spin-orbit coupling and the Hubbard repulsion on Re leads to a band structure consistent with ab initio calculations. The uncovered interplay of strong correlations and spin-orbit coupling lends partial support to our previous work, which used a local moment description to capture the spin wave dispersion found in neutron scattering measurements. We then adapt this tight-binding model to study {111}-grown bilayers of half-metallic double perovskites such as Sr2FeMoO6. The combination of spin-orbit coupling, inter-orbital hybridization and symmetry-allowed trigonal distortion leads to a rich phase diagram with tunable ferromagnetic order, topological C= +/-1, +/-2 Chern bands, and a C = +/-2 quantum anomalous Hall insulator regime. We have also performed theoretical analysis of inelastic neutron scattering (INS) experiments to investigate the magnetic excitations in the weakly distorted face-centered-cubic (fcc) iridate double perovskites La2ZnIrO 6 and La2MgIrO6. Models with dominant Kitaev exchange seem to most naturally

Accurate determination of arsenic in arsenobetaine standard solutions of BCR-626 and NMIJ CRM 7901-a by neutron activation analysis coupled with internal standard method.

Science.gov (United States)

Miura, Tsutomu; Chiba, Koichi; Kuroiwa, Takayoshi; Narukawa, Tomohiro; Hioki, Akiharu; Matsue, Hideaki

2010-09-15

Neutron activation analysis (NAA) coupled with an internal standard method was applied for the determination of As in the certified reference material (CRM) of arsenobetaine (AB) standard solutions to verify their certified values. Gold was used as an internal standard to compensate for the difference of the neutron exposure in an irradiation capsule and to improve the sample-to-sample repeatability. Application of the internal standard method significantly improved linearity of the calibration curve up to 1 microg of As, too. The analytical reliability of the proposed method was evaluated by k(0)-standardization NAA. The analytical results of As in AB standard solutions of BCR-626 and NMIJ CRM 7901-a were (499+/-55)mgkg(-1) (k=2) and (10.16+/-0.15)mgkg(-1) (k=2), respectively. These values were found to be 15-20% higher than the certified values. The between-bottle variation of BCR-626 was much larger than the expanded uncertainty of the certified value, although that of NMIJ CRM 7901-a was almost negligible. Copyright (c) 2010 Elsevier B.V. All rights reserved.

Potential of the neutron lloyd's mirror interferometer for the search for new interactions

Energy Technology Data Exchange (ETDEWEB)

Pokotilovski, Yu. N., E-mail: pokot@nf.jinr.ru [Joint Institute for Nuclear Research (Russian Federation)

2013-04-15

We discuss the potential of the neutron Lloyd's mirror interferometer in a search for new interactions at small scales. We consider three hypothetical interactions that may be tested using the interferometer. The chameleon scalar field proposed to solve the enigma of accelerating expansion of the Universe produces interaction between particles and matter. The axion-like spin-dependent coupling between a neutron and nuclei or/and electrons may result in a P- and T-noninvariant interaction with matter. Hypothetical non-Newtonian gravitational interactions mediates an additional short-range potential between neutrons and bulk matter. These interactions between the neutron and the mirror of a Lloyd-type neutron interferometer cause a phase shift of neutron waves. We estimate the sensitivity and systematic effects of possible experiments.

Reactor physics simulations with coupled Monte Carlo calculation and computational fluid dynamics

International Nuclear Information System (INIS)

Seker, V.; Thomas, J.W.; Downar, T.J.

2007-01-01

A computational code system based on coupling the Monte Carlo code MCNP5 and the Computational Fluid Dynamics (CFD) code STAR-CD was developed as an audit tool for lower order nuclear reactor calculations. This paper presents the methodology of the developed computer program 'McSTAR'. McSTAR is written in FORTRAN90 programming language and couples MCNP5 and the commercial CFD code STAR-CD. MCNP uses a continuous energy cross section library produced by the NJOY code system from the raw ENDF/B data. A major part of the work was to develop and implement methods to update the cross section library with the temperature distribution calculated by STARCD for every region. Three different methods were investigated and implemented in McSTAR. The user subroutines in STAR-CD are modified to read the power density data and assign them to the appropriate variables in the program and to write an output data file containing the temperature, density and indexing information to perform the mapping between MCNP and STAR-CD cells. Preliminary testing of the code was performed using a 3x3 PWR pin-cell problem. The preliminary results are compared with those obtained from a STAR-CD coupled calculation with the deterministic transport code DeCART. Good agreement in the k eff and the power profile was observed. Increased computational capabilities and improvements in computational methods have accelerated interest in high fidelity modeling of nuclear reactor cores during the last several years. High-fidelity has been achieved by utilizing full core neutron transport solutions for the neutronics calculation and computational fluid dynamics solutions for the thermal-hydraulics calculation. Previous researchers have reported the coupling of 3D deterministic neutron transport method to CFD and their application to practical reactor analysis problems. One of the principal motivations of the work here was to utilize Monte Carlo methods to validate the coupled deterministic neutron transport

Reliable estimation of neutron flux in BWR reactor vessel using the tort code (2) application to neutron and gamma flux estimation

Energy Technology Data Exchange (ETDEWEB)

Kurosawa, M. [Toshiba Corp., Yokohama (Japan); Tsukiyama, T.; Hayashi, K. [Hitachi Engineering Co. Ltd., Hitachi-shi (Japan)

2001-07-01

A neutron and gamma flux distribution around the core of BWR commercial plant in Japan was calculated, using a three-dimensional transport code, TORT in DOORS32 code system. In the external of the core, the bottom of the model was at an elevation of 150 cm below the bottom of active fuel, the top of the model was at an elevation of the top of the shroud head dome and the radial part of the model was to the outside of the reactor pressure vessel. The top guide beams were modeled explicitly to obtain the neutron and gamma flux distribution both in the beams and outside beams. The each control rod guide tube was also modeled with homogeneous region which included the blade wing and poison tubes so that we could obtain the neutron and gamma flux distribution around the each control rod guide tube. The calculation model mentioned above needed very large memory size which exceeded a few decade giga-bytes. As the using the splicing/coupling method had uncertainly at the splicing/coupling boundary, in this work the calculation was performed without this splicing/coupling method. On the other hand, radioactivity data were measured for a few pieces of the top guide beam, shroud and in-core monitor guide tube in the same plant which was analyzed in the above calculation. So the calculation results were able to be compared with those measured data as benchmarking and at the end of this task, the C/M values at these measured points were obtained and calculation model using TORT was evaluated. (authors)

The new electricity of France PWR: calculation scheme of neutron leakages from the reactor cavity

International Nuclear Information System (INIS)

Vergnaud, T.; Bourdet, L.; Nimal, J.C.; Brandicourt, G.; Champion, G.

1987-04-01

A new calculation scheme is adapted to evaluate neutron fluxes in the reactor cavity and the containment of next french PWR. In this scheme a large part is given to Monte Carlo method, coupled with SN-method, in order to take into account multiple neutron diffusions and the complexity of the reactor geometry

Neutron Skins and Neutron Stars

OpenAIRE

Piekarewicz, J.

2013-01-01

The neutron-skin thickness of heavy nuclei provides a fundamental link to the equation of state of neutron-rich matter, and hence to the properties of neutron stars. The Lead Radius Experiment ("PREX") at Jefferson Laboratory has recently provided the first model-independence evidence on the existence of a neutron-rich skin in 208Pb. In this contribution we examine how the increased accuracy in the determination of neutron skins expected from the commissioning of intense polarized electron be...

Thermal quasiparticle correlations and continuum coupling in nuclei far from stability

International Nuclear Information System (INIS)

Dang, Nguyen Dinh; Arima, Akito

2003-01-01

The contributions of quasiparticle correlations and continuum coupling upon the superfluid properties of neutron-rich Ni isotopes are studied within the modified BCS approximation at finite temperature. The effect of quasiparticle correlations is included using a secondary Bogoliubov transformation explicitly involving the quasiparticle occupation numbers at temperature T. The effect of continuum coupling is taken in to account via the finite widths of the single-particle resonant states. It is shown that the combination of these effects washes out the sharp superfluid-normal phase transition given by the standard finite-temperature BCS calculations. It is also found that the two-neutron separation energy for 84 Ni drops to zero at T congruent with 0.8 MeV

Measurable distributions of unpolarized neutron decay

International Nuclear Information System (INIS)

Glueck, F.

1992-05-01

Several two- and one-dimensional distributions of unpolarized free neutron decay are calculated. The results of the order-α model independent radiative correction calculations are tabulated numerically. With these corrections the theoretical distributions become precise enough to make possible the determination of the ratio of the axial-vector to the vector weak coupling constants to a precision of ∼0.001. (author) 39 refs.; 7 tabs

Design and investigations of a DD compact neutron generator in head radiotherapy

International Nuclear Information System (INIS)

Araujo, Wagner; Campos, Tarcisio Passos Ribeiro

2013-01-01

Neutron generators are device-based particle accelerators for producing neutrons through fusion reactions between hydrogen isotopes. Such devices may enable noninvasive treatments of head and neck tumors, which represent about one hundred twenty-nine thousand cases per year around the world. The present paper shows electromagnetic and nuclear simulations of a neutron generator coupled to collimator and evaluations of radiation dose in an analytical head phantom irradiated by the device. The results provide the generator design and the operation parameter in order to achieve prescribed tumor dose. Also, dose distribution in organs of head is presented, being suitable to surrounding brain tumors close to the skull. As conclusion, there is a visibility of neutron generator applied to brain tumor radiation therapy. (author)

The TE coupled RELAP5/PANTHER/COBRA code package and methodology for integrated PWR accident analysis

International Nuclear Information System (INIS)

Schneidesch, Christophe R.; Zhang, Jinzhao; Ammirabile, Luca; Dalleur, Jean-Paul

2006-01-01

At Tractebel Engineering (TE), a dynamic coupling has been developed between the best estimate thermal hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel core thermal-hydraulic analysis code COBRA 3C has been established for on-line calculation of the Departure from Nucleate Boiling Ratio (DNBR). In addition to the standard RELAP5-PANTHER coupling, the fully dynamic coupling of the RELAP5/PANTHER/COBRA3C-TE code package can be activated for evaluation purposes in which the PANTHER close-channel thermal-hydraulics module is replaced by the COBRA3C-TE with cross flow modelling and extended T-H flow conditions capabilities. The qualification of the RELAP5-PANTHER coupling demonstrated the robustness achieved by the combined 3-D neutron kinetics/system T-H code package for transient simulations. The coupled TE code package has been approved by the Belgian Safety Authorities and is used at TE for analyzing asymmetric PWR accidents with strong core-system interactions. In particular, the TE coupled code package was first used to develop a main steam line break in hot shutdown conditions (SLBHZP) accident analysis methodology based on the TE deterministic bounding approach. This methodology has been reviewed and accepted by the Belgian Safety Authorities for specific applications. Those specific applications are related to the power up-rate and steam generator replacement project of the Doel 2 plant or to the Tihange-3 SLB accident re-analysis. A coupled feedwater line break (FLB) accident analysis methodology is currently being reviewed for application approval. The results of coupled thermal-hydraulic and neutronic analysis of SLB and FLB show that there exist important margins in the traditional final safety analysis report (FSAR) accident analysis. Those margins can be used to increase the operational flexibility of the plants. Moreover, the

The TE coupled RELAP5/PANTHER/COBRA code package and methodology for integrated PWR accident analysis

Energy Technology Data Exchange (ETDEWEB)

Schneidesch, Christophe R.; Zhang, Jinzhao; Ammirabile, Luca; Dalleur, Jean-Paul [Suez-Tractebel Engineering, Avenue Ariane 7, B-1200 Brussels (Belgium)

2006-07-01

At Tractebel Engineering (TE), a dynamic coupling has been developed between the best estimate thermal hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel core thermal-hydraulic analysis code COBRA 3C has been established for on-line calculation of the Departure from Nucleate Boiling Ratio (DNBR). In addition to the standard RELAP5-PANTHER coupling, the fully dynamic coupling of the RELAP5/PANTHER/COBRA3C-TE code package can be activated for evaluation purposes in which the PANTHER close-channel thermal-hydraulics module is replaced by the COBRA3C-TE with cross flow modelling and extended T-H flow conditions capabilities. The qualification of the RELAP5-PANTHER coupling demonstrated the robustness achieved by the combined 3-D neutron kinetics/system T-H code package for transient simulations. The coupled TE code package has been approved by the Belgian Safety Authorities and is used at TE for analyzing asymmetric PWR accidents with strong core-system interactions. In particular, the TE coupled code package was first used to develop a main steam line break in hot shutdown conditions (SLBHZP) accident analysis methodology based on the TE deterministic bounding approach. This methodology has been reviewed and accepted by the Belgian Safety Authorities for specific applications. Those specific applications are related to the power up-rate and steam generator replacement project of the Doel 2 plant or to the Tihange-3 SLB accident re-analysis. A coupled feedwater line break (FLB) accident analysis methodology is currently being reviewed for application approval. The results of coupled thermal-hydraulic and neutronic analysis of SLB and FLB show that there exist important margins in the traditional final safety analysis report (FSAR) accident analysis. Those margins can be used to increase the operational flexibility of the plants. Moreover, the

Neutron-neutron probe for uranium exploration

International Nuclear Information System (INIS)

Smith, R.C.

1979-01-01

A neutron activation probe for assaying the amount of fissionable isotopes in an ore body is described which comprises a casing which is movable through a borehole in the ore body, a neutron source and a number of delayed neutron detectors arranged colinearly in the casing below the neutron source for detecting delayed neutrons

Neutron spectrometry with a monolithic silicon telescope.

Science.gov (United States)

Agosteo, S; D'Angelo, G; Fazzi, A; Para, A Foglio; Pola, A; Zotto, P

2007-01-01

A neutron spectrometer was set-up by coupling a polyethylene converter with a monolithic silicon telescope, consisting of a DeltaE and an E stage-detector (about 2 and 500 microm thick, respectively). The detection system was irradiated with monoenergetic neutrons at INFN-Laboratori Nazionali di Legnaro (Legnaro, Italy). The maximum detectable energy, imposed by the thickness of the E stage, is about 8 MeV for the present detector. The scatter plots of the energy deposited in the two stages were acquired using two independent electronic chains. The distributions of the recoil-protons are well-discriminated from those due to secondary electrons for energies above 0.350 MeV. The experimental spectra of the recoil-protons were compared with the results of Monte Carlo simulations using the FLUKA code. An analytical model that takes into account the geometrical structure of the silicon telescope was developed, validated and implemented in an unfolding code. The capability of reproducing continuous neutron spectra was investigated by irradiating the detector with neutrons from a thick beryllium target bombarded with protons. The measured spectra were compared with data taken from the literature. Satisfactory agreement was found.

Soil-Carbon Measurement System Based on Inelastic Neutron Scattering

International Nuclear Information System (INIS)

Orion, I.; Wielopolski, L.

2002-01-01

Increase in the atmospheric CO 2 is associated with concurrent increase in the amount of carbon sequestered in the soil. For better understanding of the carbon cycle it is imperative to establish a better and extensive database of the carbon concentrations in various soil types, in order to develop improved models for changes in the global climate. Non-invasive soil carbon measurement is based on Inelastic Neutron Scattering (INS). This method has been used successfully to measure total body carbon in human beings. The system consists of a pulsed neutron generator that is based on D-T reaction, which produces 14 MeV neutrons, a neutron flux monitoring detector and a couple of large NaI(Tl), 6'' diameter by 6'' high, spectrometers [4]. The threshold energy for INS reaction in carbon is 4.8 MeV. Following INS of 14 MeV neutrons in carbon 4.44 MeV photons are emitted and counted during a gate pulse period of 10 μsec. The repetition rate of the neutron generator is 104 pulses per sec. The gamma spectra are acquired only during the neutron generator gate pulses. The INS method for soil carbon content measurements provides a non-destructive, non-invasive tool, which can be optimized in order to develop a system for in field measurements

Design of a system for neutrons dosimetry