Fusion power core engineering for the ARIES-ST power plant

International Nuclear Information System (INIS)

Tillack, M.S.; Wang, X.R.; Pulsifer, J.; Malang, S.; Sze, D.K.; Billone, M.; Sviatoslavsky, I.

2003-01-01

ARIES-ST is a 1000 MWe fusion power plant based on a low aspect ratio 'spherical torus' (ST) plasma. The ARIES-ST power core was designed to accommodate the unique features of an ST power plant, to meet the top-level requirements of an attractive fusion energy source, and to minimize extrapolation from the fusion technology database under development throughout the world. The result is an advanced helium-cooled ferritic steel blanket with flowing PbLi breeder and tungsten plasma-interactive components. Design improvements, such as the use of SiC inserts in the blanket to extend the outlet coolant temperature range were explored and the results are reported here. In the final design point, the power and particle loads found in ARIES-ST are relatively similar to other advanced tokamak power plants (e.g. ARIES-RS [Fusion Eng. Des. 38 (1997) 3; Fusion Eng. Des. 38 (1997) 87]) such that exotic technologies were not required in order to satisfy all of the design criteria. Najmabadi and the ARIES Team [Fusion Eng. Des. (this issue)] provide an overview of ARIES-ST design. In this article, the details of the power core design are presented together with analysis of the thermal-hydraulic, thermomechanical and materials behavior of in-vessel components. Detailed engineering analysis of ARIES-ST TF and PF systems, nuclear analysis, and safety are given in the companion papers

Environment-based pin-power reconstruction method for homogeneous core calculations

International Nuclear Information System (INIS)

Leroyer, H.; Brosselard, C.; Girardi, E.

2012-01-01

Core calculation schemes are usually based on a classical two-step approach associated with assembly and core calculations. During the first step, infinite lattice assemblies calculations relying on a fundamental mode approach are used to generate cross-sections libraries for PWRs core calculations. This fundamental mode hypothesis may be questioned when dealing with loading patterns involving several types of assemblies (UOX, MOX), burnable poisons, control rods and burn-up gradients. This paper proposes a calculation method able to take into account the heterogeneous environment of the assemblies when using homogeneous core calculations and an appropriate pin-power reconstruction. This methodology is applied to MOX assemblies, computed within an environment of UOX assemblies. The new environment-based pin-power reconstruction is then used on various clusters of 3x3 assemblies showing burn-up gradients and UOX/MOX interfaces, and compared to reference calculations performed with APOLLO-2. The results show that UOX/MOX interfaces are much better calculated with the environment-based calculation scheme when compared to the usual pin-power reconstruction method. The power peak is always better located and calculated with the environment-based pin-power reconstruction method on every cluster configuration studied. This study shows that taking into account the environment in transport calculations can significantly improve the pin-power reconstruction so far as it is consistent with the core loading pattern. (authors)

Fuzzy logic based power-efficient real-time multi-core system

CERN Document Server

Ahmed, Jameel; Najam, Shaheryar; Najam, Zohaib

2017-01-01

This book focuses on identifying the performance challenges involved in computer architectures, optimal configuration settings and analysing their impact on the performance of multi-core architectures. Proposing a power and throughput-aware fuzzy-logic-based reconfiguration for Multi-Processor Systems on Chip (MPSoCs) in both simulation and real-time environments, it is divided into two major parts. The first part deals with the simulation-based power and throughput-aware fuzzy logic reconfiguration for multi-core architectures, presenting the results of a detailed analysis on the factors impacting the power consumption and performance of MPSoCs. In turn, the second part highlights the real-time implementation of fuzzy-logic-based power-efficient reconfigurable multi-core architectures for Intel and Leone3 processors. .

The online simulation of core physics in nuclear power plant

International Nuclear Information System (INIS)

Zhao Qiang

2005-01-01

The three-dimensional power distribution in core is one of the most important status variables of nuclear reactor. In order to monitor the 3-D in core power distribution timely and accurately, the online simulation system of core physics was designed in the paper. This system combines core physics simulation with the data, which is from the plant and reactor instrumentation. The design of the system consists of the hardware part and the software part. The online simulation system consists of a main simulation computer and a simulation operation station. The online simulation system software includes of the real-time simulation support software, the system communication software, the simulation program and the simulation interface software. Two-group and three-dimensional neutron kinetics model with six groups delayed neutrons was used in the real-time simulation of nuclear reactor core physics. According to the characteristics of the nuclear reactor, the core was divided into many nodes. Resolving the neutron equation, the method of separate variables was used. The input data from the plant and reactor instrumentation system consist of core thermal power, loop temperatures and pressure, control rod positions, boron concentration, core exit thermocouple data, Excore detector signals, in core flux detectors signals. There are two purposes using the data, one is to ensure that the model is as close as the current actual reactor condition, and the other is to calibrate the calculated power distribution. In this paper, the scheme of the online simulation system was introduced. Under the real-time simulation support system, the simulation program is being compiled. Compared with the actual operational data, the elementary simulation results were reasonable and correct. (author)

Development of An On-Line, Core Power Distribution Monitoring System

International Nuclear Information System (INIS)

Tunc ALdemir; Don Miller; Peng Wang

2007-01-01

The objective of the proposed work was to develop a software package that can construct in three-dimensional core power distributions using the signals from constant temperature power sensors distributed in the reactor core. The software developed uses a mode-based state/parameter estimation technique that is particularly attractive when there are model uncertainties and/or large signal noise. The software yields the expected value of local power at the detector locations and points in between, as well as the probability distribution of the local power density

Development of An On-Line, Core Power Distribution Monitoring System

Energy Technology Data Exchange (ETDEWEB)

Tunc ALdemir; Don Miller; Peng Wang

2007-10-02

The objective of the proposed work was to develop a software package that can construct in three-dimensional core power distributions using the signals from constant temperature power sensors distributed in the reactor core. The software developed uses a mode-based state/parameter estmation technique that is particularly attractive when there are model uncertainties and/or large signal noise. The software yields the expected value of local power at the detector locations and points in between, as well as the probability distribution of the local power density

Research on Power Output Characteristics of Magnetic Core in Energy Harvesting Devices

Directory of Open Access Journals (Sweden)

Rong-Ping GUO

2014-07-01

Full Text Available Magnetic core is the dominant factor in the performance of current transformer energy harvesting devices. The power output model of the magnetic core is established and verified through experiments. According to the actual application requirements, the concept of power density is proposed. The relationships of power density to air gap, material and dimension of the magnetic core are analyzed and verified through experiments.

Modeling the Power Variability of Core Speed Scaling on Homogeneous Multicore Systems

Directory of Open Access Journals (Sweden)

Zhihui Du

2017-01-01

Full Text Available We describe a family of power models that can capture the nonuniform power effects of speed scaling among homogeneous cores on multicore processors. These models depart from traditional ones, which assume that individual cores contribute to power consumption as independent entities. In our approach, we remove this independence assumption and employ statistical variables of core speed (average speed and the dispersion of the core speeds to capture the comprehensive heterogeneous impact of subtle interactions among the underlying hardware. We systematically explore the model family, deriving basic and refined models that give progressively better fits, and analyze them in detail. The proposed methodology provides an easy way to build power models to reflect the realistic workings of current multicore processors more accurately. Moreover, unlike the existing lower-level power models that require knowledge of microarchitectural details of the CPU cores and the last level cache to capture core interdependency, ours are easier to use and scalable to emerging and future multicore architectures with more cores. These attributes make the models particularly useful to system users or algorithm designers who need a quick way to estimate power consumption. We evaluate the family of models on contemporary x86 multicore processors using the SPEC2006 benchmarks. Our best model yields an average predicted error as low as 5%.

Power distribution gradients in WWER type cores and fuel failure root causes

Energy Technology Data Exchange (ETDEWEB)

Mikuš, Ján M., E-mail: JanMikus.nrc@hotmail.com

2014-02-15

Highlights: • Power (fission rate) distribution gradients can represent fuel failure root causes. • Positions with above gradients were investigated in WWER type cores on reactor LR-0. • Above gradients were evaluated near core heterogeneities and construction materials. • Results can be used for code validation and fuel failure occurrence investigation. - Abstract: Neutron flux non-uniformity and gradients of neutron current resulting in corresponding power (fission rate) distribution changes can represent root causes of the fuel failure. Such situation can be expected in vicinity of some core heterogeneities and construction materials. Since needed data cannot be obtained from nuclear power plant (NPP), results of some benchmark type experiments performed on light water, zero-power research reactor LR-0 were used for investigation of the above phenomenon. Attention was focused on determination of the spatial power distribution changes in fuel assemblies (FAs): Containing fuel rods (FRs) with Gd burnable absorber in WWER-440 and WWER-1000 type cores, Neighboring the core blanket and dummy steel assembly simulators on the periphery of the WWER-440 standard and low leakage type cores, resp., Neighboring baffle in WWER-1000 type cores, and Neighboring control rod (CR) in WWER-440 type cores, namely (a) power peak in axial power distribution in periphery FRs of the adjacent FAs near the area between CR fuel part and butt joint to the CR absorbing part and (b) decrease in radial power distribution in FRs near CR absorbing part. An overview of relevant experimental results from reactor LR-0 and some information concerning leaking FAs on NPP Temelín are presented. Obtained data can be used for code validation and subsequently for the fuel failure occurrence investigation.

Monitoring device for the power distribution within a nuclear reactor core

International Nuclear Information System (INIS)

Tanzawa, Tomio; Kumanomido, Hironori; Toyoshi, Isamu.

1986-01-01

Purpose: To provide a monitoring device for the power distribution in the reactor core that calculates the power distribution based on the measurement by instruments disposed within the reactor core of BWR type reactors. Constitution: The power distribution monitoring device in a reactor core comprises a signal correcting device, a signal normalizing device and a power distribution calculating device, in which the power distribution calculating device is constituted with an average power calculating device for four fuel assemblies and an average power calculating device for fuel assemblies. Gamma-ray signals corrected by the signal correcting device and signals from neutron detectors are inputted to the signal normalizing device, both of which are calibrated to determine the axial gamma-ray signal distribution in the central water gap region with the four fuel assemblies being as the unit. The average power from the four fuel assemblies are inputted to the fuel assembly average power calculating device to allocate to each of the fuel assembly average power thereby attaining the purpose. Further, thermal restriction values are calculated thereby enabling to secure the fuel integrity. (Kamimura, M.)

Fuel pin failure root causes and power distribution gradients in WWER cores

International Nuclear Information System (INIS)

Mikus, J.

2008-01-01

The purpose of this work is to investigate the influence of some core heterogeneities and reactor construction materials on space power distribution in WWER type cores, especially from viewpoint of the values and gradient occurrence that could result in static loads with some consequences, e.g., fuel pin (FP) or fuel assembly (FA) bowing and possible contribution to the FP failure root causes. Presented information were obtained by means of experiments on research reactor LR-0 concerning the: 1) Power distribution estimation on pellet surface of the FPs neighbouring a FP containing gadolinium (Gd 2 O 3 ) burnable absorber integrated into fuel in WWER-440 and -1000 type cores; 2) Power distribution measurement in periphery FAs neighbouring the baffle in WWER-1000 type cores and 3) Power distribution in FAs neighbouring the control rod absorbing part in a WWER-440 type core. (author)

Preliminary concept of a zero power nuclear reactor core

International Nuclear Information System (INIS)

Mai, Luiz Antonio; Siqueira, Paulo de Tarso D.

2011-01-01

The purpose of this work is to define a zero power core to study the neutronic behavior of a modern research reactor as the future RMB (Brazilian Nuclear Multipurpose reactor). The platform used was the IPEN/MB-01 nuclear reactor, installed at the Nuclear and Energy Research Institute (IPEN-CNEN/SP). Equilibrium among minimal changes in the current reactor facilities and an arrangement that will be as representative as possible of a future core were taken into account. The active parts of the elements (fuel and control/safety) were determined to be exactly equal the elements of a future reactor. After several technical discussions, a basic configuration for the zero power core was defined. This reactor will validate the neutronic calculations and will allow the execution of countless future experiments aiming a real core. Of all possible alternative configurations for the zero power core representative of a future reactor - named ZPC-MRR (Zero Power Core - Modern Research Reactor), it was concluded, through technical and practical arguments, that the core will have an array of 4 x 5 positions, with 19 fuel elements, identical in its active part to a standard MTR (Material Test Reactor), 4 control/safety elements having a unique flat surface and a central position of irradiation. The specifications of the fuel elements (FEs) are the same as defined to standard MTR in its active part, but the inferior nozzles are differentiated because ZPC-MRR will be a set without heat generation. A study of reactivity was performed using MCNP code, and it was estimated that it will have around 2700 pcm reactivity excess in its 19 FEs configuration (alike the present IPEN/MB-01 reactivity). The effective change in the IPEN/MB-01 reactor will be made only in the control rods drive mechanism. It will be necessary to modify the center of this mechanism. Major modifications in the facility will not be necessary. (author)

Preliminary concept of a zero power nuclear reactor core

Energy Technology Data Exchange (ETDEWEB)

Mai, Luiz Antonio; Siqueira, Paulo de Tarso D., E-mail: lamai@ipen.b, E-mail: ptsiquei@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2011-07-01

The purpose of this work is to define a zero power core to study the neutronic behavior of a modern research reactor as the future RMB (Brazilian Nuclear Multipurpose reactor). The platform used was the IPEN/MB-01 nuclear reactor, installed at the Nuclear and Energy Research Institute (IPEN-CNEN/SP). Equilibrium among minimal changes in the current reactor facilities and an arrangement that will be as representative as possible of a future core were taken into account. The active parts of the elements (fuel and control/safety) were determined to be exactly equal the elements of a future reactor. After several technical discussions, a basic configuration for the zero power core was defined. This reactor will validate the neutronic calculations and will allow the execution of countless future experiments aiming a real core. Of all possible alternative configurations for the zero power core representative of a future reactor - named ZPC-MRR (Zero Power Core - Modern Research Reactor), it was concluded, through technical and practical arguments, that the core will have an array of 4 x 5 positions, with 19 fuel elements, identical in its active part to a standard MTR (Material Test Reactor), 4 control/safety elements having a unique flat surface and a central position of irradiation. The specifications of the fuel elements (FEs) are the same as defined to standard MTR in its active part, but the inferior nozzles are differentiated because ZPC-MRR will be a set without heat generation. A study of reactivity was performed using MCNP code, and it was estimated that it will have around 2700 pcm reactivity excess in its 19 FEs configuration (alike the present IPEN/MB-01 reactivity). The effective change in the IPEN/MB-01 reactor will be made only in the control rods drive mechanism. It will be necessary to modify the center of this mechanism. Major modifications in the facility will not be necessary. (author)

On-line reconstruction of in-core power distribution by harmonics expansion method

International Nuclear Information System (INIS)

Wang Changhui; Wu Hongchun; Cao Liangzhi; Yang Ping

2011-01-01

Highlights: → A harmonics expansion method for the on-line in-core power reconstruction is proposed. → A harmonics data library is pre-generated off-line and a code named COMS is developed. → Numerical results show that the maximum relative error of the reconstruction is less than 5.5%. → This method has a high computational speed compared to traditional methods. - Abstract: Fixed in-core detectors are most suitable in real-time response to in-core power distributions in pressurized water reactors (PWRs). In this paper, a harmonics expansion method is used to reconstruct the in-core power distribution of a PWR on-line. In this method, the in-core power distribution is expanded by the harmonics of one reference case. The expansion coefficients are calculated using signals provided by fixed in-core detectors. To conserve computing time and improve reconstruction precision, a harmonics data library containing the harmonics of different reference cases is constructed. Upon reconstruction of the in-core power distribution on-line, the two closest reference cases are searched from the harmonics data library to produce expanded harmonics by interpolation. The Unit 1 reactor of DayaBay Nuclear Power Plant (DayaBay NPP) in China is considered for verification. The maximum relative error between the measurement and reconstruction results is less than 5.5%, and the computing time is about 0.53 s for a single reconstruction, indicating that this method is suitable for the on-line monitoring of PWRs.

Study on in-core fuel management for CNP1500 nuclear power plant

International Nuclear Information System (INIS)

Li Dongsheng

2005-10-01

CNP1500 is a four-loop PWR nuclear power plant with light water as moderator and coolant. The reactor core is composed of 205 AFA-3GXL fuel assemblies. The active core height at cold is 426.4 cm and equivalent diameter is 347.0 cm. The reactor thermal output is 4250 MW, and average linear power density is 179.5 W/cm. The cycle length of equilibrium cycle core is 470 equivalent full power days. For all cycles, the moderator temperature coefficients at all conditions are negative values, the nuclear enthalpy rise factors F ΔH at hot full power, all control rods out and equilibrium xenon are less than the limit value, the maximum discharge assembly burnup is less 55000 MW·d/tU, and the shutdown margin values at the end of life meet design criteria. The low-leakage core loading reduces radiation damage on pressure vessel and is beneficial to prolong use lifetime of it. The in-core fuel management design scheme and main calculation results for CNP1500 nuclear power plant are presented. (author)

The whiteStar development project: Westinghouse's next generation core design simulator and core monitoring software to power the nuclear renaissance

International Nuclear Information System (INIS)

Boyd, W. A.; Mayhue, L. T.; Penkrot, V. S.; Zhang, B.

2009-01-01

The WhiteStar project has undertaken the development of the next generation core analysis and monitoring system for Westinghouse Electric Company. This on-going project focuses on the development of the ANC core simulator, BEACON core monitoring system and NEXUS nuclear data generation system. This system contains many functional upgrades to the ANC core simulator and BEACON core monitoring products as well as the release of the NEXUS family of codes. The NEXUS family of codes is an automated once-through cross section generation system designed for use in both PWR and BWR applications. ANC is a multi-dimensional nodal code for all nuclear core design calculations at a given condition. ANC predicts core reactivity, assembly power, rod power, detector thimble flux, and other relevant core characteristics. BEACON is an advanced core monitoring and support system which uses existing instrumentation data in conjunction with an analytical methodology for on-line generation and evaluation of 3D core power distributions. This new system is needed to design and monitor the Westinghouse AP1000 PWR. This paper describes provides an overview of the software system, software development methodologies used as well some initial results. (authors)

Reconstruction of core axial power shapes using the alternating conditional expectation algorithm

International Nuclear Information System (INIS)

Lee, Eun Ki; Kim, Yong Hee; Cha, Kune Ho; Park, Moon Ghu

1999-01-01

We have introduced the alternating conditional expectation (ACE) algorithm in reconstructing 20-node axial core power shapes from five-level in-core detector powers. The core design code, Reactor Operation and Control Simulation (ROCS), calculates 3-dimensional power distributions for various core states, and the reference core-averaged axial power shapes and corresponding simulated detector powers are utilized to synthesize the axial power shape. By using the ACE algorithm, the optimal relationship between a dependent variable, the plane power, and independent variables, five detector powers, is determined without any preprocessing. A total of ∼3490 data sets per each cycle of YongGwang Nuclear (YGN) power plant units 3 and 4 is used for the regression. Continuous analytic function corresponding to each optimal transformation is calculated by simple regression model. The reconstructed axial power shapes of ∼21,200 cases are compared to the original ROCS axial power shapes. Also, to test the validity and accuracy of the new method, its performance is compared with that of the Fourier fitting method (FFM), a typical method of the deterministic approach. For a total of 21,204 data cases, the averages of root mean square (rms) error, axial peak error (ΔF z ), and axial shape index error (ΔASI) of new method are calculated as 0.81%, 0.51% and 0.00204, while those of FFM are 2.29%, 2.37% and 0.00264, respectively. We also evaluated the wide range of axial power profiles from the xenon-oscillation. The results show that the newly developed method is far superior to FFM; average rms and axial peak error are just ∼35 and ∼20% of those of FFM, respectively

Core power distribution measurement and data processing in Daya Bay Nuclear Power Station

International Nuclear Information System (INIS)

Zhang Hong

1997-01-01

For the first time in China, Daya Bay Nuclear Power Station applied the advanced technology of worldwide commercial pressurized reactors to the in-core detectors, the leading excore six-chamber instrumentation for precise axial power distribution, and the related data processing. Described in this article are the neutron flux measurement in Daya Bay Nuclear Power Station, and the detailed data processing

In-core neutron flux measurements at PARR using self powered neutron detector

International Nuclear Information System (INIS)

Hussain, A.; Ansari, S.A.

1989-10-01

This report describes experimental reactor physics measure ments at PARR using the in-core neutron detectors. Rhodium self powered neutron detectors (SPND) were used in the PARR core and several measurements were made aimed at detector calibration, response time determination and neutron flux measurements. The detectors were calibrated at low power using gold foils and full power by the thermal channel. Based on this calibration it was observed that the detector response remains almost linear throughout the power range. The self powered detectors were used for on-line determination of absolute neutron flux in the core as well as the spatial distribution of neutron flux or reactor power. The experimental, axial and horizontal flux mapping results at certain locations in the core are presented. The total response time of rhodium detector was experimentally determined to be about 5 minutes, which agree well with the theoretical results. Because of longer response time of SPND of the detectors it is not possible to use them in the reactor protection system. (author). 10 figs

Optimal Design and Analysis of the Stepped Core for Wireless Power Transfer Systems

Directory of Open Access Journals (Sweden)

Xiu Zhang

2016-01-01

Full Text Available The key of wireless power transfer technology rests on finding the most suitable means to improve the efficiency of the system. The wireless power transfer system applied in implantable medical devices can reduce the patients’ physical and economic burden because it will achieve charging in vitro. For a deep brain stimulator, in this paper, the transmitter coil is designed and optimized. According to the previous research results, the coils with ferrite core can improve the performance of the wireless power transfer system. Compared with the normal ferrite core, the stepped core can produce more uniform magnetic flux density. In this paper, the finite element method (FEM is used to analyze the system. The simulation results indicate that the core loss generated in the optimal stepped ferrite core can reduce about 10% compared with the normal ferrite core, and the efficiency of the wireless power transfer system can be increased significantly.

Modeling of the core of Atucha II nuclear power plant

International Nuclear Information System (INIS)

Blanco, Anibal

2007-01-01

This work is part of a Nuclear Engineer degree thesis of the Instituto Balseiro and it is carried out under the development of an Argentinean Nuclear Power Plant Simulator. To obtain the best representation of the reactor physical behavior using the state of the art tools this Simulator should couple a 3D neutronics core calculation code with a thermal-hydraulics system code. Focused in the neutronic nature of this job, using PARCS, we modeled and performed calculations of the nuclear power plant Atucha 2 core. Whenever it is possible, we compare our results against results obtained with PUMA (the official core code for Atucha 2). (author) [es

In core system mapping reactor power distribution

International Nuclear Information System (INIS)

Yoriyaz, H.; Moreira, J.M.L.

1989-01-01

Based on the signals of SPND'S (Self Powered Neutron Detectors) distributed inside of a core, the spatial power distribution is obtained using the MAP program, developed in this work. The methodology applied in MAP program uses a least mean square technique to calculate expansion coefficients that depend on the SPND'S signals. The final power or neutron flux distribution is obtained by a combination of certains functions or expansion modes that are provided from diffusion calculation with the CITATION code. The MAP program is written in PASCAL language and will be used in IEA-R1 reactor for assisting its operation. (author) [pt

Comparative assessment of out-of-core nuclear thermionic power systems

International Nuclear Information System (INIS)

Estabrook, W.C.; Koenig, D.R.; Prickett, W.Z.

1975-01-01

The hardware selections available for fabrication of a nuclear electric propulsion stage for planetary exploration were explored. The investigation was centered around a heat-pipe-cooled, fast-spectrum nuclear reactor for an out-of-core power conversion system with sufficient detail for comparison with the in-core system studies completed previously. A survey of competing power conversion systems still indicated that the modular reliability of thermionic converters makes them the desirable choice to provide the 240-kWe end-of-life power for at least 20,000 full power hours. The electrical energy will be used to operate a number of mercury ion bombardment thrusters with a specific impulse in the range of about 4,000-5,000 seconds. (Author)

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.)

Power reactor core safety research

International Nuclear Information System (INIS)

Rim, C.S.; Kim, W.C.; Shon, D.S.; Kim, J.

1981-01-01

As a part of nuclear safety research program, a project was launched to develop a model to predict fuel failure, to produce the data required for the localizaton of fuel design and fabrication technology, to establish safety limits for regulation of nuclear power plants and to develop reactor operation method to minimize fuel failure through the study of fuel failure mechanisms. During 1980, the first year of this project, various fuel failure mechanisms were analyzed, an experimental method for out-of-pile tests to study the stress corrosion cracking (SCC) behaviour of Zircaloy cladding underiodine environment was established, and characteristics of PWR and CANDU Zircaloy specimens were examined. Also developed during 1980 were the methods and correlations to evaluate fuel failures in the reactor core based on operating data from power reactors

Cause and countermeasure for heat up of HTTR core support plate at power rise tests

Energy Technology Data Exchange (ETDEWEB)

Fujimoto, Nozomu; Takada, Eiji; Nakagawa, Shigeaki; Tachibana, Yukio; Kawasaki, Kozo; Saikusa, Akio; Kojima, Takao; Iyoku, Tatuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

2002-01-01

HTTR has carried out many kinds of tests as power rise tests in which reactor power rises step by step after attained the first criticality. In the tests, temperature of a core support plate reached higher than expected at each power level, the temperature was expected to be higher than the maximum working temperature at 100% power level. Therefore, tests under the high temperature test operation mode, in which the core flow rate was different, were carried out to predict the temperature at 100% power precisely, and investigate the cause of the temperature rise. From the investigation, it was clear that the cause was gap flow in the core support structure. Furthermore, it was estimated that the temperature of the core support plate rose locally due to change in gap width between the core support plate and a seal plate due to change in core pressure drop. The maximum working temperature of the core support plate was revised. The integrity of core support plate under the revised maximum working temperature condition was confirmed by stress analyses. (author)

Core support structure for nuclear power plants

International Nuclear Information System (INIS)

Steinkamp, E.; Tautz, J.; Ries, H.

1979-01-01

A core support structure for nuclear power plants includes a grid of mutually crossing bridges and a support ring surrounding the grid and connected to ends of the outer bridges of the grid, the grid being formed of profile rod crosses having legs of given length, respective legs of pairs of adjacent crosses abutting one another endwise to form together a side of the smallest mesh opening of the grid, and weld means for securing the profile rod crosses to one another at the mutually abutting ends of the legs thereof; and method of producing the foregoing core support structure

Fuel management service for Tarapur Atomic Power Station core thermal hydraulics

International Nuclear Information System (INIS)

Saha, D.; Venkat Raj, V.; Markandeya, S.G.

1977-01-01

Core thermal hydraulic analysis forms an integral part of the fuel management service for the Tarapur reactors. A distinguishing feature of boiling water reactors is the dependence of core flow distribution on the power distribution. Because of the changes in the axial and radial power distribution from cycle to cycle as well as during the cycle and also the variations in leakage flow, it is necessary to evaluate the core thermal hydraulic parameters for every cycle. Some of the typical results obtained in the course of analysis for different cycles of both the units at Tarapur are presented. The use of MCPR (Minimum Critical Power Ratio), instead of MCHFR (Minimum Critical Heat Flux Ratio) as a figure of merit for fuel cladding integrity is also discussed. (K.B.)

Safety aspects of core power distribution surveillance and control

International Nuclear Information System (INIS)

Beraha, D.; Grumbach, R.; Hoeld, A.; Werner, W.

1978-01-01

The incentives for improved core surveillance and core control systems are outlined. An efficient code for evaluating the power distribution is indispensable for designing and testing such a system. The characteristics of the core simulator QUABOX/CUBBOX and the features required for off-line and on-line applications are described. The important role of the simulator for the safety assessment of a digital core control system is underlined. With regard to the safety aspects of core control, possible disturbances are classified. Simulation results are given concerning the failure of a control actuator. It is shown that means can be devised to prevent unstable behaviour of the control system and, furthermore, to contribute to a safe reactor operation by accounting for process disturbances. (author)

Microfabricated Air-core Toroidal Inductor In Very High Frequency Power Converters

DEFF Research Database (Denmark)

Lê Thanh, Hoà; Nour, Yasser; Han, Anpan

2018-01-01

Miniaturization of power supplies is required for future intelligent electronic systems e.g. internet of things devices. Inductors play an essential role, and they are by far the most bulky and expensive components in power supplies. This paper presents a miniaturized microelectromechanical systems...... (MEMS) inductor and its performance in a very high frequency (VHF) power converter. The MEMS inductor is a siliconembedded air-core toroidal inductor, and it is constructed with through-silicon vias, suspended copper windings, silicon fixtures, and a silicon support die. The air-core inductors...

Research on intelligent monitor for 3D power distribution of reactor core

International Nuclear Information System (INIS)

Xia, Hong; Li, Bin; Liu, Jianxin

2014-01-01

Highlights: • Core power distribution of ex-core measurement system has been reconstructed. • Building up an artificial intelligence model for 3-D core power distribution. • Error of the experiments has been reduced to 0.76%. • Methods for improving the accuracy of the model have been obtained. - Abstract: A real-time monitor for 3D reactor power distribution is critical for nuclear safety and high efficiency of NPP’s operation as well as for optimizing the control system, especially when the nuclear power plant (NPP) works at a certain power level or it works in load following operation. This paper was based on analyzing the monitor for 3D reactor power distribution technologies used in modern NPPs. Furthermore, considering the latest research outcomes, the paper proposed a method based on using an ex-core neutron detector system and a neural network to set up a real time monitor system for reactor’s 3D power distribution supervision. The results of the experiments performed on a reactor simulation machine illustrated that the new monitor system worked very well for a certain burn-up range during the fuel cycle. In addition, this new model could reduce the errors associated with the fitting of the distribution effectively, and several optimization methods were also obtained to improve the accuracy of the simulation model

Self powered neutron detectors as in-core detectors for Sodium-cooled Fast Reactors

Energy Technology Data Exchange (ETDEWEB)

Verma, V., E-mail: vasudha.verma@physics.uu.se [Division of Applied Nuclear Physics, Uppsala University, Box 516, SE-75120 Uppsala (Sweden); CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-lez-Durance (France); Barbot, L.; Filliatre, P. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-lez-Durance (France); Hellesen, C. [Division of Applied Nuclear Physics, Uppsala University, Box 516, SE-75120 Uppsala (Sweden); Jammes, C. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-lez-Durance (France); Svärd, S. Jacobsson [Division of Applied Nuclear Physics, Uppsala University, Box 516, SE-75120 Uppsala (Sweden)

2017-07-11

Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor. Diverse possibilities of detector system installation must be studied for various locations in the reactor vessel in order to detect any perturbations in the core. Results from a previous paper indicated that it is possible to detect changes in neutron source distribution initiated by an inadvertent withdrawal of outer control rod with in-vessel fission chambers located azimuthally around the core. It is, however, not possible to follow inner control rod withdrawal and precisely know the location of the perturbation in the core. Hence the use of complimentary in-core detectors coupled with the peripheral fission chambers is proposed to enable robust core monitoring across the radial direction. In this paper, we assess the feasibility of using self-powered neutron detectors (SPNDs) as in-core detectors in fast reactors for detecting local changes in the power distribution when the reactor is operated at nominal power. We study the neutron and gamma contributions to the total output current of the detector modelled with Platinum as the emitter material. It is shown that this SPND placed in an SFR-like environment would give a sufficiently measurable prompt neutron induced current of the order of 600 nA/m. The corresponding induced current in the connecting cable is two orders of magnitude lower and can be neglected. This means that the SPND can follow in-core power fluctuations. This validates the operability of an SPND in an SFR-like environment. - Highlights: • Studied possibility of using SPNDs as in-core detectors in SFRs. • Study done to detect local power profile changes when reactor is at nominal power. • SPND with a Pt-emitter gives measurable prompt current of the order of 600 nA/m. • Dominant proportion of prompt response is maintained throughout the operation. • Detector signal gives dynamic information on the power fluctuations.

High-power picosecond pulse delivery through hollow core photonic band gap fibers

DEFF Research Database (Denmark)

Michieletto, Mattia; Johansen, Mette Marie; Lyngsø, Jens Kristian

2015-01-01

We demonstrated robust and bend insensitive fiber delivery of high power pulsed laser with diffraction limited beam quality for two different kind of hollow core photonic band gap fibers......We demonstrated robust and bend insensitive fiber delivery of high power pulsed laser with diffraction limited beam quality for two different kind of hollow core photonic band gap fibers...

Estimation of a Reactor Core Power Peaking Factor Using Support Vector Regression and Uncertainty Analysis

International Nuclear Information System (INIS)

Bae, In Ho; Naa, Man Gyun; Lee, Yoon Joon; Park, Goon Cherl

2009-01-01

The monitoring of detailed 3-dimensional (3D) reactor core power distribution is a prerequisite in the operation of nuclear power reactors to ensure that various safety limits imposed on the LPD and DNBR, are not violated during nuclear power reactor operation. The LPD and DNBR should be calculated in order to perform the two major functions of the core protection calculator system (CPCS) and the core operation limit supervisory system (COLSS). The LPD at the hottest part of a hot fuel rod, which is related to the power peaking factor (PPF, F q ), is more important than the LPD at any other position in a reactor core. The LPD needs to be estimated accurately to prevent nuclear fuel rods from melting. In this study, support vector regression (SVR) and uncertainty analysis have been applied to estimation of reactor core power peaking factor

Low-Power Embedded DSP Core for Communication Systems

Science.gov (United States)

Tsao, Ya-Lan; Chen, Wei-Hao; Tan, Ming Hsuan; Lin, Maw-Ching; Jou, Shyh-Jye

2003-12-01

This paper proposes a parameterized digital signal processor (DSP) core for an embedded digital signal processing system designed to achieve demodulation/synchronization with better performance and flexibility. The features of this DSP core include parameterized data path, dual MAC unit, subword MAC, and optional function-specific blocks for accelerating communication system modulation operations. This DSP core also has a low-power structure, which includes the gray-code addressing mode, pipeline sharing, and advanced hardware looping. Users can select the parameters and special functional blocks based on the character of their applications and then generating a DSP core. The DSP core has been implemented via a cell-based design method using a synthesizable Verilog code with TSMC 0.35[InlineEquation not available: see fulltext.]m SPQM and 0.25[InlineEquation not available: see fulltext.]m 1P5M library. The equivalent gate count of the core area without memory is approximately 50 k. Moreover, the maximum operating frequency of a[InlineEquation not available: see fulltext.] version is 100 MHz (0.35[InlineEquation not available: see fulltext.]m) and 140 MHz (0.25[InlineEquation not available: see fulltext.]m).

CASPER: Embedding Power Estimation and Hardware-Controlled Power Management in a Cycle-Accurate Micro-Architecture Simulation Platform for Many-Core Multi-Threading Heterogeneous Processors

Directory of Open Access Journals (Sweden)

Arun Ravindran

2012-02-01

Full Text Available Despite the promising performance improvement observed in emerging many-core architectures in high performance processors, high power consumption prohibitively affects their use and marketability in the low-energy sectors, such as embedded processors, network processors and application specific instruction processors (ASIPs. While most chip architects design power-efficient processors by finding an optimal power-performance balance in their design, some use sophisticated on-chip autonomous power management units, which dynamically reduce the voltage or frequencies of idle cores and hence extend battery life and reduce operating costs. For large scale designs of many-core processors, a holistic approach integrating both these techniques at different levels of abstraction can potentially achieve maximal power savings. In this paper we present CASPER, a robust instruction trace driven cycle-accurate many-core multi-threading micro-architecture simulation platform where we have incorporated power estimation models of a wide variety of tunable many-core micro-architectural design parameters, thus enabling processor architects to explore a sufficiently large design space and achieve power-efficient designs. Additionally CASPER is designed to accommodate cycle-accurate models of hardware controlled power management units, enabling architects to experiment with and evaluate different autonomous power-saving mechanisms to study the run-time power-performance trade-offs in embedded many-core processors. We have implemented two such techniques in CASPER–Chipwide Dynamic Voltage and Frequency Scaling, and Performance Aware Core-Specific Frequency Scaling, which show average power savings of 35.9% and 26.2% on a baseline 4-core SPARC based architecture respectively. This power saving data accounts for the power consumption of the power management units themselves. The CASPER simulation platform also provides users with complete support of SPARCV9

A Novel Method To On-Line Monitor Reactor Nuclear Power And In-Core Thermal Environments

International Nuclear Information System (INIS)

Liu, Hanying; Miller, Don W.; Li, Dongxu; Radcliff, Thomas D.

2002-01-01

For current nuclear power plants, nuclear power can not be directly measured and in-core fuel thermal environments can not be monitored due to the unavailability of an appropriate measurement technology and the inaccessibility of the fuel. If the nuclear deposited power and the in-core thermal conditions (i.e. fuel or coolant temperature and heat transfer coefficient) can be monitored in-situ, then it would play a valuable and critical role in increasing nuclear power, predicting abnormal reactor operation, improving core physical models and reducing core thermal margin so as to implement higher fuel burn-up. Furthermore, the management of core thermal margin and fuel operation may be easier during reactor operation, post-accident or spent fuel storage. On the other hand, for some advanced Generation IV reactors, the sealed and long-lived reactor core design challenges traditional measurement techniques while conventional ex-core detectors and current in-core detectors can not monitor details of the in-core fuel conditions. A method is introduced in this paper that responds to the challenge to measure nuclear power and to monitor the in-core thermal environments, for example, local fuel pin or coolant heat convection coefficient and temperature. In summary, the method, which has been designed for online in-core measurement and surveillance, will be beneficial to advanced plant safety, efficiency and economics by decreasing thermal margin or increasing nuclear power. The method was originally developed for a constant temperature power sensor (CTPS). The CTPS is undergoing design and development for an advanced reactor core to measure in-core nuclear power in measurement mode and to monitor thermal environments in compensation mode. The sensor dynamics was analyzed in compensation mode to determine the environmental temperature and the heat transfer coefficient. Previous research demonstrated that a first order dynamic model is not sufficient to simulate sensor

Core management and fuel handling for nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2004-01-01

This Safety Guide supplements and elaborates upon the safety requirements for core management and fuel handling that are presented in Section 5 of the Safety Requirements publication on the operation of nuclear power plants. The present publication supersedes the IAEA Safety Guide on Safety Aspects of Core Management and Fuel Handling, issued in 1985 as Safety Series No. 50-SG-010. It is also related to the Safety Guide on the Operating Organization for Nuclear Power Plants, which identifies fuel management as one of the various functions to be performed by the operating organization. The purpose of this Safety Guide is to provide recommendations for core management and fuel handling at nuclear power plants on the basis of current international good practice. The present Safety Guide addresses those aspects of fuel management activities that are necessary in order to allow optimum reactor core operation without compromising the limits imposed by the design safety considerations relating to the nuclear fuel and the plant as a whole. In this publication, 'core management' refers to those activities that are associated with fuel management in the core and reactivity control, and 'fuel handling' refers to the movement, storage and control of fresh and irradiated fuel. Fuel management comprises both core management and fuel handling. This Safety Guide deals with fuel management for all types of land based stationary thermal neutron power plants. It describes the safety objectives of core management, the tasks that have to be accomplished to meet these objectives and the activities undertaken to perform those tasks. It also deals with the receipt of fresh fuel, storage and handling of fuel and other core components, the loading and unloading of fuel and core components, and the insertion and removal of other reactor materials. In addition, it deals with loading a transport container with irradiated fuel and its preparation for transport off the site. Transport

Core management and fuel handling for nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2002-01-01

This Safety Guide supplements and elaborates upon the safety requirements for core management and fuel handling that are presented in Section 5 of the Safety Requirements publication on the operation of nuclear power plants. The present publication supersedes the IAEA Safety Guide on Safety Aspects of Core Management and Fuel Handling, issued in 1985 as Safety Series No. 50-SG-010. It is also related to the Safety Guide on the Operating Organization for Nuclear Power Plants, which identifies fuel management as one of the various functions to be performed by the operating organization. The purpose of this Safety Guide is to provide recommendations for core management and fuel handling at nuclear power plants on the basis of current international good practice. The present Safety Guide addresses those aspects of fuel management activities that are necessary in order to allow optimum reactor core operation without compromising the limits imposed by the design safety considerations relating to the nuclear fuel and the plant as a whole. In this publication, 'core management' refers to those activities that are associated with fuel management in the core and reactivity control, and 'fuel handling' refers to the movement, storage and control of fresh and irradiated fuel. Fuel management comprises both core management and fuel handling. This Safety Guide deals with fuel management for all types of land based stationary thermal neutron power plants. It describes the safety objectives of core management, the tasks that have to be accomplished to meet these objectives and the activities undertaken to perform those tasks. It also deals with the receipt of fresh fuel, storage and handling of fuel and other core components, the loading and unloading of fuel and core components, and the insertion and removal of other reactor materials. In addition, it deals with loading a transport container with irradiated fuel and its preparation for transport off the site. Transport

Modelling of core protection and monitoring system for PWR nuclear power plant simulator

International Nuclear Information System (INIS)

Jung Kun Lee; Byoung Sung Han

1997-01-01

A nuclear power plant simulator was developed for Younggwang units 3 and 4 nuclear power plant (YGN Nos 3 and 4) in Korea; it has been in operation on training center since November 1996. The core protection calculator (CPC) and the core operating limit supervisory system (COLSS) for the simulator were also developed. The CPC is a digital computer-based core protection system, which performs on-line calculation of departure from nucleate boiling ratio (DNBR) and local power density (LPD). It initiates reactor trip when the core conditions exceed designated DNBR or LPD limitations. The COLSS is designed to assist operators by implementing the limiting conditions for operations in the technical specifications. With these systems, it is possible to increase capacity factor and safety of nuclear power plants, because the COLSS data can show accurate operation margin to plant operators and the CPC can protect reactor core. In this study, the function of CPC/COLSS is analyzed in detail, and then simulation model for CPC/COLSS is presented based on the function. Compared with the YGN Nos 3 and 4 plant operation data and CEDIPS/COLSS FORTRAN code test results, the predictions with the model show reasonable results. (Author)

Large Core Three Branch Polymer Power Splitters

Directory of Open Access Journals (Sweden)

V. Prajzler

2015-12-01

Full Text Available We report about three branch large core polymer power splitters optimized for connecting standard plastic optical fibers. A new point of the design is insertion of a rectangle-shaped spacing between the input and the central part of the splitter, which will ensure more even distribution of the output optical power. The splitters were designed by beam propagation method using BeamPROP software. Acrylic-based polymers were used as optical waveguides being poured into the Y-grooves realized by computer numerical controlled engraving on poly(methyl methacrylate substrate. Measurement of the optical insertion losses proved that the insertion optical loss could be lowered to 2.1 dB at 650 nm and optical power coupling ratio could reach 31.8% : 37.3% : 30.9%.

Whole core calculations of power reactors by Monte Carlo method

International Nuclear Information System (INIS)

Nakagawa, Masayuki; Mori, Takamasa

1993-01-01

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

On-line generation of three-dimensional core power distribution using incore detector signals to monitor safety limits

International Nuclear Information System (INIS)

Jang, Jin Wook; Lee, Ki Bog; Na, Man Gyun; Lee, Yoon Joon

2004-01-01

It is essential in commercial reactors that the safety limits imposed on the fuel pellets and fuel clad barriers, such as the Linear Power Density (LPD) and the Departure from Nucleate Boiling Ratio (DNBR), are not violated during reactor operations. In order to accurately monitor the safety limits of current reactor states, a detailed three-dimensional (3D) core power distribution should be estimated from the in-core detector signals. In this paper, we propose a calculation methodology for detailed 3D core power distribution, using in-core detector signals and core monitoring constants such as the 3D Coupling Coefficients (3DCC), node power fraction, and pin-to-node factors. Also, the calculation method for several core safety parameters is introduced. The core monitoring constants for the real core state are promptly provided by the core design code and on-line MASTER(Multi-purpose Analyzer for Static and Transient Effects of Reactors), coupled with the core monitoring program. Through the plant computer, core state variables, which include reactor thermal power, control rod bank position, boron concentration, inlet moderator temperature, and flow rate, are supplied as input data for MASTER. MASTER performs the core calculation based on the neutron balance equation and generates several core monitoring constants corresponding to the real core state in addition to the expected core power distribution. The accuracy of the developed method is verified through a comparison with the current CECOR method. Because in all the verification calculation cases the proposed method shows a more conservative value than the best estimated value and a less conservative one than the current CECOR and COLSS methods, it is also confirmed that this method secures a greater operating margin through the simulation of the YGN-3 cycle-1 core from the viewpoint of the power peaking factor for the LPD and the pseudo hot pin axial power distribution for the DNBR calculation

Approach to improve the axial power distribution for the application of a core protection system

International Nuclear Information System (INIS)

Koo, Bon Seung; Cho, Jin Young; Song, Jae Seung; Lee, Chung Chan

2008-01-01

A Core Protection Calculator System (CPCS) is a digital computer based on a safety system for generating trip signals based on a calculation of the Departure from Nucleate Boiling Ratio (DNBR) and the Local Power Density (LPD) by using several on-line measured system parameters including 3-level ex-core detector signals. A few approaches to improve the axial power distribution for the application of a core protection system were performed. For the Yonggwang unit 3 (cycle 1), axial power distributions were synthesized by applying the cubic spline method and compared with the neutronics code results. Several new cubic spline function sets were generated for the drastically distorted axial shapes for a 3-level ex-core detector system. In addition, synthesized axial shapes with a 5-level ex-core detector signals were compared with the conventional 3-level detector results. It demonstrates that the newly generated function sets appear to be better than that of the conventional CPC from the aspect of an axial power synthesis, particularly for the heavily distorted shapes. Moreover, synthesis of an axial power distribution using 5-level ex-core detector signals appears to be better than that of the 3-level ex-core detector signals. From the above results, improvement of the thermal margin is expected because of an uncertainty decreasing a core protection system. (authors)

A heuristic application of critical power ratio to pressurized water reactor core design

International Nuclear Information System (INIS)

Ahn, Seung Hoon; Jeun, Gyoo Dong

2002-01-01

The approach for evaluating the critical heat flux (CHF) margin using the departure from nucleate boiling ratio (DNBR) concept has been widely applied to PWR core design, while DNBR in this approach does not indicate appropriately the CHF margin in terms of the attainable power margin-to-CHF against a reactor core condition. The CHF power margin must be calculated by increasing power until the minimum DNBR reaches a DNBR limit. The Critical Power Ratio (CPR), defined as the ratio of the predicted CHF power to the operating power, is considered more reasonable for indicating the CHF margin and can be calculated by a CPR correlation based on the heat balance of a test bundle. This approach yields directly the CHF power margin, but the calculated CPR must be corrected to compensate for many local effects of the actual core, which are not considered in the CHF test and analysis. In this paper, correction of the calculated CPR is made so that it may become equal to the DNB overpower margin. Exemplary calculations showed that the correction tends to be increased as power distribution is more distorted, but are not unduly large

Self powered neutron detectors as in-core detectors for Sodium-cooled Fast Reactors

Science.gov (United States)

Verma, V.; Barbot, L.; Filliatre, P.; Hellesen, C.; Jammes, C.; Svärd, S. Jacobsson

2017-07-01

Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor. Diverse possibilities of detector system installation must be studied for various locations in the reactor vessel in order to detect any perturbations in the core. Results from a previous paper indicated that it is possible to detect changes in neutron source distribution initiated by an inadvertent withdrawal of outer control rod with in-vessel fission chambers located azimuthally around the core. It is, however, not possible to follow inner control rod withdrawal and precisely know the location of the perturbation in the core. Hence the use of complimentary in-core detectors coupled with the peripheral fission chambers is proposed to enable robust core monitoring across the radial direction. In this paper, we assess the feasibility of using self-powered neutron detectors (SPNDs) as in-core detectors in fast reactors for detecting local changes in the power distribution when the reactor is operated at nominal power. We study the neutron and gamma contributions to the total output current of the detector modelled with Platinum as the emitter material. It is shown that this SPND placed in an SFR-like environment would give a sufficiently measurable prompt neutron induced current of the order of 600 nA/m. The corresponding induced current in the connecting cable is two orders of magnitude lower and can be neglected. This means that the SPND can follow in-core power fluctuations. This validates the operability of an SPND in an SFR-like environment.

Measurement of power distribution in FCA-HCLWR core (Phase-1)

International Nuclear Information System (INIS)

Ohno, Akio; Osugi, Toshitaka; Satoh, Kunio

1991-11-01

In the report are described experiments with zone-type mock-up cores (FCA XIV) which consisted of uranium fuels and polystilene plates, assembled at FCA (Fast Critical Assembly), to study a series of physical characteristics for high conversion light water reactors. As one example of those characteristics, power distributions were measured by a γ-counting method in the mock-up cores by changing parametrically voidage states of the moderator, volume ratio of moderator to fuel and fuel enrichment. Fine structures of fission rate in a plate-array cell having strong heterogeneities were obtained at the center cell of the core to investigate the validity of the SRAC code for the analysis of the high conversion light water reactor. Furthermore infinite multiplication factors K ∞ which are an important physical parameter were derived from calculated migration areas and bucklings of each direction. This was obtained by fitting the measured power distributions into a cosine function. Calculated power distributions of radial directions overestimate largely the measure ones, while those of axial directions agree well with the measured values. Calculations on fine structure of fission ratio in the cell follow generally the measured values, but it is recognized that the calculation underestimates the measurement in a soft neutron spectrum core. As for infinite multiplication factors K ∞ , calculated values by the SRAC code agree within experimental errors with measured ones. No trend is observed on different voidage state of moderator and fuel enrichment in the limit of this experiment. (author)

Comparison of the fractional power motor with cores made of various magnetic materials

Science.gov (United States)

Gmyrek, Zbigniew; Lefik, Marcin; Cavagnino, Andrea; Ferraris, Luca

2017-12-01

The optimization of the motor cores, coupled with new core shapes as well as powering the motor at high frequency are the primary reasons for the use of new materials. The utilization of new materials, like SMC (soft magnetic composite), reduce the core loss and/or provide quasi-isotropic core's properties in any magnetization direction. Moreover, the use of SMC materials allows for avoiding degradation of the material portions, resulting from punching process, thereby preventing the deterioration of operating parameters of the motor. The authors examine the impact of technological parameters on the properties of a new type of SMC material and analyze the possibility of its use as the core of the fractional power motor. The result of the work is an indication of the shape of the rotor core made of a new SMC material to achieve operational parameters similar to those that have a motor with a core made of laminations.

Method of estimating thermal power distribution of core of BWR type reactor

International Nuclear Information System (INIS)

Sekimizu, Koichi

1982-01-01

Purpose: To accurately and rapidly predict the thermal power of the core of a BWR they reactor at load follow-up operating time. Method: A parameter value corrected from a correction coefficient deciding unit and a xenon density distribution value predicted and calculated from a xenon density distributor are inputted to a thermal power distribution predicting devise, the status amount such as coolant flow rate or the like predetermined at this and next high power operating times is substituted for physical model to predict and calculate the thermal power distribution. The status amount of a nuclear reactor at the time of operating in previous high power corresponding to the next high power operation to be predicted is read from the status amount of the reactor stored in time series manner is a reactor core status memory, and the physical model used in the prediction and calculation of the thermal power distribution at the time of next high power operation is corrected. (Sikiya, K.)

Power distribution investigation in the transition phase of the low moderation type MOX fueled LWR from the high conversion core to the breeding core

International Nuclear Information System (INIS)

Akie, Hiroshi; Nakano, Yoshihiro; Okubo, Tsutomu

2011-01-01

The key concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) is a core transition from a high conversion (HC) type to a plutonium breeding (BR) type in a same reactor system only by replacing fuel assemblies. Consequently in this transition phase, there are two types of assemblies in the same core. Due to the differences of the two assembly types, region-wise soft to hard neutron spectra appears and result in a large power peaking. Therefore, power distribution of FLWR in the HC to BR transition phase was studied by performing assembly and core calculations. For the whole core calculation, a new 14-group energy structure is developed to better represent the power distribution obtained with the fine 107-group structure than the 9-group structure in the previous evaluations. Calculations on few assemblies geometries show large local power peakings can be effectively reduced by considering plutonium enrichment distribution in an assembly. In the whole core calculation, there is a power level mismatch between HC and BR assemblies, but overall power distribution flattening is possible by optimizing fuel assemblies loading. Although the fuel loading should be decided also taking into account the void coefficient, transition from HC to BR type FLWR seems feasible without difficulty. (author)

Advanced in-core monitoring system for high-power reactors

International Nuclear Information System (INIS)

Mitin, V.I.; Alekseev, A.N.; Golovanov, M.N.; Zorin, A.V.; Kalinushkin, A.E.; Kovel, A.I.; Milto, N.V.; Musikhin, A.M.; Tikhonova, N.V.; Filatov, V.P.

2006-01-01

This paper encompasses such section as objective, conception and engineering solution for construction of advanced in-core instrumentation system for high power reactor, including WWER-1000. The ICIS main task is known to be an on-line monitoring of power distribution and functionals independently of design programs to avoid a common cause error. This paper shows in what way the recovery of power distribution has been carried out using the signals from in-core neutron detectors or temperature sensors. On the basis of both measured and processed data, the signals of preventive and emergency protection on local parameters (linear power of the maximum intensive fuel rods, departure from nucleate boiling ratio peaking factor) have been automatically generated. The paper presents a detection technology and processing methods for signals from SPNDs and TCs, ICIS composition and structure, computer hardware, system and applied software. Structure, composition and the taken decisions allow combining class IE and class B and C tasks in accordance with international standards of separation and safety category realization. Nowadays, ICIS-M is a system that is capable to ensure: monitoring, safety, information display and diagnostics function, which allow securing actual increase of quality, reliability and safety in operation of nuclear fuel and power units. Meanwhile, it reduce negative influence of human factor on thermal technical reliability in the operational process (Authors)

Fundamental design bases for independent core cooling in Swedish nuclear power reactors

International Nuclear Information System (INIS)

Jelinek, Tomas

2015-01-01

New regulations on design and construction of nuclear power plants came into force in 2005. The need of an independent core cooling system and if the regulations should include such a requirement was discussed. The Swedish Radiation Safety authority (SSM) decided to not include such a requirement because of open questions about the water balance and started to investigate the consequences of an independent core cooling system. The investigation is now finished and SSM is also looking at the lessons learned from the accident in Fukushima 2011. One of the most important measures in the Swedish national action plan is the implementation of an independent core cooling function for all Swedish power plants. SSM has investigated the basic design criteria for such a function where some important questions are the level of defence in depth and the acceptance criteria. There is also a question about independence between the levels of defence in depth that SSM have included in the criteria. Another issue that has to be taken into account is the complexity of the system and the need of automation where independence and simplicity are very strong criteria. In the beginning of 2014 a memorandum was finalized regarding fundamental design bases for independent core cooling in Swedish nuclear power reactors. A decision based on this memorandum with an implementation plan will be made in the first half of 2014. Sweden is also investigating the possibility to have armed personnel on site, which is not allowed currently. The result from the investigation will have impact on the possibility to use mobile equipment and the level of protection of permanent equipment. In this paper, SSM will present the memorandum for design bases for independent core cooling in Swedish nuclear power reactors that was finalized in March 20147 that also describe SSM's position regarding independence and automation of the independent core cooling function. This memorandum describes the Swedish

Reconstruction calculation of pin power for ship reactor core

International Nuclear Information System (INIS)

Li Haofeng; Shang Xueli; Chen Wenzhen; Wang Qiao

2010-01-01

Aiming at the limitation of the software that pin power distribution for ship reactor core was unavailable, the calculation model and method of the axial and radial pin power distribution were proposed. Reconstruction calculations of pin power along axis and radius was carried out by bicubic and bilinear interpolation and cubic spline interpolation, respectively. The results were compared with those obtained by professional reactor physical soft with fine mesh difference. It is shown that our reconstruction calculation of pin power is simple and reliable as well as accurate, which provides an important theoretic base for the safety analysis and operating administration of the ship nuclear reactor. (authors)

A heterogeneous multi-core platform for low power signal processing in systems-on-chip

DEFF Research Database (Denmark)

Paker, Ozgun; Sparsø, Jens; Haandbæk, Niels

2002-01-01

is based on message passing. The mini-cores are designed as parameterized soft macros intended for a synthesis based design flow. A 520.000 transistor 0.25µm CMOS prototype chip containing 6 mini-cores has been fabricated and tested. Its power consumption is only 50% higher than a hardwired ASIC and more......This paper presents a low-power and programmable DSP architecture - a heterogeneous multiprocessor platform consisting of standard CPU/DSP cores, and a set of simple instruction set processors called mini-cores each optimized for a particular class of algorithm (FIR, IIR, LMS, etc.). Communication...

Comparison of the fractional power motor with cores made of various magnetic materials

Directory of Open Access Journals (Sweden)

Gmyrek Zbigniew

2017-12-01

Full Text Available The optimization of the motor cores, coupled with new core shapes as well as powering the motor at high frequency are the primary reasons for the use of new materials. The utilization of new materials, like SMC (soft magnetic composite, reduce the core loss and/or provide quasi-isotropic core’s properties in any magnetization direction. Moreover, the use of SMC materials allows for avoiding degradation of the material portions, resulting from punching process, thereby preventing the deterioration of operating parameters of the motor. The authors examine the impact of technological parameters on the properties of a new type of SMC material and analyze the possibility of its use as the core of the fractional power motor. The result of the work is an indication of the shape of the rotor core made of a new SMC material to achieve operational parameters similar to those that have a motor with a core made of laminations.

Effects of Radial Reflector Composition on Core Reactivity and Peak Power

International Nuclear Information System (INIS)

Park, Sang Yoon; Lee, Kyung Hoon; Song, Jae Seung

2007-10-01

The effects of radial SA-240 alloy shroud on core reactivity and peak power are evaluated. The existence of radial SA-240 alloy shroud makes reflector water volume decrease, so the thermal absorption cross section of radial reflector is lower than without SA-240 alloy shroud case. Finally, the cycle length is increased from 788 EFPD to 845 EFPD and the peak power is decreased from 1.66 to 1.49. In the case of without SA-240 alloy shroud, a new core loading pattern search has been performed. For the guarantee of the same equivalent cycle length of with SA-240 alloy shroud case, the enrichment of U-235 should be increased from 4.22 w/o to 4.68 w/o. The nuclear key safety parameters of new core loading pattern have been calculated and recorded for the future

Effects of Radial Reflector Composition on Core Reactivity and Peak Power

Energy Technology Data Exchange (ETDEWEB)

Park, Sang Yoon; Lee, Kyung Hoon; Song, Jae Seung

2007-10-15

The effects of radial SA-240 alloy shroud on core reactivity and peak power are evaluated. The existence of radial SA-240 alloy shroud makes reflector water volume decrease, so the thermal absorption cross section of radial reflector is lower than without SA-240 alloy shroud case. Finally, the cycle length is increased from 788 EFPD to 845 EFPD and the peak power is decreased from 1.66 to 1.49. In the case of without SA-240 alloy shroud, a new core loading pattern search has been performed. For the guarantee of the same equivalent cycle length of with SA-240 alloy shroud case, the enrichment of U-235 should be increased from 4.22 w/o to 4.68 w/o. The nuclear key safety parameters of new core loading pattern have been calculated and recorded for the future.

Analysis of the critical and first full power operating cores for PARR using leu oxide fuel

International Nuclear Information System (INIS)

Khan, L.A.; Qazi, M.K.; Bokhari, I.H.; Fazal, R.

1989-10-01

This paper explains the analysis for determining the first full power operating core for PARR using LEU oxide fuel. The core configuration selected for this first full power operation contains about 6.13 kg of U-235 distributed in 19 standard and five control fuel elements. The neutron flux level is doubled when core is shifted from 5MW to 10 MW. Total nuclear power peaking factor of the core is 2.03. The analysis shows that the core can be operated safely at 5 MW with a flow rate of 520 meter cube per hour and at 10 MW with a flow rate of 900 meter cube per hour. (A.B.). 10 figs

Operation manual for the core flow test loop zone power-supply controller

Energy Technology Data Exchange (ETDEWEB)

Harper, R.E.

1981-11-01

The core flow test loop, which is part of the Gas-Cooled Fast Breeder Reactor Program (GCFR) at ORNL, is a high-pressure, high-temperature, out-of-reactor helium circulation system that is being constructed to permit study of the performance at steady-state and transient conditions of simulated segments of core assemblies for a GCFR demonstration plant. The simulated core segments, which are divided into zones, contain electrical heating elements to simulate the heat generated by fission. To control the power which is applied to a zone, a novel multitapped transformer and zone power control system have been designed and built which satisfy stringent design criteria. The controller can match power output to demand to within better than +-1% over a 900:1 dynamic range and perform full-power transients within 1 s. The power is applied in such a way as to minimize the electromagnetic interference at the bandwidth of the loop instrumentation, and the controller incorporates several error detection techniques, making it inherently fail-safe. The operation manual describes the specifications, operating instructions, error detection capabilities, error recovery, troubleshooting, calibration and QA procedures, and maintenance requirements. Also included are sections on the theory of operation, circuitry description, and a complete set of schematics.

Core design and fuel rod analyses of a super fast reactor with high power density

International Nuclear Information System (INIS)

Ju, Haitao; Cao, Liangzhi; Lu, Haoliang; Oka, Yoshiaki; Ikejiri, Satoshi; Ishiwatari, Yuki

2009-01-01

A Super Fast Reactor is a pressure-vessel type, fast spectrum SuperCritical Water Reactor (SCWR) that is presently researched in a Japanese project. One of the most important advantages of the Super Fast Reactor is the higher power density compared to the thermal spectrum SCWR, which reduces the capital cost. A preliminary core has an average power density of 158.8W/cc. In this paper, the principle of improving the average power density is studied and the core design is improved. After the sensitivity analyses on the fuel rod configurations, the fuel assembly configurations and the core configurations, an improved core with an average power density of 294.8W/cc is designed by 3-D neutronic/thermal-hydraulic coupled calculations. This power density is competitive with that of typical Liquid Metal Fast Breeder Reactors (LMFBR). In order to ensure the fuel rod integrity of this core design, the fuel rod behaviors on the normal operating condition are analyzed using FEMAXI-6 code. The power histories of each fuel rod are taken from the neutronics calculation results in the core design. The cladding surface temperature histories are taken from the thermal-hydraulic calculation results in the core design. Four types of the limiting fuel rods, with the Maximum Cladding Surface Temperature (MCST), Maximum Power Peak(MPP), Maximum Discharge Burnup(MDB) and Different Coolant Flow Pattern (DCFP), are chosen to cover all the fuel rods in the core. The available design range of the fuel rod design parameters, such as initial gas plenum pressure, gas plenum position, gas plenum length, grain size and gap size, are found out in order to satisfy the following design criteria: (1) Maximum fuel centerline temperature should be less than 1900degC. (2) Maximum cladding stress in circumstance direction should be less than 100MPa. (3) Pressure difference on the cladding should be less than 1/3 of buckling collapse pressure. (4) Cumulative damage faction (CDF) of the cladding should be

Core-Shell-Yarn-Based Triboelectric Nanogenerator Textiles as Power Cloths.

Science.gov (United States)

Yu, Aifang; Pu, Xiong; Wen, Rongmei; Liu, Mengmeng; Zhou, Tao; Zhang, Ke; Zhang, Yang; Zhai, Junyi; Hu, Weiguo; Wang, Zhong Lin

2017-12-26

Although textile-based triboelectric nanogenerators (TENGs) are highly promising because they scavenge energy from their working environment to sustainably power wearable/mobile electronics, the challenge of simultaneously possessing the qualities of cloth remains. In this work, we propose a strategy for TENG textiles as power cloths in which core-shell yarns with core conductive fibers as the electrode and artificial polymer fibers or natural fibrous materials tightly twined around core conductive fibers are applied as the building blocks. The resulting TENG textiles are comfortable, flexible, and fashionable, and their production processes are compatible with industrial, large-scale textile manufacturing. More importantly, the comfortable TENG textiles demonstrate excellent washability and tailorability and can be fully applied in further garment processing. TENG textiles worn under the arm or foot have also been demonstrated to scavenge various types of energy from human motion, such as patting, walking, and running. All of these merits of proposed TENG textiles for clothing uses suggest their great potentials for viable applications in wearable electronics or smart textiles in the near future.

Core-powered mass-loss and the radius distribution of small exoplanets

Science.gov (United States)

Ginzburg, Sivan; Schlichting, Hilke E.; Sari, Re'em

2018-05-01

Recent observations identify a valley in the radius distribution of small exoplanets, with planets in the range 1.5-2.0 R⊕ significantly less common than somewhat smaller or larger planets. This valley may suggest a bimodal population of rocky planets that are either engulfed by massive gas envelopes that significantly enlarge their radius, or do not have detectable atmospheres at all. One explanation of such a bimodal distribution is atmospheric erosion by high-energy stellar photons. We investigate an alternative mechanism: the luminosity of the cooling rocky core, which can completely erode light envelopes while preserving heavy ones, produces a deficit of intermediate sized planets. We evolve planetary populations that are derived from observations using a simple analytical prescription, accounting self-consistently for envelope accretion, cooling and mass-loss, and demonstrate that core-powered mass-loss naturally reproduces the observed radius distribution, regardless of the high-energy incident flux. Observations of planets around different stellar types may distinguish between photoevaporation, which is powered by the high-energy tail of the stellar radiation, and core-powered mass-loss, which depends on the bolometric flux through the planet's equilibrium temperature that sets both its cooling and mass-loss rates.

Management of radioactive waste from a major core damage in a BWR power plant

International Nuclear Information System (INIS)

Elkert, J.; Christensen, H.; Torstenfelt, B.

1990-01-01

Large amounts of fission products would be released in case of a major core damage in a nuclear power reactor. In this theoretical study the core damage is caused by a loss of coolant accident followed by a complete loss of all electric power for about 30 minutes resulting in the release of 10% of the core inventory of noble gases. A second case has also been briefly studied, in which the corresponding core damage is supposed to be created merely by the complete loss of electric power during a limited time period. It appears from the study that the radioactive waste generated as a consequence of an accident of the extent can be managed in the reference reactor with only minor modifications required in the waste plant. The detailed results of the study are reactor specific, but many of the findings and recommendations are generally applicable. (author) 28 refs

Design and fabrication of self-powered in-core neutron flux monitor assembly

International Nuclear Information System (INIS)

Chung, M.K.; Cho, S.W.; Kang, H.D.; Cho, K.K.; Cho, B.S.; Kang, S.S.

1980-01-01

This is the final report on the prototypical fabrication of an in-core neutron flux monitor detector assembly for a specific power reactor conducted by KAERI from July 1, 1978 to December 31, 1979. It is well known that power reactors require a large number of in-core neutron flux detector for reactor regulation and the structures of detector assemblies are different from reactor to reactor. Therefore, from the nature of this project, it should be noted here that the target model of the prototypical farbrication of an in-core neutron flux monitor detector assembly is a VFD-2 System for Wolsung CANDU. It is concluded that fabrication of in-core neutron flux monitor detector assembly for CANDU reactor is technically feasible and will bring economical benefit as much as 50 % of the unit price if they are fabricated in Korea by using partially materials which are available from local market. (author)

Application of MSHIM core control strategy for westinghouse AP1000 nuclear power plant

International Nuclear Information System (INIS)

Onoue, Masaaki; Kawanishi, Tomohiro; Carlson, William R.; Morita, Toshio

2003-01-01

Westinghouse has developed a new core control strategy, in which two independently moving Rod Cluster Control Assembly (RCCA) groups are utilized for core control; one group for reactivity/temperature control, the other for axial power distribution (Axial Offset) control. This control procedure eliminates the need for Chemical Shim adjustments during power maneuvers, such as load follow, and is designated MSHIM (Mechanical Shim). This core control strategy is utilized in the AP1000. In the AP1000, it is possible to perform MSHIM load follow maneuvers for up to 95% of cycle life without changing the soluble boron concentration in the moderator. This core control strategy has been evaluated, via computer simulations, to provide appropriate margins to core and fuel design limits during normal operation maneuvers (including load follow operations) and also during anticipated Condition II accident transients. The primary benefits of MSHIM as a control strategy are as follows; Power change operation can be totally automated due to the elimination of boron concentration adjustments. Full load follow capability is achievable for up to more than 95% of cycle life. Load follow operations performed solely by mechanical devices results in a significant reduction in the boron system requirements and a significant reduction in daily effluent to be processed. (author)

Effective Domain Partitioning for Multi-Clock Domain IP Core Wrapper Design under Power Constraints

Science.gov (United States)

Yu, Thomas Edison; Yoneda, Tomokazu; Zhao, Danella; Fujiwara, Hideo

The rapid advancement of VLSI technology has made it possible for chip designers and manufacturers to embed the components of a whole system onto a single chip, called System-on-Chip or SoC. SoCs make use of pre-designed modules, called IP-cores, which provide faster design time and quicker time-to-market. Furthermore, SoCs that operate at multiple clock domains and very low power requirements are being utilized in the latest communications, networking and signal processing devices. As a result, the testing of SoCs and multi-clock domain embedded cores under power constraints has been rapidly gaining importance. In this research, a novel method for designing power-aware test wrappers for embedded cores with multiple clock domains is presented. By effectively partitioning the various clock domains, we are able to increase the solution space of possible test schedules for the core. Since previous methods were limited to concurrently testing all the clock domains, we effectively remove this limitation by making use of bandwidth conversion, multiple shift frequencies and properly gating the clock signals to control the shift activity of various core logic elements. The combination of the above techniques gains us greater flexibility when determining an optimal test schedule under very tight power constraints. Furthermore, since it is computationally intensive to search the entire expanded solution space for the possible test schedules, we propose a heuristic 3-D bin packing algorithm to determine the optimal wrapper architecture and test schedule while minimizing the test time under power and bandwidth constraints.

Ultrahigh temperature vapor core reactor-MHD system for space nuclear electric power

Science.gov (United States)

Maya, Isaac; Anghaie, Samim; Diaz, Nils J.; Dugan, Edward T.

1991-01-01

The conceptual design of a nuclear space power system based on the ultrahigh temperature vapor core reactor with MHD energy conversion is presented. This UF4 fueled gas core cavity reactor operates at 4000 K maximum core temperature and 40 atm. Materials experiments, conducted with UF4 up to 2200 K, demonstrate acceptable compatibility with tungsten-molybdenum-, and carbon-based materials. The supporting nuclear, heat transfer, fluid flow and MHD analysis, and fissioning plasma physics experiments are also discussed.

Results and interpretation of noise measurements using in-core self powered neutron detector strings at Unit 2 of the Paks Nuclear Power Plant

International Nuclear Information System (INIS)

Gloeckler, O.; Por, G.; Valko, J.

1986-11-01

In-core neutron noise and fuel assembly outlet temperature noise measurements were performed at Unit 2 of Paks Nuclear Power Plant. Characteristics of the reactor and the noise measuring equipment are briefly described. The in-core Rhodium emitter selfpowered neutron detector strings positioned axially above the other show high coherence and linear phase at low frequencies indicating a marked transport effect, not regularly measured in PWRs. The coherence between horizontally placed neutron detectors is small and the phase is zero. A transport effect of different nature is obtained between neutron detectors (in-core and ex-core) and fuel assembly outlet thermocouples. The observed characteristics depend on reactor and fuel assembly power in a way supporting interpretation in terms of coolant density and void content changes and power feedback effects. During routine analysis vibration of 1.1 Hz appeared as a strong peak in the power spectra. The control assembly that was responsible for the observed behaviour could be localized with high certainty. (author)

Thermal limits validation of gamma thermometer power adaption in CFE Laguna Verde 2 reactor core

Energy Technology Data Exchange (ETDEWEB)

Cuevas V, G.; Banfield, J. [GE-Hitachi Nuclear Energy Americas LLC, Global Nuclear Fuel, Americas LLC, 3901 Castle Hayne Road, Wilmingtonm, North Carolina (United States); Avila N, A., E-mail: Gabriel.Cuevas-Vivas@ge.com [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Carretera Cardel-Nautla Km 42.5, Alto Lucero, Veracruz (Mexico)

2016-09-15

This paper presents the status of GEH work on Gamma Thermometer (GT) validation using the signals of the instruments installed in the Laguna Verde Unit 2 reactor core. The long-standing technical collaboration between Comision Federal de Electricidad (CFE), Global Nuclear Fuel - Americas LLC (GNF) and GE-Hitachi Nuclear Energy Americas LLC (GEH) is moving forward with solid steps to a final implementation of GTs in a nuclear reactor core. Each GT is integrated into a slightly modified Local Power Range Monitor (LPRM) assembly. Six instrumentation strings are equipped with two gamma field detectors for a total of twenty-four bundles whose calculated powers are adapted to the instrumentation readings in addition to their use as calibration instruments for LPRMs. Since November 2007, the six GT instrumentation strings have been operable with almost no degradation by the strong neutron and gamma fluxes in the Laguna Verde Unit 2 reactor core. In this paper, the thermal limits, Critical Power Ratio (CPR) and maximum Linear Heat Generation Rate (LHGR), of bundles directly monitored by either Traverse In-core Probes (TIPs) or GTs are used to establish validation results that confirm the viability of TIP system replacement with automatic fixed in-core probes (AFIPs, GTs, in a Boiling Water Reactor. The new GNF steady-state reactor core simulator AETNA02 is used to obtain power and exposure distribution. Using this code with an updated methodology for GT power adaption, a reduced value of the GT interpolation uncertainty is obtained that is fed into the LHGR calculation. This new method achieves margin recovery for the adapted thermal limits for use in the Economic Simplified Boiling Water Reactor (ESBWR) or any other BWR in the future that employs a GT based AFIP system for local power measurements. (Author)

Thermal limits validation of gamma thermometer power adaption in CFE Laguna Verde 2 reactor core

International Nuclear Information System (INIS)

Cuevas V, G.; Banfield, J.; Avila N, A.

2016-09-01

This paper presents the status of GEH work on Gamma Thermometer (GT) validation using the signals of the instruments installed in the Laguna Verde Unit 2 reactor core. The long-standing technical collaboration between Comision Federal de Electricidad (CFE), Global Nuclear Fuel - Americas LLC (GNF) and GE-Hitachi Nuclear Energy Americas LLC (GEH) is moving forward with solid steps to a final implementation of GTs in a nuclear reactor core. Each GT is integrated into a slightly modified Local Power Range Monitor (LPRM) assembly. Six instrumentation strings are equipped with two gamma field detectors for a total of twenty-four bundles whose calculated powers are adapted to the instrumentation readings in addition to their use as calibration instruments for LPRMs. Since November 2007, the six GT instrumentation strings have been operable with almost no degradation by the strong neutron and gamma fluxes in the Laguna Verde Unit 2 reactor core. In this paper, the thermal limits, Critical Power Ratio (CPR) and maximum Linear Heat Generation Rate (LHGR), of bundles directly monitored by either Traverse In-core Probes (TIPs) or GTs are used to establish validation results that confirm the viability of TIP system replacement with automatic fixed in-core probes (AFIPs, GTs, in a Boiling Water Reactor. The new GNF steady-state reactor core simulator AETNA02 is used to obtain power and exposure distribution. Using this code with an updated methodology for GT power adaption, a reduced value of the GT interpolation uncertainty is obtained that is fed into the LHGR calculation. This new method achieves margin recovery for the adapted thermal limits for use in the Economic Simplified Boiling Water Reactor (ESBWR) or any other BWR in the future that employs a GT based AFIP system for local power measurements. (Author)

Solution of the 'MIDICORE' WWER-1000 core periphery power distribution benchmark by KARATE and MCNP

International Nuclear Information System (INIS)

Temesvari, E.; Hegyi, G.; Hordosy, G.; Maraczy, C.

2011-01-01

The 'MIDICORE' WWER-1000 core periphery power distribution benchmark was proposed by Mr. Mikolas on the twentieth Symposium of AER in Finland in 2010. This MIDICORE benchmark is a two-dimensional calculation benchmark based on the WWER-1000 reactor core cold state geometry with taking into account the geometry of explicit radial reflector. The main task of the benchmark is to test the pin by pin power distribution in selected fuel assemblies at the periphery of the WWER-1000 core. In this paper we present our results (k eff , integral fission power) calculated by MCNP and the KARATE code system in KFKI-AEKI and the comparison to the preliminary reference Monte Carlo calculation results made by NRI, Rez. (Authors)

Overview of the TITAN-II reversed-field pinch aqueous fusion power core design

Energy Technology Data Exchange (ETDEWEB)

Wong, C.P.C.; Creedon, R.L.; Grotz, S.; Cheng, E.T.; Sharafat, S.; Cooke, P.I.H.

1988-03-01

TITAN-II is a compact, high power density Reversed-Field Pinch fusion power reactor design based on the aqueous lithium solution fusion power core concept. The selected breeding and structural materials are LiNO/sub 3/ and 9-C low activation ferritic steel, respectively. TITAN-II is a viable alternative to the TITAN-I lithium self-cooled design for the Reversed-Field Pinch reactor to operate at a neutron wall loading of 18 MWm/sup 2/. Submerging the complete fusion power core and the primary loop in a large pool of cool water will minimize the probability of radioactivity release. Since the protection of the large pool integrity is the only requirement for the protection of the public, TITAN-II is a passive safety assurance design. 13 refs., 3 figs., 1 tab.

Overview of the TITAN-II reversed-field pinch aqueous fusion power core design

Energy Technology Data Exchange (ETDEWEB)

Wong, C.P.C.; Creedon, R.L.; Cheng, E.T. (General Atomic Co., San Diego, CA (USA)); Grotz, S.P.; Sharafat, S.; Cooke, P.I.H. (California Univ., Los Angeles (USA). Dept. of Mechanical, Aerospace and Nuclear Engineering; California Univ., Los Angeles, CA (USA). Inst. for Plasma and Fusion Research); TITAN Research Group

1989-04-01

TITAN-II is a compact, high-power-density Reversed-Field Pinch fusion power reactor design based on the aqueous lithium solution fusion power core concept. The selected breeding and structural materials are LiNO/sub 3/ and 9-C low activation ferritic steel, respectively. TITAN-II is a viable alternative to the TITAN-I lithium self-cooled design for the Reversed-Field Pinch reactor to operate at a neutron wall loading of 18 MW/m/sup 2/. Submerging the complete fusion power core and the primary loop in a large pool of cool water will minimize the probability of radioactivity release. Since the protection of the large pool integrity is the only requirement for the protection of the public, TITAN-II is a level 2 of passive safety assurance design. (orig.).

Neutron flux and power in RTP core-15

Energy Technology Data Exchange (ETDEWEB)

Rabir, Mohamad Hairie, E-mail: m-hairie@nuclearmalaysia.gov.my; Zin, Muhammad Rawi Md; Usang, Mark Dennis; Bayar, Abi Muttaqin Jalal; Hamzah, Na’im Syauqi Bin [Nuclear and reactor Physics Section, Nuclear Technology Center, Technical Support Division, Malaysian Nuclear Agency, Bangi, 43000 Kajang, Selangor (Malaysia)

2016-01-22

PUSPATI TRIGA Reactor achieved initial criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes. This paper describes the reactor parameters calculation for the PUSPATI TRIGA REACTOR (RTP); focusing on the application of the developed reactor 3D model for criticality calculation, analysis of power and neutron flux distribution of TRIGA core. The 3D continuous energy Monte Carlo code MCNP was used to develop a versatile and accurate full model of the TRIGA reactor. The model represents in detailed all important components of the core with literally no physical approximation. The consistency and accuracy of the developed RTP MCNP model was established by comparing calculations to the available experimental results and TRIGLAV code calculation.

Determination of power peak factor using control rods, ex-core detectors and neural networks

International Nuclear Information System (INIS)

Souza, Rose Mary Gomes do Prado

2005-01-01

This work presents a methodology based on the artificial neural network technique to predict in real time the power peak factor in a form that can be implemented in reactor protection systems. The neural network inputs were those available in the reactor protection systems, namely, the axial and quadrant power differences obtained from measured ex-core detector signals, and the position of control rods. The response of ex core detector signals was measured in experiments especially performed in the IPEN/MB-01 zero-power reactor. Several reactor states with different power density distribution were obtained by positioning the control rods in different configurations. The power distribution and its peak factor were calculated for each of these reactor states using the Citation code. The obtained results show that the power peak factor correlates well with the control rod position and the quadrant power difference, and with a lesser degree with the axial power differences. The data presented an inherent organisation and could be classified into different classes of power peak factor behaviour as a function of position of control rods, axial power difference and quadrant power difference. The RBF networks were able to identify classes and interpolate the power peak factor values. The relative error for the power peak factor estimation ranged from 0.19 % to 0.67 %, less than the one that was obtained performing a power density distribution map with in-core detectors. It was observed that the positions of control rods bear the detailed and localised information about the power density distribution, and that the axial and the quadrant power difference describe its global variations in the axial and radial directions. The results showed that the RBF and MLP networks produced similar results, and that a neural network correlation can be implemented in power reactor protection systems. (author)

Time delays between core power production and external detector response from Monte Carlo calculations

International Nuclear Information System (INIS)

Valentine, T.E.; Mihalczo, J.T.

1996-01-01

One primary concern for design of safety systems for reactors is the time response of external detectors to changes in the core. This paper describes a way to estimate the time delay between the core power production and the external detector response using Monte Carlo calculations and suggests a technique to measure the time delay. The Monte Carlo code KENO-NR was used to determine the time delay between the core power production and the external detector response for a conceptual design of the Advanced Neutron Source (ANS) reactor. The Monte Carlo estimated time delay was determined to be about 10 ms for this conceptual design of the ANS reactor

Three-dimensional Core Design of a Super Fast Reactor with a High Power Density

International Nuclear Information System (INIS)

Cao, Liangzhi; Oka, Yoshiaki; Ishiwatari, Yuki; Ikejiri, Satoshi; Ju, Haitao

2010-01-01

The SuperCritical Water-cooled Reactor (SCWR) pursues high power density to reduce its capital cost. The fast spectrum SCWR, called a super fast reactor, can be designed with a higher power density than thermal spectrum SCWR. The mechanism of increasing the average power density of the super fast reactor is studied theoretically and numerically. Some key parameters affecting the average power density, including fuel pin outer diameter, fuel pitch, power peaking factor, and the fraction of seed assemblies, are analyzed and optimized to achieve a more compact core. Based on those sensitivity analyses, a compact super fast reactor is successfully designed with an average power density of 294.8 W/cm 3 . The core characteristics are analyzed by using three-dimensional neutronics/thermal-hydraulics coupling method. Numerical results show that all of the design criteria and goals are satisfied

Comparison study on in-core neutron detector for online neutron flux mapping of research and power reactor

International Nuclear Information System (INIS)

Zareen Khan Abdul Jalil Khan; Mohd Idris Taib; Izhar Abu Husin; Nurfarhana Ayuni

2010-01-01

This paper presents the comparison study on In-Core neutron detector using for online flux mapping of Research and Power reactor. Technical description of in-core neutron also taken into consideration to identify the different characterization of neutron detector and describe on Self Power neutron detector (SPND) for online neutron flux mapping. Able to provide information on the neutron flux distribution and understand how in-core neutron detector are being used in nuclear power plant including to enable to state the principles of neutron detector. (author)

Emergency core cooling systems in CANDU nuclear power plants

International Nuclear Information System (INIS)

1981-12-01

This report contains the responses by the Advisory Committee on Nuclear Safety to three questions posed by the Atomic Energy Control Board concerning the need for Emergency Core Cooling Systems (ECCS) in CANDU nuclear power plants, the effectiveness requirement for such systems, and the extent to which experimental evidence should be available to demonstrate compliance with effectiveness standards

Core concept of fast power reactor with zero sodium void reactivity

International Nuclear Information System (INIS)

Matveev, V.I.; Chebeskov, A.N.; Krivitsky, I.Y.

1991-01-01

The paper presents a core concept of BN-800 - type fast power reactor with zero sodium void reactivity (SVR). Consideration is given to the layout-and some design features of such a core. Some considerations on the determination of the required SVR value as one of the fast reactor safety criteria in accidents with coolant boiling are presented. Some methodical considerations an the development of calculation models that give a correct description of the new core features are stated. The results of the integral SVR calculation studies are included. reactivity excursions under different scenarios of sodium boiling are estimated, some corrections into the calculated SVR value are discussed. (author)

INCA: method of analyzing in-core detector data in power reactors

International Nuclear Information System (INIS)

Ober, T.G.; Terney, W.B.; Marks, G.H.

1975-04-01

A method (INCA) is described by which signals from fixed in-core detectors are related to estimates of the three dimensional power distribution in an operating reactor core and to the maximum linear heat rate in the core. A description of the large library of data accompanying the method is provided. A detailed examination of the analytical verifications performed using the method is presented, and a summary of the uncertainty associated with the method is given. The uncertainty assigned to the maximum linear heat rate inferred by the method from operating reactor data is found to be 5.8 percent at a 95/95 confidence level. (U.S.)

A Practical Core Loss Model for Filter Inductors of Power Electronic Converters

DEFF Research Database (Denmark)

Matsumori, Hiroaki; Shimizu, Toshihisa; Wang, Xiongfei

2018-01-01

This paper proposes a core loss model for filter inductors of power electronic converters. The model allows a computationally efficient analysis on the core loss of the inductor under the square voltage excitation and the premagnetization condition. First, the core loss of the filter inductor under...... buck chopper excitation is evaluated with the proposed model and compared with the conventional methods. The comparison shows that the proposed method results in a better core loss prediction under the premagnetized condition than that of conventional alternatives. Then, the core loss of the filter...... inductor with the pulsewidth modulated inverter excitation is evaluated, which shows that the proposed model not only accurately predicts the core loss but also identifies the hysteresis loss part. These results demonstrate that the approach can further be used for the development of magnetic materials...

Direct measurement of the transition from edge to core power coupling in a light-ion helicon source

Science.gov (United States)

Piotrowicz, P. A.; Caneses, J. F.; Showers, M. A.; Green, D. L.; Goulding, R. H.; Caughman, J. B. O.; Biewer, T. M.; Rapp, J.; Ruzic, D. N.

2018-05-01

We present time-resolved measurements of an edge-to-core power transition in a light-ion (deuterium) helicon discharge in the form of infra-red camera imaging of a thin stainless steel target plate on the Proto-Material Exposure eXperiment device. The time-resolved images measure the two-dimensional distribution of power deposition in the helicon discharge. The discharge displays a mode transition characterized by a significant increase in the on-axis electron density and core power coupling, suppression of edge power coupling, and the formation of a fast-wave radial eigenmode. Although the self-consistent mechanism that drives this transition is not yet understood, the edge-to-core power transition displays characteristics that are consistent with the discharge entering a slow-wave anti-resonant regime. RF magnetic field measurements made across the plasma column, together with the power deposition results, provide direct evidence to support the suppression of the slow-wave in favor of core plasma production by the fast-wave in a light-ion helicon source.

High-power picosecond pulse delivery through hollow core photonic band gap fibers

DEFF Research Database (Denmark)

Michieletto, Mattia; Johansen, Mette Marie; Lyngsø, Jens Kristian

2016-01-01

We demonstrated robust and bend insensitive fiber delivery of high power laser with diffraction limited beam quality for two different kinds of hollow core band gap fibers. The light source for this experiment consists of ytterbium-doped double clad fiber aeroGAIN-ROD-PM85 in a high power amplifier...

Formulation of detector response function to calculate the power density profiles using in-core neutron detectors

International Nuclear Information System (INIS)

Ahmed, S. A.; Peter, J. K.; Semmler, W.; Shultis, J. K.

2007-01-01

By measuring neutron fluxes at different locations throughout a core, it's possible to derive the power-density profile P k (W cm - 3), at an axial depth z of fuel rod k. Micro-pocket fission detectors (MPFD) have been fabricated to perform such in-core neutron flux measurements. The purpose of this study is to develop a mathematical model to obtain axial power density distributions in the fuel rods from the in-core responses of the MPFDs

VG-400 atomic power and technological installation. Possible core design

International Nuclear Information System (INIS)

Komarov, E.V.; Laptev, F.V.; Lyubivyj, A.G.; Mitenkov, F.M.; Samojlov, O.B.; Sukhachevskij, Yu.B.

1979-01-01

The main characteristics, basic circuit and configuration of equipment of the VG-400 atomic power and technological installation are considered. This installation is intended for supplying with highly-potential heat of thermal electrochemical hydrogen production and for power generation in the steam-turbine cycle. The main installation characteristics: HTGR reactor heat power 1100 MW, electric power 300 MW, helium coolant pressure 50 atm, output temperature 950 deg C, steam pressure in the second contour 175 atm, temperature 535 deg C, core diameter and height 6.4 m and 4 m, respectively, number of spherical fuel elements 8.5x10 5 . The installation can ensure hydrogen production of 10 5 Nxm 3 /h. For the VG-400 reactor block the integral arrangement of the first circuit equipment in the reinforced concrete is chosen. Two versions of the reactor core with prismatic and spherical fuel elements are compared. It is shown that taking into account great potentialities of the spherical zone in a case of further temperature increase and its positive qualities with respect to construction and processing of fuel elements and graphite blocks, the utilization of simplier units and mechanisms in the overloading system and in the process of profiling of energy distribution the choice of the spherical configuration for the VG-400 pilot plant installation seems to be valid

Gas-core reactor power transient analysis. Final report

International Nuclear Information System (INIS)

Kascak, A.F.

1972-01-01

The gas core reactor is a proposed device which features high temperatures. It has applications in high specific impulse space missions, and possibly in low thermal pollution MHD power plants. The nuclear fuel is a ball of uranium plasma radiating thermal photons as opposed to gamma rays. This thermal energy is picked up before it reaches the solid cavity liner by an inflowing seeded propellant stream and convected out through a rocket nozzle. A wall-burnout condition will exist if there is not enough flow of propellant to convect the energy back into the cavity. A reactor must therefore operate with a certain amount of excess propellant flow. Due to the thermal inertia of the flowing propellant, the reactor can undergo power transients in excess of the steady-state wall burnout power for short periods of time. The objective of the study was to determine how long the wall burnout power could be exceeded without burning out the cavity liner. The model used in the heat-transfer calculation was one-dimensional, and thermal radiation was assumed to be a diffusion process. (auth)

Results of an analysis of in-core measurements during the first core cycle of the Greifswald nuclear power plant, unit 3

International Nuclear Information System (INIS)

Gehre, G.

1982-01-01

First results of an analysis of flux and temperature values obtained from the in-core system in the third unit of the Greifswald nuclear power plant during the first core cycle are presented. The analysis has been performed with the aid of the computer code INCA. Possibilities and limits of this code are shown. (author)

Solid-Core, Gas-Cooled Reactor for Space and Surface Power

International Nuclear Information System (INIS)

King, Jeffrey C.; El-Genk, Mohamed S.

2006-01-01

The solid-core, gas-cooled, Submersion-Subcritical Safe Space (S and 4) reactor is developed for future space power applications and avoidance of single point failures. The Mo-14%Re reactor core is loaded with uranium nitride fuel in enclosed cavities, cooled by He-30%Xe, and sized to provide 550 kWth for seven years of equivalent full power operation. The beryllium oxide reflector disassembles upon impact on water or soil. In addition to decreasing the reactor and shadow shield mass, Spectral Shift Absorber (SSA) materials added to the reactor core ensure that it remains subcritical in the worst-case submersion accident. With a 0.1 mm thick boron carbide coating on the outside surface of the core block and 0.25 mm thick iridium sleeves around the fuel stacks, the reflector outer diameter is 43.5 cm and the combined reactor and shadow shield mass is 935.1 kg. With 12.5 atom% gadolinium-155 added to the fuel, 2.0 mm diameter gadolinium-155 sesquioxide intersititial pins, and a 0.1 mm thick gadolinium-155 sesquioxide coating, the S and 4 reactor has a slightly smaller reflector outer diameter of 43.0 cm, and a total reactor and shield mass of 901.7 kg. With 8.0 atom% europium-151 added to the fuel, 2.0 mm diameter europium-151 sesquioxide interstitial pins, and a 0.1 mm thick europium-151 sesquioxide coating, the reflector's outer diameter and the total reactor and shield mass are further reduced to 41.5 cm and 869.2 kg, respectively

Analysis of in-core coolant temperatures of FFTF instrumented fuels tests at full power

International Nuclear Information System (INIS)

Hoth, C.W.

1981-01-01

Two full size highly instrumented fuel assemblies were inserted into the core of the Fast Flux Test Facility in December of 1979. The major objectives of these instrumented tests are to provide verification of the FFTF core conditions and to characterize temperature patterns within FFTF driver fuel assemblies. A review is presented of the results obtained during the power ascents and during irradiation at a constant reactor power of 400 MWt. The results obtained from these instrumented tests verify the conservative nature of the design methods used to establish core conditions in FFTF. The success of these tests also demonstrates the ability to design, fabricate, install and irradiate complex, instrumented fuel tests in FFTF using commercially procured components

Gas core reactor power plants designed for low proliferation potential

International Nuclear Information System (INIS)

Lowry, L.L.

1977-09-01

The feasibility of gas core nuclear power plants to provide adequate power while maintaining a low inventory and low divertability of fissile material is studied. Four concepts were examined. Two used a mixture of UF 6 and helium in the reactor cavities, and two used a uranium-argon plasma, held away from the walls by vortex buffer confinement. Power levels varied from 200 to 2500 MWth. Power plant subsystems were sized to determine their fissile material inventories. All reactors ran, with a breeding ratio of unity, on 233 U born from thorium. Fission product removal was continuous. Newly born 233 U was removed continuously from the breeding blanket and returned to the reactor cavities. The 2500-MWth power plant contained a total of 191 kg of 233 U. Less than 4 kg could be diverted before the reactor shut down. The plasma reactor power plants had smaller inventories. In general, inventories were about a factor of 10 less than those in current U.S. power reactors

Computation of the Mutual Inductance between Air-Cored Coils of Wireless Power Transformer

International Nuclear Information System (INIS)

Anele, A O; Hamam, Y; Djouani, K; Chassagne, L; Alayli, Y; Linares, J

2015-01-01

Wireless power transfer system is a modern technology which allows the transfer of electric power between the air-cored coils of its transformer via high frequency magnetic fields. However, due to its coil separation distance and misalignment, maximum power transfer is not guaranteed. Based on a more efficient and general model available in the literature, rederived mathematical models for evaluating the mutual inductance between circular coils with and without lateral and angular misalignment are presented. Rather than presenting results numerically, the computed results are graphically implemented using MATLAB codes. The results are compared with the published ones and clarification regarding the errors made are presented. In conclusion, this study shows that power transfer efficiency of the system can be improved if a higher frequency alternating current is supplied to the primary coil, the reactive parts of the coils are compensated with capacitors and ferrite cores are added to the coils. (paper)

Analysis of the methodical component of core power density field calculation error on the basis of Mochovce-1 commissioning tests

International Nuclear Information System (INIS)

Brik, A.

2009-01-01

In the first decade of June 2008, during the power commissioning of the reactor at the Mochovce NPP unit 1, the experiment with reducing the thermal power of core almost to the balance-of-plant (BOP) needs was performed. After the reactor has operated for seven hours at low power (about 200 220 MW (thermal)), its power was increased (at a rate of about 0.25% of N nom /min) to the initial level, close to 107% (1471 MW). During the experiment, core parameters, which were subsequently used for comparing the measured data with the results of experiment simulation calculations, were recorded in the reactor in-core monitoring system database. Calculated and measured levels of critical concentrations of boric acid were compared, along with power density distributions by fuel elements and assemblies obtained both by the KRUIZ in-core monitoring system and on the basis of calculations simulating reactor operation in accordance with the given core power variation schedule. The final stage consisted of assessing the methodical component of power density micro- and macro-fields calculation error in the core of Mochovce-1 reactor operating with varying load. (author)

Analysis of the methodical component of core power density field calculation error on the basis of Mochovce-1 commissioning tests

International Nuclear Information System (INIS)

Brik, A.

2009-01-01

In the first decade of June 2008, during the power commissioning of the reactor at Mochovce NPP unit 1, the experiment with reducing the thermal power of core almost to the balance-of-plant needs was performed. After the reactor has operated for seven hours at low power (about 200 220 MW (thermal)), its power was increased (at a rate of about 0.25% of N nom /min) to the initial level, close to 107% (1471 MW). During the experiment, core parameters, which were subsequently used for comparing the measured data with the results of experiment simulation calculations, were recorded in the reactor in-core monitoring system's database. Calculated and measured levels of critical concentrations of boric acid were compared, along with power density distributions by fuel elements and assemblies obtained both by the KRUIZ in-core monitoring system and on the basis of calculations simulating reactor operation in accordance with the given core power variation schedule. The final stage consisted of assessing the methodical component of power density micro- and macro-fields' calculation error in the core of Mochovce-1 reactor operating with varying load. (Authors)

Design and Fabrication of Air-core Inductors for Power Conversion

DEFF Research Database (Denmark)

Lê Thanh, Hoà; Mizushima, Io; Tang, Peter Torben

supply on chip (PwrSoC) [1]. Examples of PwrSoC applications are power adaptors for LED illumination and the “Internet of Things”. We report an air-core MEMS inductor. Our process is scalable and universal for making inductors with versatile geometries e.g. spiral, solenoid, toroid, and advanced...

In-core Instrument Subcritical Verification (INCISV) - Core Design Verification Method - 358

International Nuclear Information System (INIS)

Prible, M.C.; Heibel, M.D.; Conner, S.L.; Sebastiani, P.J.; Kistler, D.P.

2010-01-01

According to the standard on reload startup physics testing, ANSI/ANS 19.6.1, a plant must verify that the constructed core behaves sufficiently close to the designed core to confirm that the various safety analyses bound the actual behavior of the plant. A large portion of this verification must occur before the reactor operates at power. The INCISV Core Design Verification Method uses the unique characteristics of a Westinghouse Electric Company fixed in-core self powered detector design to perform core design verification after a core reload before power operation. A Vanadium self powered detector that spans the length of the active fuel region is capable of confirming the required core characteristics prior to power ascension; reactivity balance, shutdown margin, temperature coefficient and power distribution. Using a detector element that spans the length of the active fuel region inside the core provides a signal of total integrated flux. Measuring the integrated flux distributions and changes at various rodded conditions and plant temperatures, and comparing them to predicted flux levels, validates all core necessary core design characteristics. INCISV eliminates the dependence on various corrections and assumptions between the ex-core detectors and the core for traditional physics testing programs. This program also eliminates the need for special rod maneuvers which are infrequently performed by plant operators during typical core design verification testing and allows for safer startup activities. (authors)

Improvement of the Cubic Spline Function Sets for a Synthesis of the Axial Power Distribution of a Core Protection System

International Nuclear Information System (INIS)

Koo, Bon-Seung; Lee, Chung-Chan; Zee, Sung-Quun

2006-01-01

Online digital core protection system(SCOPS) for a system-integrated modular reactor is being developed as a part of a plant protection system at KAERI. SCOPS calculates the minimum CHFR and maximum LPD based on several online measured system parameters including 3-level ex-core detector signals. In conventional ABB-CE digital power plants, cubic spline synthesis technique has been used in online calculations of the core axial power distributions using ex-core detector signals once every 1 second in CPC. In CPC, pre-determined cubic spline function sets are used depending on the characteristics of the ex-core detector responses. But this method shows an unnegligible power distribution error for the extremely skewed axial shapes by using restrictive function sets. Therefore, this paper describes the cubic spline method for the synthesis of an axial power distribution and it generates several new cubic spline function sets for the application of the core protection system, especially for the severely distorted power shapes needed reactor type

Nuclear piston engine and pulsed gaseous core reactor power systems

International Nuclear Information System (INIS)

Dugan, E.T.

1976-01-01

The investigated nuclear piston engines consist of a pulsed, gaseous core reactor enclosed by a moderating-reflecting cylinder and piston assembly and operate on a thermodynamic cycle similar to the internal combustion engine. The primary working fluid is a mixture of uranium hexafluoride, UF 6 , and helium, He, gases. Highly enriched UF 6 gas is the reactor fuel. The helium is added to enhance the thermodynamic and heat transfer characteristics of the primary working fluid and also to provide a neutron flux flattening effect in the cylindrical core. Two and four-stroke engines have been studied in which a neutron source is the counterpart of the sparkplug in the internal combustion engine. The piston motions which have been investigated include pure simple harmonic, simple harmonic with dwell periods, and simple harmonic in combination with non-simple harmonic motion. The results of the conducted investigations indicate good performance potential for the nuclear piston engine with overall efficiencies of as high as 50 percent for nuclear piston engine power generating units of from 10 to 50 Mw(e) capacity. Larger plants can be conceptually designed by increasing the number of pistons, with the mechanical complexity and physical size as the probable limiting factors. The primary uses for such power systems would be for small mobile and fixed ground-based power generation (especially for peaking units for electrical utilities) and also for nautical propulsion and ship power

Influence of FA pin power minimisation on neutron-physical core characteristics

International Nuclear Information System (INIS)

Mikolas, P.

2005-01-01

The influence of optimisation of radial fuel enrichment profilation in fuel assembly of Gd-2 type on lowering its pin power non-uniformity and subsequent lowering of F dH in WWER-440 cores are described in this paper (Author)

Physicians' Professionally Responsible Power: A Core Concept of Clinical Ethics.

Science.gov (United States)

McCullough, Laurence B

2016-02-01

The gathering of power unto themselves by physicians, a process supported by evidence-based practice, clinical guidelines, licensure, organizational culture, and other social factors, makes the ethics of power--the legitimation of physicians' power--a core concept of clinical ethics. In the absence of legitimation, the physician's power over patients becomes problematic, even predatory. As has occurred in previous issues of the Journal, the papers in the 2016 clinical ethics issue bear on the professionally responsible deployment of power by physicians. This introduction explores themes of physicians' power in papers from an international group of authors who address autonomy and trust, the virtues of perinatal hospice, conjoined twins in ethics and law, addiction and autonomy in clinical research on addicting substances, euthanasia of patients with dementia in Belgium, and a pragmatic approach to clinical futility. © The Author 2015. Published by Oxford University Press, on behalf of the Journal of Medicine and Philosophy Inc. All rights reserved. For permissions, please e-mail: journals.permissions@oup.com.

Automated in-core image generation from video to aid visual inspection of nuclear power plant cores

Energy Technology Data Exchange (ETDEWEB)

Murray, Paul, E-mail: paul.murray@strath.ac.uk [Department of Electronic and Electrical Engineering, University of Strathclyde, Technology and Innovation Centre, 99 George Street, Glasgow, G1 1RD (United Kingdom); West, Graeme; Marshall, Stephen; McArthur, Stephen [Dept. Electronic and Electrical Engineering, University of Strathclyde, Royal College Building, 204 George Street, Glasgow G1 1XW (United Kingdom)

2016-04-15

Highlights: • A method is presented which improves visual inspection of reactor cores. • Significant time savings are made to activities on the critical outage path. • New information is extracted from existing data sources without additional overhead. • Examples from industrial case studies across the UK fleet of AGR stations. - Abstract: Inspection and monitoring of key components of nuclear power plant reactors is an essential activity for understanding the current health of the power plant and ensuring that they continue to remain safe to operate. As the power plants age, and the components degrade from their initial start-of-life conditions, the requirement for more and more detailed inspection and monitoring information increases. Deployment of new monitoring and inspection equipment on existing operational plant is complex and expensive, as the effect of introducing new sensing and imaging equipment to the existing operational functions needs to be fully understood. Where existing sources of data can be leveraged, the need for new equipment development and installation can be offset by the development of advanced data processing techniques. This paper introduces a novel technique for creating full 360° panoramic images of the inside surface of fuel channels from in-core inspection footage. Through the development of this technique, a number of technical challenges associated with the constraints of using existing equipment have been addressed. These include: the inability to calibrate the camera specifically for image stitching; dealing with additional data not relevant to the panorama construction; dealing with noisy images; and generalising the approach to work with two different capture devices deployed at seven different Advanced Gas Cooled Reactor nuclear power plants. The resulting data processing system is currently under formal assessment with a view to replacing the existing manual assembly of in-core defect montages. Deployment of the

Power Production Analysis of the OE Buoy WEC for the CORES Project

DEFF Research Database (Denmark)

Lavelle, John; Kofoed, Jens Peter

This report describes the analysis performed on the OE Buoy for the CORES project by the wave energy group at Aalborg University, Denmark. OE Buoy is a type of Oscillating Water Column (OWC) wave energy converter as part of the CORES project. This type of device is one of the most developed...... to extract energy from the ocean (1). Typically, a Wells turbine is used for the Power Take Off (PTO) for OWCs. The Wells turbine has the advantage that it is self-rectifying – with the ability to operate with either direction of airflow, which changes during each cycle of the wave. This type of turbine...... which a total of 39 hours of power production data was collected. A data acquisition system was used to sample the sensors on board and the generator shaft power time-series data was used in the analysis here. A wave-rider buoy, located at the site of OE Buoy and operated by the Marine Institute Ireland...

Start-up fuel and power flattening of sodium-cooled candle core

International Nuclear Information System (INIS)

Takaki, Naoyuki; Sagawa, Yu; Umino, Akitake; Sekimoto, Hiroshi

2013-01-01

The hard neutron spectrum and unique power shape of CANDLE enable its distinctive performances such as achieving high burnup more than 30% and exempting necessity of both enrichment and reprocessing. On the other hand, they also cause several challenging problems. One is how the initial fuel can be prepared to start up the first CANDLE reactor because the equilibrium fuel composition that enables stable CANDLE burning is complex both in axial and radial directions. Another prominent problem is high radial power peaking factor that worsens averaged burnup, namely resource utilization factor in once-through mode and shorten the life time of structure materials. The purposes of this study are to solve these two problems. Several ideas for core configurations and startup fuel using single enrichment uranium and iron as a substitute of fission products are studied. As a result, it is found that low enriched uranium is applicable to ignite the core but all concepts examined here exceeded heat limits. Adjustment in enrichment and height of active and burnt zone is opened for future work. Sodium duct assemblies and thorium fuel assemblies loaded in the center region are studied as measures to reduce radial power peaking factor. Replacing 37 fuels by thorium fuel assemblies in the zeroth to third row provides well-balanced performance with flattened radial power distribution. The CANDLE core loaded with natural uranium in the outer and thorium in the center region achieved 35.6% of averaged burnup and 7.0 years of cladding life time owing to mitigated local fast neutron irradiation at the center. Using thorium with natural or depleted uranium in CANDLE reactor is also beneficial to diversifying fission resource and extending available term of fission energy without expansion of needs for enrichment and reprocessing

Power-level regulation and simulation of nonlinear pressurized water reactor core with xenon oscillation using H-infinity loop shaping control

Directory of Open Access Journals (Sweden)

Li Gang

2016-01-01

Full Text Available This investigation is to solve the power-level control issue of a nonlinear pressurized water reactor core with xenon oscillations. A nonlinear pressurized water reactor core is modeled using the lumped parameter method, and a linear model of the core is then obtained through the small perturbation linearization way. The H∞loop shapingcontrolis utilized to design a robust controller of the linearized core model.The calculated H∞loop shaping controller is applied to the nonlinear core model. The nonlinear core model and the H∞ loop shaping controller build the nonlinear core power-level H∞loop shaping control system.Finally, the nonlinear core power-level H∞loop shaping control system is simulatedconsidering two typical load processes that are a step load maneuver and a ramp load maneuver, and simulation results show that the nonlinear control system is effective.

Development of Core Monitoring System for Nuclear Power Plants (I)

Energy Technology Data Exchange (ETDEWEB)

Lee, S.H.; Kim, Y.B.; Park, M.G; Lee, E.K.; Shin, H.C.; Lee, D.J. [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

1997-12-31

1.Object and Necessity of the Study -The main objectives of this study are (1)conversion of APOLLO version BEACON system to HP-UX version core monitoring system, (2)provision of the technical bases to enhance the in-house capability of developing more advanced core monitoring system. 2.Results of the Study - In this study, the revolutionary core monitoring technologies such as; nodal analysis and isotope depletion calculation method, advanced schemes for power distribution control, and treatment of nuclear databank were established. The verification and validation work has been successfully performed by comparing the results with those of the design code and measurement data. The advanced graphic user interface and plant interface method have been implemented to ensure the future upgrade capability. The Unix shell scripts and system dependent software are also improved to support administrative functions of the system. (author). 14 refs., 112 figs., 52 tabs.

Self-powered in-core neutron detector assembly with uniform perturbation characteristics

International Nuclear Information System (INIS)

1981-01-01

An in-core neutron detector assembly consisting of a number of longitudinally extending self-powered detectors is described. The uniform mechanical structures and materials are placed symmetrically at each active detector portion thus ensuring that local perturbation factors are uniform. (U.K.)

Aerosol core nuclear reactor for space-based high energy/power nuclear-pumped lasers

International Nuclear Information System (INIS)

Prelas, M.A.; Boody, F.P.; Zediker, M.S.

1987-01-01

An aerosol core reactor concept can overcome the efficiency and/or chemical activity problems of other fuel-reactant interface concepts. In the design of a laser using the nuclear energy for a photon-intermediate pumping scheme, several features of the aerosol core reactor concept are attractive. First, the photon-intermediate pumping concept coupled with photon concentration methods and the aerosol fuel can provide the high power densities required to drive high energy/power lasers efficiently (about 25 to 100 kW/cu cm). Secondly, the intermediate photons should have relatively large mean free paths in the aerosol fuel which will allow the concept to scale more favorably. Finally, the aerosol core reactor concept can use materials which should allow the system to operate at high temperatures. An excimer laser pumped by the photons created in the fluorescer driven by a self-critical aerosol core reactor would have reasonable dimensions (finite cylinder of height 245 cm and radius of 245 cm), reasonable laser energy (1 MJ in approximately a 1 millisecond pulse), and reasonable mass (21 kg uranium, 8280 kg moderator, 460 kg fluorescer, 450 kg laser medium, and 3233 kg reflector). 12 references

Physicians’ Professionally Responsible Power: A Core Concept of Clinical Ethics

Science.gov (United States)

McCullough, Laurence B.

2016-01-01

The gathering of power unto themselves by physicians, a process supported by evidence-based practice, clinical guidelines, licensure, organizational culture, and other social factors, makes the ethics of power—the legitimation of physicians’ power—a core concept of clinical ethics. In the absence of legitimation, the physician’s power over patients becomes problematic, even predatory. As has occurred in previous issues of the Journal, the papers in the 2016 clinical ethics issue bear on the professionally responsible deployment of power by physicians. This introduction explores themes of physicians’ power in papers from an international group of authors who address autonomy and trust, the virtues of perinatal hospice, conjoined twins in ethics and law, addiction and autonomy in clinical research on addicting substances, euthanasia of patients with dementia in Belgium, and a pragmatic approach to clinical futility. PMID:26671961

On-line extraction of the variance caused by burn-up in in-core three-dimensional power distribution

International Nuclear Information System (INIS)

Wang Yaqi; Luo Zhengpei; Li Fu; Liu Wenfeng

2001-01-01

In most of PWRs, the ex-core ion-chambers are the sole real-time sensors to respond to in-core power and its axial offset. However, the calibration coefficient of the ion-chambers depends on the (3D) power distribution and varies with the burn-up. People expect to know the variance in distribution caused by burn-up directly from the signals of ion-chambers. This expectation is not realized as yet, because an ion-chamber almost only responds to its nearest fuel assemblies. The authors then developed a two-step method for burn-up characteristic extraction: the harmonics synthesis method and harmonics' burn-up grouping. Using the extracted burn-up characteristics, the relationship between the readings of the ex-core ion-chambers and the in-core 3D power distribution is set up. Through the simulation on the heating reactor, the method of burn-up characteristic extraction is verified under engineering conditions. It is possible to on-line extract the variance caused by burn-up in 3D power distribution

Self-powered in-core neutron detector assembly with uniform perturbation characteristics

International Nuclear Information System (INIS)

Todt, W.H.; Playfoot, K.C.

1979-01-01

Disclosed is a self-powered in-core neutron detector assembly in which a plurality of longitudinally extending self-powered detectors have neutron responsive active portions spaced along a longitudinal path. A low neutron absorptive extension extends from the active portions of the spaced active portions of the detectors in symmetrical longitudinal relationship with the spaced active detector portions of each succeeding detector. The detector extension terminates with the detector assembly to provide a uniform perturbation characteristic over the entire assembly length

Neutronics simulations on hypothetical power excursion and possible core melt scenarios in CANDU6

International Nuclear Information System (INIS)

Kim, Yonghee

2015-01-01

LOCA (Loss of coolant accident) is an outstanding safety issue in the CANDU reactor system since the coolant void reactivity is strongly positive. To deal with the LOCA, the CANDU systems are equipped with specially designed quickly-acting secondary shutdown system. Nevertheless, the so-called design-extended conditions are requested to be taken into account in the safety analysis for nuclear reactor systems after the Fukushima accident. As a DEC scenario, the worst accident situation in a CANDU reactor system is a unprotected LOCA, which is supposed to lead to a power excursion and possibly a core melt-down. In this work, the hypothetical unprotected LOCA scenario is simulated in view of the power excursion and fuel temperature changes by using a simplified point-kinetics (PK) model accounting for the fuel temperature change. In the PK model, the core reactivity is assumed to be affected by a large break LOCA and the fuel temperature is simulated to account for the Doppler effect. In addition, unlike the conventional PK simulation, we have also considered the Xe-I model to evaluate the impact of Xe during the LOCA. Also, we tried to simulate the fuel and core melt-down scenario in terms of the reactivity through a series of neutronics calculations for hypothetical core conditions. In case of a power excursion and possible fuel melt-down situation, the reactor system behavior is very uncertain. In this work, we tried to understand the impacts of fuel melt and relocation within the pressure vessel on the core reactivity and failure of pressure and calandria tubes. (author)

Development of Optimized Core Design and Analysis Methods for High Power Density BWRs

Science.gov (United States)

Shirvan, Koroush

Increasing the economic competitiveness of nuclear energy is vital to its future. Improving the economics of BWRs is the main goal of this work, focusing on designing cores with higher power density, to reduce the BWR capital cost. Generally, the core power density in BWRs is limited by the thermal Critical Power of its assemblies, below which heat removal can be accomplished with low fuel and cladding temperatures. The present study investigates both increases in the heat transfer area between ~he fuel and coolant and changes in operating parameters to achieve higher power levels while meeting the appropriate thermal as well as materials and neutronic constraints. A scoping study is conducted under the constraints of using fuel with cylindrical geometry, traditional materials and enrichments below 5% to enhance its licensability. The reactor vessel diameter is limited to the largest proposed thus far. The BWR with High power Density (BWR-HD) is found to have a power level of 5000 MWth, equivalent to 26% uprated ABWR, resulting into 20% cheaper O&M and Capital costs. This is achieved by utilizing the same number of assemblies, but with wider 16x16 assemblies and 50% shorter active fuel than that of the ABWR. The fuel rod diameter and pitch are reduced to just over 45% of the ABWR values. Traditional cruciform form control rods are used, which restricts the assembly span to less than 1.2 times the current GE14 design due to limitation on shutdown margin. Thus, it is possible to increase the power density and specific power by 65%, while maintaining the nominal ABWR Minimum Critical Power Ratio (MCPR) margin. The plant systems outside the vessel are assumed to be the same as the ABWR-Il design, utilizing a combination of active and passive safety systems. Safety analyses applied a void reactivity coefficient calculated by SIMULA TE-3 for an equilibrium cycle core that showed a 15% less negative coefficient for the BWR-HD compared to the ABWR. The feedwater

Core-power and decay-time limits for disabled automatic-actuation of LOFT ECCS

International Nuclear Information System (INIS)

Hanson, G.H.

1978-01-01

The Emergency Core Cooling System (ECCS) for the LOFT reactor may need to be disabled for modifications or repairs of hardware or instrumentation or for component testing during periods when the reactor system is hot and pressurized, or it may be desirable to enable the ECCS to be disabled without the necessity of cooling down and depressurizing the reactor. LTR 113-47 has shown that the LOFT ECCS can be safely bypassed or disabled when the total core power does not exceed 25 kW. A modified policy involves disabling the automatic actuation of the LOFT ECCS, but still retaining the manual activation capability. Disabling of the automatic actuation can be safely utilized, without subjecting the fuel cladding to unacceptable temperatures, when the LOFT power decays to 70 kW; this power level permits a maximum delay of 20 minutes following a LOCA for the manual actuation of ECCS

The change of radial power factor distribution due to RCCA insertion at the first cycle core of AP1000

Science.gov (United States)

Susilo, J.; Suparlina, L.; Deswandri; Sunaryo, G. R.

2018-02-01

The using of a computer program for the PWR type core neutronic design parameters analysis has been carried out in some previous studies. These studies included a computer code validation on the neutronic parameters data values resulted from measurements and benchmarking calculation. In this study, the AP1000 first cycle core radial power peaking factor validation and analysis were performed using CITATION module of the SRAC2006 computer code. The computer code has been also validated with a good result to the criticality values of VERA benchmark core. The AP1000 core power distribution calculation has been done in two-dimensional X-Y geometry through ¼ section modeling. The purpose of this research is to determine the accuracy of the SRAC2006 code, and also the safety performance of the AP1000 core first cycle operating. The core calculations were carried out with the several conditions, those are without Rod Cluster Control Assembly (RCCA), by insertion of a single RCCA (AO, M1, M2, MA, MB, MC, MD) and multiple insertion RCCA (MA + MB, MA + MB + MC, MA + MB + MC + MD, and MA + MB + MC + MD + M1). The maximum power factor of the fuel rods value in the fuel assembly assumedapproximately 1.406. The calculation results analysis showed that the 2-dimensional CITATION module of SRAC2006 code is accurate in AP1000 power distribution calculation without RCCA and with MA+MB RCCA insertion.The power peaking factor on the first operating cycle of the AP1000 core without RCCA, as well as with single and multiple RCCA are still below in the safety limit values (less then about 1.798). So in terms of thermal power generated by the fuel assembly, then it can be considered that the AP100 core at the first operating cycle is safe.

Research on 3D power distribution of PWR reactor core based on RBF neural network

International Nuclear Information System (INIS)

Xia Hong; Li Bin; Liu Jianxin

2014-01-01

Real-time monitor for 3D power distribution is critical to nuclear safety and high efficiency of NPP's operation as well as the control system optimization. A method was proposed to set up a real-time monitor system for 3D power distribution by using of ex-core neutron detecting system and RBF neural network for improving the instantaneity of the monitoring results and reducing the fitting error of the 3D power distribution. A series of experiments were operated on a 300 MW PWR simulation system. The results demonstrate that the new monitor system works very well under condition of certain burnup range during the fuel cycle and reconstructs the real-time 3D distribution of reactor core power. The accuracy of the model is improved effectively with the help of several methods. (authors)

Optimization design of toroidal core for magnetic energy harvesting near power line by considering saturation effect

Science.gov (United States)

Park, Bumjin; Kim, Dongwook; Park, Jaehyoung; Kim, Kibeom; Koo, Jay; Park, HyunHo; Ahn, Seungyoung

2018-05-01

Recently, magnetic energy harvesting technologies have been studied actively for self-sustainable operation of applications around power line. However, magnetic energy harvesting around power lines has the problem of magnetic saturation, which can cause power performance degradation of the harvester. In this paper, optimal design of a toroidal core for magnetic energy harvesters has been proposed with consideration of magnetic saturation near power lines. Using Permeability-H curve and Ampere's circuital law, the optimum dimensional parameters needed to generate induced voltage were analyzed via calculation and simulation. To reflect a real environment, we consider the nonlinear characteristic of the magnetic core material and supply current through a 3-phase distribution panel used in the industry. The effectiveness of the proposed design methodology is verified by experiments in a power distribution panel and takes 60.9 V from power line current of 60 A at 60 Hz.

The ARIES-RS power core - recent development in Li/V designs

International Nuclear Information System (INIS)

Sze Dai-Kai; Billone, M.C.; Hua, T.Q.; Tillack, M.; Najmabadi, F.; Wang Xueren; Malang, S.; El-Guebaly, L.A.; Sviatoslavsky, I.N.; Blanchard, J.P.; Crowell, J.A.; Khater, H.Y.; Mogahed, E.A.; Waganer, L.M.; Lee, D.; Cole, D.

1998-01-01

The ARIES-RS fusion power plant design study is based on reversed-shear (RS) physics with a Li/V (lithium breeder and vanadium structure) blanket. The reversed-shear discharge has been documented in many large tokamak experiments. The plasma in the RS mode has a high beta, low current, and low current drive requirement. Therefore, it is an attractive physics regime for a fusion power plant. The blanket system based on Li/V has high temperature operating capability, good tritium breeding, excellent high heat flux removal capability, long structural life time, low activation, low after heat and good safety characteristics. For these reasons, the ARIES-RS reactor study selected Li/V as the reference blanket. The combination of attractive physics and attractive blanket engineering is expected to result in a superior power plant design. This paper summarizes the power core design of the ARIES-RS power plant study. (orig.)

High Power Spark Delivery System Using Hollow Core Kagome Lattice Fibers

Directory of Open Access Journals (Sweden)

Ciprian Dumitrache

2014-08-01

Full Text Available This study examines the use of the recently developed hollow core kagome lattice fibers for delivery of high power laser pulses. Compared to other photonic crystal fibers (PCFs, the hollow core kagome fibers have larger core diameter (~50 µm, which allows for higher energy coupling in the fiber while also maintaining high beam quality at the output (M2 = 1.25. We have conducted a study of the maximum deliverable energy versus laser pulse duration using a Nd:YAG laser at 1064 nm. Pulse energies as high as 30 mJ were transmitted for 30 ns pulse durations. This represents, to our knowledge; the highest laser pulse energy delivered using PCFs. Two fiber damage mechanisms were identified as damage at the fiber input and damage within the bulk of the fiber. Finally, we have demonstrated fiber delivered laser ignition on a single-cylinder gasoline direct injection engine.

Determination of the decay power for a U3O8 designed core using the ORIGEN 2.1 code

International Nuclear Information System (INIS)

Castro, Jose; Gallardo, Alberto; Madariaga, Marcelo

2014-01-01

After the operation of a nuclear research reactor at a higher power (more than 300 kW), a cooling time is required to remove the residual heat from the core due to the heat produced by the energy emitted by fission products, this fact is common in reactors. There is a short time where the heat output falls to 6 % after the reactor shutdown, the importance of knowing this power is because of the accidental events that this power could cause and affect the fuel after a sudden shutdown in the cooling system of the reactor and there is any other refrigeration system, only that one surrounding the reactor core. This report shows the results of the calculation of the U 3 O 8 core residual power a for the RP-10, using the ORIGEN 2.1 calculation code, verifying the safety of the proposed core within the safety limits accepted for the reactor. (authors).

Fuel and Core Design Verification for Extended Power Up-rate in Ringhals Unit 3

International Nuclear Information System (INIS)

Gabrielsson, Petter; Stepniewski, Marek; Almberger, Jan

2006-01-01

Vattenfall's Westinghouse 3-loop PWR Ringhals 3 at the western coast of Sweden is scheduled for an extended power up-rate from 2783 to 3160 MWt in 2007, in the frame of the so called GREAT-project. The project will realize an up-rating initially planned and analysed back in 1995, but with a number of significant improvements outlined in this paper. For the licensing of the up-rated power level, a complete revision of the safety analyses, radiological analyses and systems verifications in FSAR is being performed by Westinghouse Electrics Belgium. The work is performed in close cooperation with Vattenfall in the areas of core calculations and input data. For more than a decade, Vattenfall has performed all core design and reload safety evaluations (RSE) for Ringhals, independent of fuel vendors and safety analysts. In GREAT all core parameters in the safety analysis checklist (SAC) used for the safety analyses are determined based upon a set of nine reference loading patterns designed by Vattenfall covering a wide range of fuel and core designs and extreme cycle-to-cycle variations. To facilitate the calculation of SAC parameters Westinghouse has provided a Reload Safety Evaluation Procedure report (RSEP) with detailed specifications for the calculation of all core parameters used in the analyses. The procedure has been automatized by Vattenfall in a set of scripts executing 3D core simulator calculations and extracting the key results. The same tools will be used in Vattenfall's future RSE for Ringhals 3. This approach is taken to obtain consistency between core designs and core calculations for the safety analyses and the cycle specific calculations, to minimize the risk for future violations of the safety analyses. (authors)

Core fusion accidents in nuclear power reactors. Knowledge review

International Nuclear Information System (INIS)

Bentaib, Ahmed; Bonneville, Herve; Clement, Bernard; Cranga, Michel; Fichot, Florian; Koundy, Vincent; Meignen, Renaud; Corenwinder, Francois; Leteinturier, Denis; Monroig, Frederique; Nahas, Georges; Pichereau, Frederique; Van-Dorsselaere, Jean-Pierre; Cenerino, Gerard; Jacquemain, Didier; Raimond, Emmanuel; Ducros, Gerard; Journeau, Christophe; Magallon, Daniel; Seiler, Jean-Marie; Tourniaire, Bruno

2013-01-01

This reference document proposes a large and detailed review of severe core fusion accidents occurring in nuclear power reactors. It aims at presenting the scientific aspects of these accidents, a review of knowledge and research perspectives on this issue. After having recalled design and operation principles and safety principles for reactors operating in France, and the main studied and envisaged accident scenarios for the management of severe accidents in French PWRs, the authors describe the physical phenomena occurring during a core fusion accident, in the reactor vessel and in the containment building, their sequence and means to mitigate their effects: development of the accident within the reactor vessel, phenomena able to result in an early failure of the containment building, phenomena able to result in a delayed failure with the corium-concrete interaction, corium retention and cooling in and out of the vessel, release of fission products. They address the behaviour of containment buildings during such an accident (sizing situations, mechanical behaviour, bypasses). They review and discuss lessons learned from accidents (Three Mile Island and Chernobyl) and simulation tests (Phebus-PF). A last chapter gives an overview of software and approaches for the numerical simulation of a core fusion accident

On-power verification of the dynamic response of self-powered in-core detectors

International Nuclear Information System (INIS)

Serdula, K.; Beaudet, M.

1996-01-01

Self-powered in-core detectors are used for on-line safety and regulation purposes in CANDU reactors. Such applications require use of detectors whose response is primarily prompt to changes in flux. In-service verification of the detectors' response is required to ensure significant degradation in performance has not occurred during long-term operation. Changes in the detector characteristics occur due to nuclear interactions and failures. Present verification requires significant station resources and disrupts power production. Use of the 'noise' in the detector signal is being investigated as an alternative to assess the dynamic response of the detectors during long-term operation. Measurements of reference 'signatures' were obtained from replacement shutdown system detectors. Results show 'noise' measurements are a promising alternative to the current verification method. Identification of changes in the detector response function assist in accurate diagnosis and prognosis of changes in detector signals due to process changes. (author)

Fabrication of 3D Air-core MEMS Inductors for High Frequency Power Electronic Applications

DEFF Research Database (Denmark)

Lê Thanh, Hoà; Mizushima, Io; Nour, Yasser

2018-01-01

footprints have an inductance from 34.2 to 44.6 nH and a quality factor from 10 to 13 at frequencies ranging from 30 to 72 MHz. The air-core inductors show threefold lower parasitic capacitance and up to a 140% higher-quality factor and a 230% higher-operation frequency than silicon-core inductors. A 33 MHz...... boost converter mounted with an air-core toroidal inductor achieves an efficiency of 68.2%, which is better than converters mounted with a Si-core inductor (64.1%). Our inductors show good thermal cycling stability, and they are mechanically stable after vibration and 2-m-drop tests.......We report a fabrication technology for 3D air-core inductors for small footprint and very-high-frequency power conversions. Our process is scalable and highly generic for fabricating inductors with a wide range of geometries and core shapes. We demonstrate spiral, solenoid, and toroidal inductors...

Implementation of a fast running full core pin power reconstruction method in DYN3D

International Nuclear Information System (INIS)

Gomez-Torres, Armando Miguel; Sanchez-Espinoza, Victor Hugo; Kliem, Sören; Gommlich, Andre

2014-01-01

Highlights: • New pin power reconstruction (PPR) method for the nodal diffusion code DYN3D. • Flexible PPR method applicable to a single, a group or to all fuel assemblies (square, hex). • Combination of nodal with pin-wise solutions (non-conform geometry). • PPR capabilities shown for REA of a Minicore (REA) PWR whole core. - Abstract: This paper presents a substantial extension of the pin power reconstruction (PPR) method used in the reactor dynamics code DYN3D with the aim to better describe the heterogeneity within the fuel assembly during reactor simulations. The flexibility of the new implemented PPR permits the local spatial refinement of one fuel assembly, of a cluster of fuel assemblies, of a quarter or eight of a core or even of a whole core. The application of PPR in core regions of interest will pave the way for the coupling with sub-channel codes enabling the prediction of local safety parameters. One of the main advantages of considering regions and not only a hot fuel assembly (FA) is the fact that the cross flow within this region can be taken into account by the subchannel code. The implementation of the new PPR method has been tested analysing a rod ejection accident (REA) in a PWR minicore consisting of 3 × 3 FA. Finally, the new capabilities of DNY3D are demonstrated by the analysing a boron dilution transient in a PWR MOX core and the pin power of a VVER-1000 reactor at stationary conditions

Implementation of a fast running full core pin power reconstruction method in DYN3D

Energy Technology Data Exchange (ETDEWEB)

Gomez-Torres, Armando Miguel [Instituto Nacional de Investigaciones Nucleares, Department of Nuclear Systems, Carretera Mexico – Toluca s/n, La Marquesa, 52750 Ocoyoacac (Mexico); Sanchez-Espinoza, Victor Hugo, E-mail: victor.sanchez@kit.edu [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology, Hermann-vom-Helmhotz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Kliem, Sören; Gommlich, Andre [Helmholtz-Zentrum Dresden-Rossendorf, Bautzner Landstraße 400, 01328 Dresden (Germany)

2014-07-01

Highlights: • New pin power reconstruction (PPR) method for the nodal diffusion code DYN3D. • Flexible PPR method applicable to a single, a group or to all fuel assemblies (square, hex). • Combination of nodal with pin-wise solutions (non-conform geometry). • PPR capabilities shown for REA of a Minicore (REA) PWR whole core. - Abstract: This paper presents a substantial extension of the pin power reconstruction (PPR) method used in the reactor dynamics code DYN3D with the aim to better describe the heterogeneity within the fuel assembly during reactor simulations. The flexibility of the new implemented PPR permits the local spatial refinement of one fuel assembly, of a cluster of fuel assemblies, of a quarter or eight of a core or even of a whole core. The application of PPR in core regions of interest will pave the way for the coupling with sub-channel codes enabling the prediction of local safety parameters. One of the main advantages of considering regions and not only a hot fuel assembly (FA) is the fact that the cross flow within this region can be taken into account by the subchannel code. The implementation of the new PPR method has been tested analysing a rod ejection accident (REA) in a PWR minicore consisting of 3 × 3 FA. Finally, the new capabilities of DNY3D are demonstrated by the analysing a boron dilution transient in a PWR MOX core and the pin power of a VVER-1000 reactor at stationary conditions.

Stability region for a prompt power variation of a coupled-core system with positive prompt feedback

International Nuclear Information System (INIS)

Watanabe, S.; Nishina, K.

1984-01-01

A stability analysis using a one-group model is presented for a coupled-core system. Positive prompt feedback of a γp /SUB j/ form is assumed, where p /SUB j/ is the fractional power variation of core j. Prompt power variations over a range of a few milliseconds after a disturbance are analyzed. The analysis combines Lapunov's method, prompt jump approximation, and the eigenfunction expansion of coupling region response flux. The last is treated as a pseudo-delayed neutron precursor. An asymptotic stability region is found for p /SUB j/. For an asymmetric flux variation over a system of two coupled cores, either p /SUB I/ or p /SUB II/ can slightly exceed, by virtue of the coupling effect, the critical value (β/γ-1) of a single-core case. Such a stability region is increased by additional inclusion of the coupling region fundamental mode in the treatment. The coupling region contributes to stability through its delayed response and coupling. An optimum core separation distance for stability is found

Nuclear analysis and performance of the Light Water Breeder Reactor (LWBR) core power operation at Shippingport

International Nuclear Information System (INIS)

Hecker, H.C.

1984-04-01

This report presents the nuclear analysis and discusses the performance of the LWBR core at Shippingport during power operation from initial startup through end-of-life at 28,730 EFPH. Core follow depletion calculations confirmed that the reactivity bias and power distributions were well within the uncertainty allowances used in the design and safety analysis of LWBR. The magnitude of the core follow reactivity bias has shown that the calculational models used can predict the behavior of U 233 -Th systems with closely spaced fuel rod lattices and movable fuel. In addition, the calculated final fissile loading is sufficiently greater than the initial fissile inventory that the measurements to be performed for proof-of-breeding evaluations are expected to confirm that breeding has occurred

Determining the magnetically nonlinear characteristics of a three phase core-type power transformer

International Nuclear Information System (INIS)

Dolinar, Matjaz; Stumberger, Gorazd; Polajzer, Bostjan; Dolinar, Drago

2006-01-01

This paper presents nonlinear iron core model of a three-phase, three-limb power transformer which is given by the current-dependant characteristics of flux linkages. The magnetically nonlinear characteristics are determined by controlled magnetic excitation of all three limbs which allows to take into account the variable magnetic-cross couplings between different coils placed on limbs, caused by saturation. The corresponding partial derivatives of measured flux linkage characteristics are used in the transformer circuit model as a magnetically nonlinear iron core model in order to analyze the behaviour of a nonsymmetrically excited transformer. Numerical results using transformer model with the determined iron core model agree very well with the measured results

Use of in-core self-powered detectors in the nuclear reactor control system

International Nuclear Information System (INIS)

Mitel'man, M.G.; Andreev, L.G.; Batenin, I.V.

1975-01-01

The paper describes the results of experiments conducted at the Obninsk atomic power station on the establishment of an integrated system for the control and automatic regulation of the distribution of energy liberation. The pick-ups used for this system are direct-charge detectors with rhodium emitters. The system, which consists of 12 integral DPZ-7 detectors distributed evenly on 2 mutually perpendicular diamaters and 2 assemblies, each containing 4 1n direct-charge detectors, monitors the distribution of neutron flow density according to core height. Increased signals from the detectors were fed into the summation block, in which the outgoing signal was also adjusted. The half-life of the rhodium was calculated and the non-inertia of the entire system ascertained. The total operating time of the system for the automatic control of reactor power is approximately 1 h. The influence of interference was not determined. During the experiment, the power of the reactor remained constant, i.e. its regulation by the direct-charge detectors inside the core was effective. (author)

Mini-cavity plasma core reactors for dual-mode space nuclear power/propulsion systems

International Nuclear Information System (INIS)

Chow, S.

1976-01-01

A mini-cavity plasma core reactor is investigated for potential use in a dual-mode space power and propulsion system. In the propulsive mode, hydrogen propellant is injected radially inward through the reactor solid regions and into the cavity. The propellant is heated by both solid driver fuel elements surrounding the cavity and uranium plasma before it is exhausted out the nozzle. The propellant only removes a fraction of the driver power, the remainder is transferred by a coolant fluid to a power conversion system, which incorporates a radiator for heat rejection. In the power generation mode, the plasma and propellant flows are shut off, and the driver elements supply thermal power to the power conversion system, which generates electricity for primary electric propulsion purposes

Core-power and decay-time limits for disabled automatic-actuation of LOFT ECCS

International Nuclear Information System (INIS)

Hanson, G.H.

1978-01-01

The Emergency Core Cooling System (ECCS) for the LOFT reactor may need to be disabled for modifications or repairs of hardware or instrumentation or for component testing during periods when the reactor system is hot and pressurized, or it may be desirable to enable the ECCS to be disabled without the necessity of cooling down and depressurizing the reactor. A policy involves disabling the automatic-actuation of the LOFT ECCS, but still retaining the manual actuation capability. Disabling of the automatic actuation can be safely utilized, without subjecting the fuel cladding to unacceptable temperatures, when the LOFT power decays to 33 kW; this power level permits a maximum delay of 20 minutes following a LOCA for the manual actuation of ECCS. For the operating power of the L2-2 Experiment, the required decay-periods (with operating periods of 40 and 2000 hours) are about 21 and 389 hours, respectively. With operating periods of 40 and 2000 hours at Core-I full power, the required decay-periods are about 42 and 973 hours, respectively. After these decay periods the automatic actuation of the LOFT ECCS can be disabled assuming a maximum delay of 20 minutes following a LOCA for the manual actuation of ECCS. The automatic and manual lineup of the ECCS may be waived if decay power is less than 11 kW

Reactor core cooling device for nuclear power plant

International Nuclear Information System (INIS)

Tsuda, Masahiko.

1992-01-01

The present invention concerns a reactor core cooling facility upon rupture of pipelines in a BWR type nuclear power plant. That is, when rupture of pipelines should occur in the reactor container, an releasing safety valve operates instantly and then a depressurization valve operates to depressurize the inside of a reactor pressure vessel. Further, an injection valve of cooling water injection pipelines is opened and cooling water is injected to cool the reactor core from the time when the pressure is lowered to a level capable of injecting water to the pressure vessel by the static water head of a pool water as a water source. Further, steams released from the pressure vessel and steams in the pressure vessel are condensed in a high pressure/low pressure emergency condensation device and the inside of the reactor container is depressurized and cooled. When the reactor is isolated, since the steams in the pressure vessel are condensed in the state that the steam supply valve and the return valve of a steam supply pipelines are opened and a vent valve is closed, the reactor can be maintained safely. (I.S.)

Feasibility of maintaining natural convection mode core cooling in research reactor power upgrades

International Nuclear Information System (INIS)

Ha, J.J.; Belhadj, M.; Aldemir, T.; Christensen, R.N.

1987-01-01

Two operational concerns for natural convection coooled research reactors using plate type fuels are: 1) pool top 16 N activity (PTNA), and 2) nucleate boiling in core channels. The feasibility assessment of a power upgrade while maintaining natural convection mode core cooling requires addressing these operational concerns. Previous studies have shown that: a) The conventional technique for reducing PTNA by plume dispersion may not be effective in a large power upgrade of research reactors with small pools. b) Currently used correlations to predict onset of nucleate boiling (ONB) in thin, rectangular core channels are not valid for low-velocity, upward flows such as encountered in natural convection cooling. The PTNA depends on the velocity distribution in the reactor pool. COMMIX-1A code is used to determine the three-dimensional velocity fields in The Ohio State University Research Reactor (OSURR) pool as a function of varying design conditions, following a power upgrade to 500 kW with LEU fuel. It is shown that a sufficiently deep stagnant water layer can be created below the pool top by properly choosing the disperser flow rate. The ONB heat flux is experimentally determined for channel gaps and upward flow velocities in the range 2mm-4mm and 3-16 cm/sec., respectively. Two alternatives to plume dispersion for reducing PTNA and a new correlation to determine the ONB heat flux in thin, rectangular channels under low-velocity, upward flow conditions are proposed. (Author)

Power distribution monitoring system in the boiling water cooled reactor core

International Nuclear Information System (INIS)

Leshchenko, Yu.I.; Sadulin, V.P.; Semidotskij, I.I.

1987-01-01

Consideration is being given to the system of physical power distribution monitoring, used during several years in the VK-50 tank type boiling water cooled reactor. Experiments were conducted to measure the ratios of detector prompt and activation currents, coefficients of detector relative sensitivity with respect to neutrons and effective cross sections of 103 Rh interaction with thermal and epithermal neutrons. Mobile self-powered detectors (SPD) with rhodium emitters are used as the power distribution detectors in the considered system. All detectors move simultaneously with constant rate in channels, located in fuel assembly central tubes, when conducting the measurements. It is concluded on the basis of analyzing the obtained data, that investigated system with calibrated SPD enables to monitor the absolute power distribution in fuel assemblies under conditions of boiling water cooled reactor and is independent of thermal engineering measurements conducted by in core instruments

Anisotropic nanolaminated CoNiFe cores integrated into microinductors for high-frequency dc–dc power conversion

International Nuclear Information System (INIS)

Kim, Jooncheol; Kim, Minsoo; Herrault, Florian; Kim, Jung-Kwun; Allen, Mark G

2015-01-01

This paper presents a rectangular, anisotropic nanolaminated CoNiFe core that possesses a magnetically hard axis in the long geometric axis direction. Previously, we have developed nanolaminated cores comprising tens to hundreds of layers of 300–1000 nm thick metallic alloys (i.e. Ni 80 Fe 20 or Co 44 Ni 37 Fe 19 ) based on sequential electrodeposition, demonstrating suppressed eddy-current losses at MHz frequencies. In this work, magnetic anisotropy was induced to the nanolaminated CoNiFe cores by applying an external magnetic field (50–100 mT) during CoNiFe film electrodeposition. The fabricated cores comprised tens to hundreds of layers of 500–1000 nm thick CoNiFe laminations that have the hard-axis magnetic property. Packaged in a 22-turn solenoid test inductor, the anisotropic core showed 10% increased effective permeability and 25% reduced core power losses at MHz operation frequency, compared to an isotropic core of the identical geometry. Operating the anisotropic nanolaminated CoNiFe core in a step-down dc–dc converter (15 V input to 5 V output) demonstrated 81% converter efficiency at a switching frequency of 1.1 MHz and output power of 6.5 W. A solenoid microinductor with microfabricated windings integrated with the anisotropic nanolaminated CoNiFe core was fabricated, demonstrating a constant inductance of 600 nH up to 10 MHz and peak quality factor exceeding 20 at 4 MHz. The performance of the microinductor with the anisotropic nanolaminated CoNiFe core is compared with other previously reported microinductors. (fast track communication)

Self-Powered Neutron and Gamma Detectors for In-Core Measurements

International Nuclear Information System (INIS)

Strindehag, O.

1971-11-01

The performance of various types of self-powered neutron and gamma detectors intended for control and power distribution measurements in water cooled reactors is discussed. The self-powered detectors are compared with other types of in-core detectors and attention is paid to such properties as neutron and gamma sensitivity, high-temperature performance, burn-up rate and time of response. Also treated are the advantages and disadvantages of using gamma detector data for power distribution calculations instead of data from neutron detectors. With regard to neutron-sensitive detectors, results from several long-term experiments with vanadium and cobalt detectors are presented. The results include reliability and stability data for these two detector types and the Co build-up in cobalt detectors. Experimental results which reveal the fast response of cobalt detectors are presented, and the use of cobalt detectors in reactor safety systems is discussed. Experience of the design and installation of complete flux probes, electronic units and data processing systems for power reactors is reported. The investigation of gamma-sensitive detectors includes detectors with emitters of lead, zirconium, magnesium and Inconel. Measured gamma sensitivities from calibrations both in a reactor and in a gamma cell are given, and the signal levels of self-powered neutron and gamma detectors when applied to power reactors are compared

Self-Powered Neutron and Gamma Detectors for In-Core Measurements

Energy Technology Data Exchange (ETDEWEB)

Strindehag, O

1971-11-15

The performance of various types of self-powered neutron and gamma detectors intended for control and power distribution measurements in water cooled reactors is discussed. The self-powered detectors are compared with other types of in-core detectors and attention is paid to such properties as neutron and gamma sensitivity, high-temperature performance, burn-up rate and time of response. Also treated are the advantages and disadvantages of using gamma detector data for power distribution calculations instead of data from neutron detectors. With regard to neutron-sensitive detectors, results from several long-term experiments with vanadium and cobalt detectors are presented. The results include reliability and stability data for these two detector types and the Co build-up in cobalt detectors. Experimental results which reveal the fast response of cobalt detectors are presented, and the use of cobalt detectors in reactor safety systems is discussed. Experience of the design and installation of complete flux probes, electronic units and data processing systems for power reactors is reported. The investigation of gamma-sensitive detectors includes detectors with emitters of lead, zirconium, magnesium and Inconel. Measured gamma sensitivities from calibrations both in a reactor and in a gamma cell are given, and the signal levels of self-powered neutron and gamma detectors when applied to power reactors are compared

Nuclear characteristic simulation device for reactor core

International Nuclear Information System (INIS)

Arakawa, Akio; Kobayashi, Yuji.

1994-01-01

In a simulation device for nuclear characteristic of a PWR type reactor, there are provided a one-dimensional reactor core dynamic characteristic model for simulating one-dimensional neutron flux distribution in the axial direction of the reactor core and average reactor power based on each of inputted signals of control rod pattern, a reactor core flow rate, reactor core pressure and reactor core inlet enthalphy, and a three-dimensional reactor core dynamic characteristic mode for simulating three-dimensional power distribution of the reactor core, and a nuclear instrumentation model for calculating read value of the nuclear instrumentation disposed in the reactor based on the average reactor core power and the reactor core three-dimensional power distribution. A one-dimensional neutron flux distribution in the axial direction of the reactor core, a reactor core average power, a reactor core three-dimensional power distribution and a nuclear instrumentation read value are calculated. As a result, the three-dimensional power distribution and the power level are continuously calculated. Further, since the transient change of the three-dimensional neutron flux distribution is calculated accurately on real time, more actual response relative to a power monitoring device of the reactor core and operation performance can be simulated. (N.H.)

Automatic determination of pressurized water reactor core loading patterns which maximize end-of-cycle reactivity within power peaking and burnup constraints

International Nuclear Information System (INIS)

Hobson, G.H.

1985-01-01

An automated procedure for determining the optimal core loading pattern for a pressurized water reactor which maximizes end-of-cycle k/sub eff/ while satisfying constraints on power peaking and discharge burnup has been developed. The optimization algorithm combines a two energy group, two-dimensional coarse-mesh finite difference diffusion theory neutronics model to simulate core conditions, a perturbation theory approach to determine reactivity, flux, power and burnup changes as a function of assembly shuffling, and Monte Carlo integer programming to select the optimal loading pattern solution. The core examined was a typical Cycle 2 reload with no burnable poisons. Results indicate that the core loading pattern that maximizes end-of-cycle k/sub eff/ results in a 5.4% decrease in fuel cycle costs compared with the core loading pattern that minimizes the maximum relative radial power peak

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-02-15

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

Research on instability design method without occurring boiling transition for hyper ABWR plants of extended core power density

International Nuclear Information System (INIS)

Okamoto, T.; Hotta, A.; Ama, T.

2008-01-01

The hyper ABWR (Advanced Boiling Water Reactor) project aims to develop an advanced BWR concept that is competitive in the global market with both highly economic and safety features. Expecting plant construction within the coming ten years, a research program for substantiating the basic design of a high core power density ABWR was conducted. By inheriting the conventional ABWR design, it is possible to reduce construction costs. In order to achieve the rated core power of over 1650MWe which is almost equivalent to that of the EPR (European Pressurized Water Reactor), the core power density of ABWR will be up-rated by at least 25%. Three key subjects linked to this target were recognized. They are, (1) fuel design applicable to the high power density core, (2) improvement of the evaluation method for the coupled neutronic and thermal-hydraulic instability under a wider power-flow operating range, and (3) improvement of the steam separator performance under high quality conditions. In this paper, the second subject has been focused on. In the second subject, the uncertainty approach was introduced in the instability analysis where the best-estimate plant simulator was combined with a direct prediction of boiling transition by the sub-channel code. By employing the CSAU like method, a safety evaluation system that enables to include influences of uncertainties has been developed. Based on the correlation between the time margin for reaching the boiling transition under power oscillations and the decay ratio in the power-flow operation map, an automatic power oscillation suppressing system was designed. The set-point for activating suppression mechanisms (i.e. scram or SRI) could be determined based on this correlation. It was proposed that the present conservative acceptance criterion of the deterministic decay ratio can be replaced with a more rational one of the time margin with including uncertainties. (author)

Influence of core NA on thermal-induced mode instabilities in high power fiber amplifiers

International Nuclear Information System (INIS)

Tao, Rumao; Ma, Pengfei; Wang, Xiaolin; Zhou, Pu; Liu, Zejin

2015-01-01

We report on the influence of core NA on thermal-induced mode instabilities (MI) in high power fiber amplifiers. The influence of core NA and the V-parameter on MI has been investigated numerically. It shows that core NA has a larger influence on MI for fibers with a smaller core-cladding-ratio, and the influence of core NA on the threshold is more obvious when the amplifiers are pumped at 915 nm. The dependence of the threshold on the V-parameter revealed that the threshold increases linearly as the V-parameter decreases when the V-parameter is larger than 3.5, and the threshold shows an exponential increase as the V-parameter decreases when the V-parameter is less than 3.5. We also discussed the effect of linewidth on MI, which indicates that the influence of linewidth can be neglected for a linewidth smaller than 1 nm when the fiber core NA is smaller than 0.07 and the fiber length is shorter than 20 m. Fiber amplifiers with different core NA were experimentally analyzed, which agreed with the theoretical predictions. (letter)

Ex-vessel molten core debris interactions at CANDU nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Lewis, M J; Oyinloye, J O; Chambers, I [Electrowatt Consulting Engineers and Scientists, Warrington, Cheshire (United Kingdom); Scott, C K [Atlantic Nuclear Services, Fredericton, NB (Canada); Omar, A M [Atomic Energy Control Board, Ottawa, ON (Canada)

1991-12-31

Currently, the Atomic Energy Control Board (AECB) of Canada is sponsoring a project with a long term objective of obtaining an evaluation, independent of the industry, of the consequences to the public and the environment of postulated severe accidents at a Canadian nuclear power plant. Phase 1 of this project is a scoping study conducted to establish the relative consequences of a number of postulated event sequences. The studies in this paper model a multi-unit CANDU reactor at which pre-defined initiating events and their consequences could lead to severe core damage and relocation of the core debris onto the floor of the concrete reactor vault. Depending on the accident sequence assumptions made, an overlying pool of water may or may not be present. The US-NRC computer code CORCON Mod 2.0 was used to calculate the behaviour of the core material interacting with the concrete. The code calculates the decomposition of concrete by the molten core, and also the gases produced, which are released into the containment. The challenges to containment integrity are described, from the viewpoint of foundation decomposition and failure due to overpressure. The containment thermal-hydraulic behaviour is examined using an in-house computer code (CREM) written for this purpose. It is found that the containment envelope, in the absence of mitigating operator actions or design safety features, even for a case involving early core disassembly with the vacuum building unavailable, is unlikely to be failed within the 48 hours time frame examined. The paper identifies several areas for improvement in the models for future studies of core-concrete interactions for CANDU reactor plants. (author). 8 refs., 1 tab., 5 figs.

Ex-vessel molten core debris interactions at CANDU nuclear power plants

International Nuclear Information System (INIS)

Lewis, M.J.; Oyinloye, J.O.; Chambers, I.; Scott, C.K.; Omar, A.M.

1990-01-01

Currently, the Atomic Energy Control Board (AECB) of Canada is sponsoring a project with a long term objective of obtaining an evaluation, independent of the industry, of the consequences to the public and the environment of postulated severe accidents at a Canadian nuclear power plant. Phase 1 of this project is a scoping study conducted to establish the relative consequences of a number of postulated event sequences. The studies in this paper model a multi-unit CANDU reactor at which pre-defined initiating events and their consequences could lead to severe core damage and relocation of the core debris onto the floor of the concrete reactor vault. Depending on the accident sequence assumptions made, an overlying pool of water may or may not be present. The US-NRC computer code CORCON Mod 2.0 was used to calculate the behaviour of the core material interacting with the concrete. The code calculates the decomposition of concrete by the molten core, and also the gases produced, which are released into the containment. The challenges to containment integrity are described, from the viewpoint of foundation decomposition and failure due to overpressure. The containment thermal-hydraulic behaviour is examined using an in-house computer code (CREM) written for this purpose. It is found that the containment envelope, in the absence of mitigating operator actions or design safety features, even for a case involving early core disassembly with the vacuum building unavailable, is unlikely to be failed within the 48 hours time frame examined. The paper identifies several areas for improvement in the models for future studies of core-concrete interactions for CANDU reactor plants. (author). 8 refs., 1 tab., 5 figs

Design study of nuclear power systems for deep space explorers. (2) Electricity supply capabilities of solid cores

International Nuclear Information System (INIS)

Yamaji, Akifumi; Takizuka, Takakazu; Nabeshima, Kunihiko; Iwamura, Takamichi; Akimoto, Hajime

2009-01-01

This study has been carried out in series with the other study, 'Criticality of Low Enriched Uranium Fueled Core' to explore the possibilities of a solid reactor electricity generation system for supplying propulsion power of a deep space explorer. The design ranges of two different systems are determined with respect to the electric power, the radiator mass, and the operating temperatures of the heat-pipes and thermoelectric converters. The two systems are the core surface cooling with heat-pipe system (CSHP), and the core direct cooling with heat-pipe system (CDHP). The evaluated electric powers widely cover the 1 to 100 kW range, which had long been claimed to be the range that lacked the power sources in space. Therefore, the concepts shown by this study may lead to a breakthrough of the human activities in space. The working temperature ranges of the main components, namely the heat-pipes and thermoelectric converters, are wide and covers down to relatively low temperatures. This is desirable from the viewpoints of broadening the choices, reducing the development needs, and improving the reliabilities of the devices. Hence, it is advantageous for an early establishment of the concept. (author)

Sensitivity Analysis of Core Damage by Loss of Auxiliary Feed Water during the Extended Loss of All AC Power

Energy Technology Data Exchange (ETDEWEB)

Jeon, Woo Jae; Chung, Soon Il; Hwang, Su Hyun; Lee, Kyung Jin; Lee, Byung Chul [FNC Tech., Yongin (Korea, Republic of); Yun, Duk Joo; Lee, Seung Chan [Korea Hydro and Nuclear Power Co. Ltd., Daejeon (Korea, Republic of)

2015-10-15

In this study, the reactor core damage time for OPR1000 type Nuclear Power Plant (NPP) was analyzed to develop a strategy to handle ELAP and to apply to the EOP. The reactor core damage time in the ELAP condition was calculated according to the time of Auxiliary Feedwater (AFW) loss. Fukushima accident was caused by long hours of Station Black Out (SBO) caused by natural disaster beyond Design Based Accident (DBA) criteria. It led to the reactor core damage. After the accident, the regulatory authorities of each country (Japan, US, EU, IAEA, and etc.) recommended developing the necessary systems and strategies in order to cover up the Extended Loss of All AC Power (ELAP) such as one occurred in the Fukushima accident. And the need of procedure or guideline to cope with ELAP has been raised through the stress test for Wolsong Nuclear Power Plant unit 1. Current Emergency Operating Procedures (EOP) used in domestic nuclear power plant are seemed to be insufficient to cope with ELAP. Therefore, it has been required to be improved. As the result, the time of AFW loss in the ELAP condition influences greatly on core damage time.

Calculation of core axial power shapes using alternating conditional expectation algorithm

International Nuclear Information System (INIS)

Lee, Eun Ki; Kim, Yong Hee; Cha, Kune Ho; Park, Moon Kyu

1998-01-01

We have introduced the alternating conditional expectation (ACE) algorithm in the method of reconstructing 20 node axial power shapes from five level detector powers. The ACE algorithm was used to find the optimal relationships between each plane power and normalized five detector powers. The obtained all optimal transformations had simple forms to be represented with polynomials. The reference axial power shapes and simulated detector powers were drawn out of the 3-dimensional results of Reactor Operation and Control Simulation (ROCS) code for various core states. By the ACE algorithm, we obtained the optimal relationship between dependent variable plane power, y, and independent variable detector powers, {Di, i=1,...,5 without any preprocessing, where a total of ≅3490 data sets per each cycle of YongGwang Nuclear (YGN) Power Plant units 3 and 4 are used. To test the validity and accuracy of the new method, about 21,200 cases of reconstructed axial power shapes are compared to original ROCS axial power shapes, and they are also contrasted with those obtained by Fourier fitting method (FFM). The average error of root mean square (rms), axial peak (DFZ), and axial shape index (DASI) of our new method for total 21204 data cases are 0.81%, 0.51% and 0.00204, while FFM 2.29%, 2.37% and 0.00264, respectively. The evaluation results for the data sets not used in the ACE transformations also show that the accuracy of new method is much better than that of FFM

Toward power scaling in an acetylene mid-infrared hollow-core optical fiber gas laser: effects of pressure, fiber length, and pump power

Science.gov (United States)

Weerasinghe, H. W. Kushan; Dadashzadeh, Neda; Thirugnanasambandam, Manasadevi P.; Debord, Benoît.; Chafer, Matthieu; Gérôme, Frédéric; Benabid, Fetah; Corwin, Kristan L.; Washburn, Brian R.

2018-02-01

The effect of gas pressure, fiber length, and optical pump power on an acetylene mid-infrared hollow-core optical fiber gas laser (HOFGLAS) is experimentally determined in order to scale the laser to higher powers. The absorbed optical power and threshold power are measured for different pressures providing an optimum pressure for a given fiber length. We observe a linear dependence of both absorbed pump energy and lasing threshold for the acetylene HOFGLAS, while maintaining a good mode quality with an M-squared of 1.15. The threshold and mode behavior are encouraging for scaling to higher pressures and pump powers.

Introduction of virtual detectors for core monitoring system of korean standard nuclear power plant

International Nuclear Information System (INIS)

Eun, Ki Lee.; Yong, Hee Kim.; Jybe, Ho Cha.; Moon, Ghu Park.

2000-01-01

A novel algorithm known as the virtual detector method (VDM) is introduced to reconstruct the axial power shape (APS) for the on-line core monitoring system of the Korean Standard Nuclear Power Plant (KSNP). A pure statistical method (SM) is also introduced and the results are compared with the currently implemented five-mode Fourier fitting method (FFM). VDM adopts nine virtual detector informations coupled with a regression model based on the Alternating Conditional Expectation (ACE) algorithm. VDM uses Fourier fitting with the information of nine virtual detectors expanded from the currently implemented FFM, which uses five-level detector information. By introducing virtual detectors, we can increase the number of axial detectors, and thus expect the computational errors of APS to be reduced. The two methods (SM and VDM) are applied to in-core mapping data from six cycles of Yong Gwang nuclear power plant Units 3 and 4. For ∼ 3500 cases of APSs extracted from a cycle of operation which is simulated by a three-dimensional nodal code, the accuracy of the three methods (SM, VDM, FFM) is compared. The average root mean square (RMS) error and average of axial peaking error of SM and VDM resulted in reduction of more than 50 % and 70 %, respectively, relative to FFM. VDM and SM also show more realistic axial profiles and predict more accurate axial peaking than FFM. These improvements can contribute to a larger thermal margin. SM shows the most accurate results for all cases. VDM can almost obtain the same results as SM, and using far fewer computation steps. VDM can be a useful tool for precisely reconstructing axial power shapes in a core monitoring system. (authors)

The Multi一physics Research on I ron一Core Vibration Noise of Power Reactor

Directory of Open Access Journals (Sweden)

LI U Ja

2017-02-01

Full Text Available On the basis of theoretical research releted to the magnetostriction and maxwell’.s equations，the fi- nite element coupling in the transient electromagnetic field coupling，structure and sound field coupling has been developed In thts paper by using the flnlte element sOftWare CO}IS01., Whleh establish a serles three-phase COT’e re- actor model, to analyzing the power frequency magnetic field distribution，core magnetostrictive displacement，max- well force displacement and sound pressure level of the three-phase series core reactor under the power frequency working state. According to transient magnetic field distribution in the simulation of the reactor，the magnetic flux density distribution inside the reactor and the vibration displacement distribution are calculated，the acoustic field distribution is measured alao. It is shown that physical field simulation results and measured data are basically in consisent by experiment，it is proved multi-physics coupling is an effective method for forecast of noise.

Improving the calculated core stability by the core nuclear design optimization

International Nuclear Information System (INIS)

Partanen, P.

1995-01-01

Three different equilibrium core loadings for TVO II reactor have been generated in order to improve the core stability properties at uprated power level. The reactor thermal power is assumed to be uprated from 2160 MW th to 2500 MW th , which moves the operating point after a rapid pump rundown where the core stability has been calculated from 1340 MW th and 3200 kg/s to 1675 MW th and 4000 kg/s. The core has been refuelled with ABB Atom Svea-100 -fuel, which has 3,64% w/o U-235 average enrichment in the highly enriched zone. PHOENIX lattice code has been used to provide the homogenized nuclear constants. POLCA4 static core simulator has been used for core loadings and cycle simulations and RAMONA-3B program for simulating the dynamic response to the disturbance for which the stability behaviour has been evaluated. The core decay ratio has been successfully reduced from 0,83 to 0,55 mainly by reducing the power peaking factors. (orig.) (7 figs., 1 tab.)

Partial thorium loading in the initial core of Kakrapar atomic power reactor

International Nuclear Information System (INIS)

Balakrishnan, M.R.

1993-01-01

The first unit of Kakrapar nuclear power station has gone critical with some thorium oxide fuel bundles loaded in its core. The thorium helps to flatten the power by reducing neutron flux in the centre of the reactor. However, the placing of the thorium had to be planned with care, because if the neutron flux at a point where a safety rod is located is depressed, the reactivity worth of the safety rod gets reduced. Using a dynamic programing approach, the Reactor Engineering Division of Bhabha Atomic Research Centre worked out a satisfactory configuration for loading the thorium bundles

Comparison of Battery-Powered and Manual Bone Biopsy Systems for Core Needle Biopsy of Sclerotic Bone Lesions.

Science.gov (United States)

Cohen, Micah G; McMahon, Colm J; Kung, Justin W; Wu, Jim S

2016-05-01

The purpose of this study was to compare manual and battery-powered bone biopsy systems for diagnostic yield and procedural factors during core needle biopsy of sclerotic bone lesions. A total of 155 consecutive CT-guided core needle biopsies of sclerotic bone lesions were performed at one institution from January 2006 to November 2014. Before March 2012, lesions were biopsied with manual bone drill systems. After March 2012, most biopsies were performed with a battery-powered system and either noncoaxial or coaxial biopsy needles. Diagnostic yield, crush artifact, CT procedure time, procedure radiation dose, conscious sedation dose, and complications were compared between the manual and battery-powered core needle biopsy systems by Fisher exact test and t test. One-way ANOVA was used for subgroup analysis of the two battery-powered systems for procedure time and radiation dose. The diagnostic yield for all sclerotic lesions was 60.0% (93/155) and was significantly higher with the battery-powered system (73.0% [27/37]) than with the manual systems (55.9% [66/118]) (p = 0.047). There was no significant difference between the two systems in terms of crush artifact, procedure time, radiation dose, conscious sedation administered, or complications. In subgroup analysis, the coaxial battery-powered biopsies had shorter procedure times (p = 0.01) and lower radiation doses (p = 0.002) than the coaxial manual systems, but the noncoaxial battery-powered biopsies had longer average procedure times and higher radiation doses than the coaxial manual systems. In biopsy of sclerotic bone lesions, use of a battery-powered bone drill system improves diagnostic yield over use of a manual system.

Axial power distribution calculation using a neural network in the nuclear reactor core

Energy Technology Data Exchange (ETDEWEB)

Kim, Y H; Cha, K H; Lee, S H [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

1998-12-31

This paper is concerned with an algorithm based on neural networks to calculate the axial power distribution using excore detector signals in the nuclear reactor core. The fundamental basis of the algorithm is that the detector response can be fairly accurately estimated using computational codes. In other words, the training set, which represents relationship between detector signals and axial power distributions, for the neural network can be obtained through calculations instead of measurements. Application of the new method to the Yonggwang nuclear power plant unit 3 (YGN-3) shows that it is superior to the current algorithm in place. 7 refs., 4 figs. (Author)

Axial power distribution calculation using a neural network in the nuclear reactor core

Energy Technology Data Exchange (ETDEWEB)

Kim, Y. H.; Cha, K. H.; Lee, S. H. [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

1997-12-31

This paper is concerned with an algorithm based on neural networks to calculate the axial power distribution using excore detector signals in the nuclear reactor core. The fundamental basis of the algorithm is that the detector response can be fairly accurately estimated using computational codes. In other words, the training set, which represents relationship between detector signals and axial power distributions, for the neural network can be obtained through calculations instead of measurements. Application of the new method to the Yonggwang nuclear power plant unit 3 (YGN-3) shows that it is superior to the current algorithm in place. 7 refs., 4 figs. (Author)

Noise variation by compressive stress on the model core of power transformers

Energy Technology Data Exchange (ETDEWEB)

Mizokami, Masato, E-mail: mizokami.g76.masato@jp.nssmc.com; Kurosaki, Yousuke

2015-05-01

The reduction of audible noise generated by cores for power transformers has been required due to environmental concern. It is known that compressive stress in the rolling direction of electrical steel affects magnetostriction and it can result in an increase in noise level. In this research, the effect of compressive stress to noise was investigated on a 3-phase 3-limb model core. Compressive stress was applied in the rolling direction of the limbs from the outside of the core. It increased the sound pressure levels and the slope of the rise was about 2 dBA/MPa. Magnetostriction on single sheet samples was also measured under compressive stress and the harmonic components of the magnetostriction were compared with those of noise. It revealed that the variation in magnetostriction with compressive stress did not entirely correspond to that in noise. In one of the experiments, localized bending happened on one limb during compressing the core. While deformation of the core had not been intended, the noise was measured. The deformation increased the noise by more than 10 dBA and it occurred on most of the harmonic components. - Highlights: • Audible noise was measured on a model core to which compressive stress was applied. • The stress in the rolling direction of the steel causes a rise in noise level. • The slope of the rise in sound pressure level up to 2.5 MPa is about 2 dBA/MPa. • Variation in magnetostriction by stress does not entirely agree with that in noise. • Bend arisen in the core causes an extreme increase in noise.

Noise variation by compressive stress on the model core of power transformers

International Nuclear Information System (INIS)

Mizokami, Masato; Kurosaki, Yousuke

2015-01-01

The reduction of audible noise generated by cores for power transformers has been required due to environmental concern. It is known that compressive stress in the rolling direction of electrical steel affects magnetostriction and it can result in an increase in noise level. In this research, the effect of compressive stress to noise was investigated on a 3-phase 3-limb model core. Compressive stress was applied in the rolling direction of the limbs from the outside of the core. It increased the sound pressure levels and the slope of the rise was about 2 dBA/MPa. Magnetostriction on single sheet samples was also measured under compressive stress and the harmonic components of the magnetostriction were compared with those of noise. It revealed that the variation in magnetostriction with compressive stress did not entirely correspond to that in noise. In one of the experiments, localized bending happened on one limb during compressing the core. While deformation of the core had not been intended, the noise was measured. The deformation increased the noise by more than 10 dBA and it occurred on most of the harmonic components. - Highlights: • Audible noise was measured on a model core to which compressive stress was applied. • The stress in the rolling direction of the steel causes a rise in noise level. • The slope of the rise in sound pressure level up to 2.5 MPa is about 2 dBA/MPa. • Variation in magnetostriction by stress does not entirely agree with that in noise. • Bend arisen in the core causes an extreme increase in noise

A benefit assessment of using in-core neutron detector signals in core protection calculator system (CPCS)

International Nuclear Information System (INIS)

Han, S.; Park, S.J.; Seong, P.H.

1997-01-01

A Core Protection Calculator System (CPCS) is a digital computer based safety system generating trip signals based on the calculation of Departure from Nucleate Boiling Ratio (DNBR) and Local Power Density (LPD). Currently, CPCS uses ex-core detector signals for core power calculation and it has some uncertainties. In this study, in-core detector signals which directly measure inside flux of core are applied to CPCS to get more accurate power distribution profile, DNBR and LPD. In order to improve axial power distribution calculation, piece-wise cubic Spline method is applied; from the 40 nodes of expanded signals, more accurate and detailed core information can be provided. Simulation is carried out to verify its applicability to power distribution calculation. Simulation result shows that the improved method reduces the calculational uncertainties significantly and it allows larger operational margin. It is also expected that no power reduction is required while Core Operating Limit Supervisory System (COLSS) is out-of-service due to reduced uncertainties when the improved method is applied. In this study, a quantitative economic benefit assessment of using in-core neutron detector signals is also carried out. (authors)

A benefit assessment of using in-core neutron detector signals in core protection calculator system(CPCS)

International Nuclear Information System (INIS)

Han, Seung

1996-02-01

A Core Protection Calculator System(CPCS) is a digital computer based safety system generating trip signals based on the calculation of Departure from Nucleate Boiling Ratio(DNBR) and Local Power Density(LPD). Currently, CPCS uses ex-core detector signals for core power calculation and it has some uncertainties. In this study, In-core detector signals which directly measure inside flux of core are applied to CPCS to get more accurate power distribution profile, DNBR and LPD. In order to improve axial power distribution calculation, piecewise cubic spline method is applied: From the 40 nodes of expanded signals, more accurate and detailed core information can be provided. Simulation is carried out to verify its applicability to power distribution calculation. Simulation result shows that the improved method reduces the calculational uncertainties significantly and it allows larger operational margin. It is also expected that no power reduction is required while Core Operating Limit Supervisory System(COLSS) is out-of-service due to reduced uncertainties when the improved method is applied. In this study, a quantitative economic benefit assessment of using in-core neutron detector signals is also carried out

A combined compensation method for the output voltage of an insulated core transformer power supply

Energy Technology Data Exchange (ETDEWEB)

Yang, L.; Yang, J., E-mail: jyang@mail.hust.edu.cn; Liu, K. F.; Qin, B.; Chen, D. Z. [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China)

2014-06-15

An insulated core transformer (ICT) power supply is an ideal high-voltage generator for irradiation accelerators with energy lower than 3 MeV. However, there is a significant problem that the structure of the segmented cores leads to an increase in the leakage flux and voltage differences between rectifier disks. A high level of consistency in the output of the disks helps to achieve a compact structure by improving the utilization of both the rectifier components and the insulation distances, and consequently increase the output voltage of the power supply. The output voltages of the disks which are far away from the primary coils need to be improved to reduce their inhomogeneity. In this study, by investigating and comparing the existing compensation methods, a new combined compensation method is proposed, which increases the turns on the secondary coils and employs parallel capacitors to improve the consistency of the disks, while covering the entire operating range of the power supply. This method turns out to be both feasible and effective during the development of an ICT power supply. The non-uniformity of the output voltages of the disks is less than 3.5% from no-load to full-load, and the power supply reaches an output specification of 350 kV/60 mA.

Determining the axial power profile of partly flooded fuel in a compact core assembled in reactor LR-0

International Nuclear Information System (INIS)

Košťál, Michal; Švadlenková, Marie; Baroň, Petr; Rypar, Vojtěch; Milčák, Ján

2016-01-01

Highlights: • Fission density in partly flooded compact core. • Calculation of fission density axial profile. • Significant calculational under prediction of experimental axial profile. - Abstract: Measurement and calculation of the axial power profile near the boundary of a moderated and non-moderated core is used to analyze the suitability of the neutron-physical process description, mainly the angular cross-section of a water-moderated uranium system. This is also an important issue because it affects the radiation situation above the partly flooded core of a water-moderated reactor. Axial power profiles of various fuel pins irradiated on reactor LR-0 were measured and the results were compared with MCNP6 code calculations using the ENDF/B-VII.0 nuclear data library. The calculated power profile in positions above the moderator level significantly underestimates experimental results. This might be caused by an improper description of the angular distribution of scattered neutrons in a water-moderated uranium system.

Fuel assembly outlet temperature profile influence on core by-pass flow and power distribution determination in WWER -440 reactors

International Nuclear Information System (INIS)

Petenyi, V.; Klucarova, K.; Remis, J.

2003-01-01

The in core instrumentation of the WWER-440 reactors consists of the thermocouple system and the system of self powered detectors (SPD). The thermocouple systems are positioned about 50 cm above the fuel bundle upper flow-mixing grid. The usual assumption is that, the coolant is well mixed in the Tc location, i.e. the temperature is constant through the flow cross-section area. The present evaluations by using the FLUENT 5.5.14 code reveal that, this assumption is not fulfilled. There exists a temperature profile that depends on fuel assembly geometry and on inner power profile of the fuel assembly. The paper presents the estimation of this effect and its influence on the core power distribution and the core by-pass flow determination. Comparison with measurements in Mochovce NPP will also be a part of this presentation (Authors)

Dual-core Itanium Processor

CERN Multimedia

2006-01-01

Intel’s first dual-core Itanium processor, code-named "Montecito" is a major release of Intel's Itanium 2 Processor Family, which implements the Intel Itanium architecture on a dual-core processor with two cores per die (integrated circuit). Itanium 2 is much more powerful than its predecessor. It has lower power consumption and thermal dissipation.

Heterogeneous gas core reactor

International Nuclear Information System (INIS)

Han, K.I.

1977-01-01

Preliminary investigations of a heterogeneous gas core reactor (HGCR) concept suggest that this potential power reactor offers distinct advantages over other existing or conceptual reactor power plants. One of the most favorable features of the HGCR is the flexibility of the power producing system which allows it to be efficiently designed to conform to a desired optimum condition without major conceptual changes. The arrangement of bundles of moderator/coolant channels in a fissionable gas or mixture of gases makes a truly heterogeneous nuclear reactor core. It is this full heterogeneity for a gas-fueled reactor core which accounts for the novelty of the heterogeneous gas core reactor concept and leads to noted significant advantages over previous gas core systems with respect to neutron and fuel economy, power density, and heat transfer characteristics. The purpose of this work is to provide an insight into the design, operating characteristics, and safety of a heterogeneous gas core reactor system. The studies consist mainly of neutronic, energetic and kinetic analyses of the power producing and conversion systems as a preliminary assessment of the heterogeneous gas core reactor concept and basic design. The results of the conducted research indicate a high potential for the heterogeneous gas core reactor system as an electrical power generating unit (either large or small), with an overall efficiency as high as 40 to 45%. The HGCR system is found to be stable and safe, under the conditions imposed upon the analyses conducted in this work, due to the inherent safety of ann expanding gaseous fuel and the intrinsic feedback effects of the gas and water coolant

Conceptual study of axial offset fluctuations upon stepwise power changes in a thorium–plutonium core to improve load-following conditions

International Nuclear Information System (INIS)

Lau, Cheuk Wah; Dykin, Victor; Nylén, Henrik; Björk, Klara Insulander; Sandberg, Urban

2014-01-01

Highlights: • Thorium–plutonium mixed oxide to improve nuclear reactors load-following capability. • SIMULATE-3 was the main calculation tool. • The Ringhals-3 PWR unit in Sweden was used as a reference. • Lower xenon poisoning and shorter reactor dead time. - Abstract: The increased share of renewable energy, such as wind and solar power, will increase the demand for load-following power sources, and nuclear reactors could be one option. However, during rapid load-following events, traditional UOX cores could be restricted by the volatile oscillation of the power distribution. Therefore, a conceptual study on stability properties of Th-MOX PWR concerning axial offset power excursion during load-following events are investigated and discussed. The study is performed in SIMULATE-3 for a realistic PWR core (Ringhals-3) at the end of cycle, where the largest amplitude of the axial offset oscillations is expected. It is shown that the Th-MOX core possesses much better stability characteristics and shorter reactor dead time compared with a traditional UOX core, and the main reasons are the lower sensitivity to perturbations in the neutron spectrum, lower xenon poisoning and lower thermal neutron flux

Safety-related investigations on power distribution in MOX fuel elements in LWR cores

International Nuclear Information System (INIS)

Kramer, E.; Langenbuch, S.

1991-01-01

For the concept of thermal recycling various fuel assembly designs have been developped during the last years. An overview is given describing the present status of MOX-fuel assembly design for PWR and BWR. The local power distribution within the MOX-fuel assembly and influences between neighbouring MOX- and Uranium fuel assemblies have been analyzed by own calculations. These investigations are limited to specific aspects of the spatial power distribution, which are related to the use of MOX-fuel assemblies within the reactor core of LWR. (orig.) [de

Core design experience of WWER-440 reactors when they working on increased power level

International Nuclear Information System (INIS)

Adeev, V.; Panov, A.; Melenchuk, I.

2015-01-01

The Kola NPP continues commercial operation of 2nd generation fuel (FA-2) and trial operation of 3rd generation fuel (FA-3), which has a number of design features providing the best operational characteristics. This report gives the results of VVER-440 core operation with FA-2 and FA-3 with enrichment increased up to 4.87%, and at the power level uprated to 107% of nominal power level. Brief analysis of obtained data is carried out. Peculiarities and techniques of developing loading patterns with new types of nuclear fuel for operation at the uprated power level are reviewed. (authors)

Evaluation of local power distribution with fine-mesh core model for the HTTR

International Nuclear Information System (INIS)

Murata, Isao; Yamashita, Kiyonobu; Maruyama, So; Shindo, Ryuichi; Fujimoto, Nozomu; Sudo, Yukio; Nakata, Tetsuo.

1991-01-01

An evaluation method of the local power distribution was developed considering the radial and axial heterogeneity caused by fuel rods, BP rods and block end graphite for the High Temperature Engineering Test Reactor (HTTR) in Japan Atomic Energy Research Institute (JAERI). The evaluation method was verified through the analyses of critical assembly experiments. A good agreement was obtained between calculations and measurements and the difference was less than 3 % on the power distribution. This method was applied to the core design for the HTTR to evaluate the maximum fuel temperature. From these results, it was confirmed that this evaluation method has an enough accuracy and is able to predict the detailed power distribution of the HTTR. (author)

Development of Liquid-Vapor Core Reactors with MHD Generator for Space Power and Propulsion Applications

International Nuclear Information System (INIS)

Samim Anghaie

2002-01-01

Any reactor that utilizes fuel consisting of a fissile material in a gaseous state may be referred to as a gaseous core reactor (GCR). Studies on GCRs have primarily been limited to the conceptual phase, mostly due to budget cuts and program cancellations in the early 1970's. A few scientific experiments have been conducted on candidate concepts, primarily of static pressure fissile gas filling a cylindrical or spherical cavity surrounded by a moderating shell, such as beryllium, heavy water, or graphite. The main interest in this area of nuclear power generation is for space applications. The interest in space applications has developed due to the promise of significant enhancement in fuel utilization, safety, plant efficiency, special high-performance features, load-following capabilities, power conversion optimization, and other key aspects of nuclear power generation. The design of a successful GCR adapted for use in space is complicated. The fissile material studied in the pa st has been in a fluorine compound, either a tetrafluoride or a hexafluoride. Both of these molecules have an impact on the structural material used in the making of a GCR. Uranium hexafluoride as a fuel allows for a lower operating temperature, but at temperatures greater than 900K becomes essentially impossible to contain. This difficulty with the use of UF6 has caused engineers and scientists to use uranium tetrafluoride, which is a more stable molecule but has the disadvantage of requiring significantly higher operating temperatures. Gas core reactors have traditionally been studied in a steady state configuration. In this manner a fissile gas and working fluid are introduced into the core, called a cavity, that is surrounded by a reflector constructed of materials such as Be or BeO. These reactors have often been described as cavity reactors because the density of the fissile gas is low and criticality is achieved only by means of the reflector to reduce neutron leakage from the core

Harmonic Domain Modelling of Transformer Core Nonlinearities Using the DIgSILENT PowerFactory Software

DEFF Research Database (Denmark)

Bak, Claus Leth; Bak-Jensen, Birgitte; Wiechowski, Wojciech

2008-01-01

This paper demonstrates the results of implementation and verification of an already existing algorithm that allows for calculating saturation characteristics of singlephase power transformers. The algorithm was described for the first time in 1993. Now this algorithm has been implemented using...... the DIgSILENT Programming Language (DPL) as an external script in the harmonic domain calculations of a power system analysis tool PowerFactory [10]. The algorithm is verified by harmonic measurements on a single-phase power transformer. A theoretical analysis of the core nonlinearities phenomena...... in single and three-phase transformers is also presented. This analysis leads to the conclusion that the method can be applied for modelling nonlinearities of three-phase autotransformers....

Automatic determination of pressurized water reactor core loading patterns that maximize beginning-of-cycle reactivity within power-peaking and burnup constraints

International Nuclear Information System (INIS)

Hobson, G.H.; Turinsky, P.J.

1986-01-01

Computational capability has been developed to automatically determine a good estimate of the core loading pattern, which minimizes fuel cycle costs for a pressurized water reactor (PWR). Equating fuel cycle cost minimization with core reactivity maximization, the objective is to determine the loading pattern that maximizes core reactivity while satisfying power peaking, discharge burnup, and other constraints. The method utilizes a two-dimensional, coarse-mesh, finite difference scheme to evaluate core reactivity and fluxes for an initial reference loading pattern. First-order perturbation theory is applied to determine the effects of assembly shuffling on reactivity, power distribution, end-of-cycle burnup. Monte Carlo integer programming is then used to determine a near-optimal loading pattern within a range of loading patterns near the reference pattern. The process then repeats with the new loading pattern as the reference loading pattern and terminates when no better loading pattern can be determined. The process was applied with both reactivity maximization and radial power-peaking minimization as objectives. Results on a typical large PWR indicate that the cost of obtaining an 8% improvement in radial power-peaking margin is ≅2% in fuel cycle costs, for the reload core loaded without burnable poisons that was studied

Design and test of-80 kV snubber core assemblies for MFTF sustaining-neutral-beam power supplies

International Nuclear Information System (INIS)

Bishop, S.R.; Mayhall, D.J.; Wilson, J.H.; De Vore, K.R.; Ross, R.I.; Sears, R.G.

1981-01-01

Core snubbers, located near the neutral beam source ends of the Mirror Fusion Test Facility (MFTF) Sustaining Neutral Beam Power Supply System (SNBPSS) source cables, protect the neutral beam source extractor grid wires from overheating and sputtering during internal sparkdowns. The snubbers work by producing an induced counter-emf which limits the fault current and by absorbing the capacitive energy stored on the 80 kV source cables and power supplies. A computer program STACAL was used in snubber magnetic design to choose appropriate tape wound cores to provide 400 Ω resistance and 25 J energy absorption. The cores are mounted horizontally in a dielectric structure. The central source cable bundle passes through the snubber and terminates on three copper buses. Multilam receptacles on the buses connect to the source module jumper cables. Corona rings and shields limit electric field stresses to allow close clearances between snubbers

Accelerator driven sub-critical core

Science.gov (United States)

McIntyre, Peter M; Sattarov, Akhdiyor

2015-03-17

Systems and methods for operating an accelerator driven sub-critical core. In one embodiment, a fission power generator includes a sub-critical core and a plurality of proton beam generators. Each of the proton beam generators is configured to concurrently provide a proton beam into a different area of the sub-critical core. Each proton beam scatters neutrons within the sub-critical core. The plurality of proton beam generators provides aggregate power to the sub-critical core, via the proton beams, to scatter neutrons sufficient to initiate fission in the sub-critical core.

The power-sharing formula for a seed/blanket core-resolution of a paradox

International Nuclear Information System (INIS)

Radkowsky, A.; Segev, M.; Galperin, A.

1986-01-01

The ''classical'' formula for the sharing of power between a seed and blanket was based on the two-group diffusion theory model and gave good agreement with experiments conducted in the original Shippingport program and with transport theory. Recently an extensive series of calculations on seed/blanket assemblies showed that the power sharing deviates widely from the classical formula but paradoxically is in good agreement with the one-group formula, which neglects the back leakage of thermal neutrons from the blanket to the seed. The power-sharing formula has now been rederived, and the paradox is resolved by taking into account epithermal absorptions in the seeds. The diffusion theory model is important as a guide to formulating innovative concepts for improved core designs

Off-resonance frequency operation for power transfer in a loosely coupled air core transformer

Science.gov (United States)

Scudiere, Matthew B

2012-11-13

A power transmission system includes a loosely coupled air core transformer having a resonance frequency determined by a product of inductance and capacitance of a primary circuit including a primary coil. A secondary circuit is configured to have a substantially same product of inductance and capacitance. A back EMF generating device (e.g., a battery), which generates a back EMF with power transfer, is attached to the secondary circuit. Once the load power of the back EMF generating device exceeds a certain threshold level, which depends on the system parameters, the power transfer can be achieved at higher transfer efficiency if performed at an operating frequency less than the resonance frequency, which can be from 50% to 95% of the resonance frequency.

CORE DESIGNS OF ABWR FOR PROPOSED OF THE FIRST NUCLEAR POWER PLANT IN INDONESIA

Directory of Open Access Journals (Sweden)

Yohannes Sardjono

2015-04-01

Full Text Available Indonesia as an archipelago has been experiencing high growth industry and energy demand due to high population growth, dynamic economic activities. The total population is around 230 million people and 75 % to the total population is living in Java. The introduction of Nuclear Power Plant on Java Bali electricity grid will be possible in 2022 for 2 GWe, using proven technology reactor like ABWR or others light water reactor with nominal power 1000 MWe. In this case, the rated thermal power for the equilibrium cycles is 3926 MWt, the cycle length is 18 month and overall capacity factor is 87 %. The designs were performed for an 872-fuel bundles ABWR core using GE-11 fuel type in an 9×9 fuel rod arrays with 2 Large Central Water Rods (LCWR. The calculations were divided into two steps; the first is to generate bundle library and the other is to make the thermal and reactivity limits satisfied for the core designs. Toshiba General Electric Bundle lattice Analysis (TGBLA and PANACEA computer codes were used as designs tools. TGBLA is a General Electric proprietary computer code which is used to generate bundle lattice library for fuel designs. PANACEA is General Electric proprietary computer code which is used as thermal hydraulic and neutronic coupled BWR core simulator. This result of core designs describes reactivity and thermal margins i.e.; Maximum Linear Heat Generation rate (MLHGR is lower than 14.4 kW/ft, Minimum Critical Power Ratio (MCPR is upper than 1.25, Hot Excess Reactivity (HOTXS is upper than 1 %Dk at BOC and 0.8 %Dk at 200 MWD/ST and Cold Shutdown Margin Reactivity (CSDM is upper than 1 %Dk. It is concluded that the equilibrium core design using GE-11 fuel bundle type satisfies the core design objectives for the proposed of the firs Indonesia ABWR Nuclear Power Plant. Keywords: The first NPP in Indonesia, ABWR-1000 MWe, and core designs.   Indonesia adalah sebagai negara kepulauan yang laju pertumbuhan industri, energi, penduduk

Determination of PWR core water level using ex-core detectors signals

International Nuclear Information System (INIS)

Bernal, Alvaro; Abarca, Agustin; Miro, Rafael; Verdu, Gumersindo

2013-01-01

The core water level provides relevant neutronic and thermalhydraulic information of the reactor such as power, k eff and cooling ability; in fact, core water level monitoring could be used to predict LOCA and cooling reduction which may deal with core damage. Although different detection equipment is used to monitor several parameters such as the power, core water level monitoring is not an evident task. However, ex-core detectors can measure the fast neutrons leaking the core and several studies demonstrate the existence of a relationship between fast neutron leakage and core water level due to the shielding effect of the water. In addition, new ex-core detectors are being developed, such as silicon carbide semiconductor radiation detectors, monitoring the neutron flux with higher accuracy and in higher temperatures conditions. Therefore, a methodology to determine this relationship has been developed based on a Monte Carlo calculation using MCNP code and applying variance reduction with adjoint functions based on the adjoint flux obtained with the discrete ordinates code TORT. (author)

Core Power Limits For A Lead-Bismuth Natural Circulation Actinide Burner Reactor

Energy Technology Data Exchange (ETDEWEB)

Davis, Cliff Bybee; Kim, D.; Todreas, N. E.; Mujid S. Kazimi

2002-04-01

The Idaho National Engineering and Environmental Laboratory and Massachusetts Institute of Technology are investigating the suitability of lead-bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The design being considered here is a pool type reactor that burns actinides and utilizes natural circulation of the primary coolant, a conventional steam power conversion cycle, and a passive decay heat removal system. Thermal-hydraulic evaluations of the actinide burner reactor were performed to determine allowable core power ratings that maintain cladding temperatures below corrosion-established temperature limits during normal operation and following a loss-of-feedwater transient. An economic evaluation was performed to optimize various design parameters by minimizing capital cost. The transient power limit was initially much more restrictive than the steady-state limit. However, enhancements to the reactor vessel auxiliary cooling system for transient decay heat removal resulted in an increased power limit of 1040 MWt, which was close to the steady-state limit. An economic evaluation was performed to estimate the capital cost of the reactor and its sensitivity to the transient power limit. For the 1040 MWt power level, the capital cost estimate was 49 mills per kWhe based on 1999 dollars.

Contribution of Anticipated Transients Without Scram (ATWS) to core melt at United States nuclear power plants

International Nuclear Information System (INIS)

Giachetti, R.T.

1989-09-01

This report looks at WASH-1400 and several other Probabilistic Risk Assessments (PRAs) and Probabilistic Safety Studies (PSSs) to determine the contribution of Anticipated Transients Without Scram (ATWS) events to the total core melt probability at eight nuclear power plants in the United States. After considering each plant individually, the results are compared from plant to plant to see if any generic conclusions regarding ATWS, or core melt in general, can be made. 8 refs., 34 tabs

Dynamic Voltage-Frequency and Workload Joint Scaling Power Management for Energy Harvesting Multi-Core WSN Node SoC

Directory of Open Access Journals (Sweden)

Xiangyu Li

2017-02-01

Full Text Available This paper proposes a scheduling and power management solution for energy harvesting heterogeneous multi-core WSN node SoC such that the system continues to operate perennially and uses the harvested energy efficiently. The solution consists of a heterogeneous multi-core system oriented task scheduling algorithm and a low-complexity dynamic workload scaling and configuration optimization algorithm suitable for light-weight platforms. Moreover, considering the power consumption of most WSN applications have the characteristic of data dependent behavior, we introduce branches handling mechanism into the solution as well. The experimental result shows that the proposed algorithm can operate in real-time on a lightweight embedded processor (MSP430, and that it can make a system do more valuable works and make more than 99.9% use of the power budget.

Two-dimensional thermal-hydraulic behavior in core in SCTF Core-II cold leg injection tests

International Nuclear Information System (INIS)

Iwamura, Takamichi; Sobajima, Makoto; Okubo, Tsutomu; Ohnuki, Akira; Abe, Yutaka; Adachi, Hiromichi

1985-07-01

Major purpose of the Slab Core Test Program is to investigate the two-dimensional thermal-hydraulic behavior in the core during the reflood phase in a PWR-LOCA. In order to investigate the effects of radial power profile, three cold leg injection tests with different radial power profiles under the same total heating power and core stored energy were performed by using the Slab Core Test Facility (SCTF) Core-II. It was revealed by comparing these three tests that the heat transfer was enhanced in the higher power bundles and degraded in the lower power bundles in the non-uniform radial power profile tests. The turnaround temperature in the high power bundles were evaluated to be reduced by about 40 to 120 K. On the other hand, a two-dimensional flow in the core was also induced by the non-uniform water accumulation in the upper plenum and the quench was delayed resultantly in the bundles corresponding to the peripheral bundles of a PWR. However, the effect of the non-uniform upper plenum water accumulation on the turnaround temperature was small because the effect dominated after the turnaround of the cladding temperature. Selected data from Tests S2-SH1, S2-SH2 and S2-O6 are also presented in this report. Some data from Tests S2-SH1 and S2-SH2 were compared with TRAC post-test calculations performed by the Los Alamos National Laboratory. (author)

Design of the zero power reactor core of Instituto de Energia Atomica, SP, Brazil

International Nuclear Information System (INIS)

Ferreira, Antonio Carlos de Almeida

1974-01-01

The main characteristics of a graphite moderated core of a critical assembly to be installed in the zero power reactor of the Instituto de Energia Atomica have been defined. Several simple geometric configurations have been selected and criticality studies have been made. The necessary quantity of fissile uranium has been calculated. (author)

Temporary core liquid level depression during cold-leg small-break LOCA effect of break size and power level

International Nuclear Information System (INIS)

Koizumi, Y.; Kumamaru, H.; Mimura, Y.; Kukita, Y.; Tasaka, K.

1989-01-01

Cold-leg small break LOCA experiments (0.5-10% break) were conducted at the large scale test facility (LSTF), a volumetrically-scaled (1/48) simulator of a PWR, of the ROSA-IV Program. When a break area was less than 2.5% of the scaled cold-leg flow area, the core liquid level was temporarily further depressed to the bottom elevation of the crossover leg during the loop seal clearing early in the transient only by the manometric pressure balance since no coolant remained in the upper portion of the primary system. When the break size was larger than 5%, the core liquid level was temporarily further depressed lower than the bottom elevation of the crossover leg during the loop seal clearing since coolant remained at the upper portion of the primary system; the steam generator (SG) U-tube upflow side and the SG inlet plenum, due to counter current flow limiting by updrafting steam while the coolant drained. The amount of coolant trapped there was dependent on the vapor velocity (core power); the larger the core power, the lower the minimum core liquid level. The RELAP5/MOD2 code reasonable predicted phenomena observed in the experiments. (orig./DG)

Incorporation of Resonance Upscattering and Intra-Pellet Power Profile in Direct Whole Core Calculation

International Nuclear Information System (INIS)

Lim, Chang Hyun; Jung Yeon Sang; Joo Han Gyu

2012-01-01

It was generally known that the Doppler feedback effect computed by most industrial reactor analysis codes is underestimated than the actual values. Part of the underestimation was attributed to the neglect of the resonance upscattering during the slowing down calculation. On the contrary, the edge peaked power profile noted in burned fuel pins due to more plutonium buildup at the periphery of fuel pellets might lead to smaller power defects than the predicted values obtained with a flat profile. This work is to mitigate these problems with a direct whole core calculation code nTRACER which is capable of handling ringwise depletion as well as incorporating nonuniform power profiles inside a fuel pellet

Effects of space-dependent cross sections on core physics parameters for compact fast spectrum space power reactors

International Nuclear Information System (INIS)

Lell, R.M.; Hanan, N.A.

1987-01-01

Effects of multigroup neutron cross section generation procedures on core physics parameters for compact fast spectrum reactors have been examined. Homogeneous and space-dependent multigroup cross section sets were generated in 11 and 27 groups for a representative fast reactor core. These cross sections were used to compute various reactor physics parameters for the reference core. Coarse group structure and neglect of space-dependence in the generation procedure resulted in inaccurate computations of reactor flux and power distributions and in significant errors regarding estimates of core reactivity and control system worth. Delayed neutron fraction was insensitive to cross section treatment, and computed reactivity coefficients were only slightly sensitive. However, neutron lifetime was found to be very sensitive to cross section treatment. Deficiencies in multigroup cross sections are reflected in core nuclear design and, consequently, in system mechanical design

Real-Time Power-Efficient Integration of Multi-Sensor Occupancy Grid on Many-Core

OpenAIRE

Rakotovao , Tiana; Mottin , Julien; Puschini , Diego; Laugier , Christian

2015-01-01

International audience; Safe Autonomous Vehicles (AVs) will emerge when comprehensive perception systems will be successfully integrated into vehicles. Advanced perception algorithms, estimating the position and speed of every obstacle in the environment by using data fusion from multiple sensors, were developed for AV prototypes. Computational requirements of such application prevent their integration into AVs on current low-power embedded hardware. However, recent emerging many-core archite...

Powered bone marrow biopsy procedures produce larger core specimens, with less pain, in less time than with standard manual devices

Directory of Open Access Journals (Sweden)

Larry J. Miller

2011-07-01

Full Text Available Bone marrow sampling remains essential in the evaluation of hematopoietic and many non-hematopoietic disorders. One common limitation to these procedures is the discomfort experienced by patients. To address whether a Powered biopsy system could reduce discomfort while providing equivalent or better results, we performed a randomized trial in adult volunteers. Twenty-six subjects underwent bilateral biopsies with each device. Core samples were obtained in 66.7% of Manual insertions; 100% of Powered insertions (P=0.002. Initial mean biopsy core lengths were 11.1±4.5 mm for the Manual device; 17.0±6.8 mm for the Powered device (P<0.005. Pathology assessment for the Manual device showed a mean length of 6.1±5.6 mm, width of 1.0±0.7 mm, and volume of 11.0±10.8 mm3. Powered device measurements were mean length of 15.3±6.1 mm, width of 2.0±0.3 mm, and volume of 49.1±21.5 mm3 (P<0.001. The mean time to core ejection was 86 seconds for Manual device; 47 seconds for the Powered device (P<0.001. The mean second look overall pain score was 33.3 for the Manual device; 20.9 for the Powered (P=0.039. We conclude that the Powered biopsy device produces superior sized specimens, with less overall pain, in less time.

The thermal-mechanical behavior of fuel pins during power's maneuvering regime at stationary core loading on 2nd unit of KHNPP

International Nuclear Information System (INIS)

Ieremenko, M.; Ovdiyenko, Y.; Khalimonchuk, V.

2007-01-01

Results of thermal-mechanical behaviour of fuel pins during daily power's maneuvering regime that were proposed for second unit of Khmelnitsky NPP are presented. Calculations were performed for campaign's moments 100 and 160 fpd and for different type of regulation. Additionally calculations were performed for campaign 7. It is the design variant of the campaign and reactor core contains the high burnt fuel. Calculations of macro-core parameters (Kq, Kv) was performed by spatial computer code DYN3D. Calculations of micro-core parameters (fuel pin power) was performed by computer code DERAB. Calculations of thermal-mechanical behaviour of fuel pins was performed by computer code TRANSURANUS (Authors)

Special power supply and control system for the gas-cooled fast reactor-core flow test loop

International Nuclear Information System (INIS)

Hudson, T.L.

1981-09-01

The test bundle in the Gas-Cooled Fast Reactor-Core Flow Test Loop (GCFR-CFTL) requires a source of electrical power that can be controlled accurately and reliably over a wide range of steady-state and transient power levels and skewed power distributions to simulate GCFR operating conditions. Both ac and dc power systems were studied, and only those employing silicon-controlled rectifiers (SCRs) could meet the requirements. This report summarizes the studies, tests, evaluations, and development work leading to the selection. it also presents the design, procurement, testing, and evaluation of the first 500-kVa LMPL supply. The results show that the LMPL can control 60-Hz sine wave power from 200 W to 500 kVA

Fast core prediction simulator for load follow control

International Nuclear Information System (INIS)

Yim, Man Sung; Lee, Sang Hoon; Lee, Un Chul

1990-01-01

An operator-assisting system for the reactor core control under power changing operating condition was developed. The system is consisted of core simulator routine and Xenon and Iodine initial condition generation routine. The initial condition generation routine, without exactly knowing the core status, is capable of providing accurate number densities and axial offset conditions of Xenon and Iodine after several hours of predictor- corrector calculations using the plant instrumentation signals of power level and power axial offset. The core simulator routine, even with the two node core model, gives equivalently accurate results as the one-dimensional model for the core behaviour simulation under power changing condition and can provide proper control strategies for load follow operation. The core simulator can also be used by the operator to develop remedial actions to restore the distorted power distribution by using its prediction capability

Reactor core for LMFBR type reactors

International Nuclear Information System (INIS)

Masumi, Ryoji; Azekura, Kazuo; Kurihara, Kunitoshi; Bando, Masaru; Watari, Yoshio.

1987-01-01

Purpose: To reduce the power distribution fluctuations and obtain flat and stable power distribution throughout the operation period in an LMFBR type reactor. Constitution: In the inner reactor core region and the outer reactor core region surrounding the same, the thickness of the inner region is made smaller than the axial height of the reactor core region and the radial width thereof is made smaller than that of the reactor core region and the volume thereof is made to 30 - 50 % for the reactor core region. Further, the amount of the fuel material per unit volume in the inner region is made to 70 - 90 % of that in the outer region. The difference in the neutron infinite multiplication factor between the inner region and the outer region is substantially constant irrespective of the burnup degree and the power distribution fluctuation can be reduced to about 2/3, by which the effect of thermal striping to the reactor core upper mechanisms can be moderated. Further, the maximum linear power during operation can be reduced by 3 %, by which the thermal margin in the reactor core is increased and the reactor core fuels can be saved by 3 %. (Kamimura, M.)

Self-powered in-core detectors of cobalt type

International Nuclear Information System (INIS)

Jonsson, Georg

1975-01-01

Testing and development of self-powered neutron detectors with a cobalt emitter is described. Long term irradiation at 400 deg C is expected to indicate insulation quality, change in calibration and 60 Co build-up. Dynamic tests to investigate possible transient effects due to temperature changes are being performed on a number of detectors up to about 600 deg C. A long term irradiation at low temperature has been terminated after 4.5 years. On completion, neutron dose was estimated to be 5.6 x 10 21 nvt and the 60 Co background was 9.3 % of the full flux signal. A recently introduced long term test is expected to provide data on instability effects due to 61 Co. For a BWR in-core detector installation, the main advantage of cobalt detectors, apart from the small size, appears to be long life. Development work is being done on detectors with vanadium-cobalt emitters, electronic separation of fast and delayed signals and reduction of gamma sensitivity. (O.T.)

Two-dimensional thermal-hydraulic behavior in core in SCTF Core-II forced feed reflood tests

International Nuclear Information System (INIS)

Iwamura, Takamichi; Sobajima, Makoto; Okubo, Tsutomu; Ohnuki, Akira; Abe, Yutaka; Adachi, Hiromichi

1987-01-01

Major purpose of the Slab Core Test Program is to investigate the two-dimensional thermal-hydraulic behavior in the core during the reflood phase of a PWR-LOCA. It was revealed in the previous Slab Core Test Facility (SCTF) Core-II test results that the heat transfer was enhanced in the higher power bundles and degraded in the lower power bundles in the non-uniform radial power profile tests. In order to separately evaluate the effect of the radial power (Q) distribution itself and the effect of the radial temperature (T) distribution, four tests were performed with steep Q and T, flat Q and T, steep Q and flat T, and flat Q and steep T. Based on the test results, it was concluded that the radial temperature distribution which accompanied the radial power distribution was the dominant factor of the two-dimensional thermal-hydraulic behavior in the core during the initial period. Selected data from these four tests are also presented in this report. Some data from Test S2-12 (steep Q, T) were compared with TRAC post-test calculations performed by the Los Alamos National Laboratory. (author)

Minimum critical power ratio control device for nuclear power plants

International Nuclear Information System (INIS)

Kurosawa, Tsuneo.

1991-01-01

Reactor core flowrate is determined by comparing a minimum critical power ratio calculated based on the status amount of a nuclear power plant and a control value for the minimum critical power ratio that depends on the reactor core flowrate. Further, the minimum critical power ratio and a control value for the minimum critical power ratio that depends on the reactor thermal power are compared to set a reactor thermal power converted to a reactor core flowrate. Deviation between the thus determined reactor core flowrate and the present reactor core flowrate is calculated. When the obtained deviation is lower than a rated value, a reactor core flowrate set signal is generated to a reactor flowrate control means, to control the reactor power by a recycling flowrate control system of the reactor. On the other hand, when the deviation exceeds the determined value, the reactor core flowrate set signal is converted into a reactor thermal power, to control the position of control rods and control the reactor power. Then, monitor and control can be conducted safely and automatically without depending on operator's individual ability over the entire operation range corresponding to load following operation. (N.H.)

Error Analysis of High Frequency Core Loss Measurement for Low-Permeability Low-Loss Magnetic Cores

DEFF Research Database (Denmark)

Niroumand, Farideh Javidi; Nymand, Morten

2016-01-01

in magnetic cores is B-H loop measurement where two windings are placed on the core under test. However, this method is highly vulnerable to phase shift error, especially for low-permeability, low-loss cores. Due to soft saturation and very low core loss, low-permeability low-loss magnetic cores are favorable...... in many of the high-efficiency high power-density power converters. Magnetic powder cores, among the low-permeability low-loss cores, are very attractive since they possess lower magnetic losses in compared to gapped ferrites. This paper presents an analytical study of the phase shift error in the core...... loss measuring of low-permeability, low-loss magnetic cores. Furthermore, the susceptibility of this measurement approach has been analytically investigated under different excitations. It has been shown that this method, under square-wave excitation, is more accurate compared to sinusoidal excitation...

Review of advanced core designs for LMFBRs

International Nuclear Information System (INIS)

Yoshida, Kazuo

1986-01-01

It is a matter of great importance for the development of LMFBR to reduce its power cost to the level of the other power generating means. For this purpose, some ideas that use advanced core concepts to reduce LMFBR's power cost by improving its fuel cycle economics have recently been proposed. In this report, two hopeful ideas that use advanced core concepts: (1) Ultra Long Life Core (ULLC) - non-refueling over LMFBR power plant life; (2) Integral Fast Reactor (IFR) concept - metal fueled core and pyrometallurgical reprocessing; are picked up and their economical effect and technical probrems are investigated. (author)

Neutronic studies of the long life core concept: Part 1, Design and performance of 1000 MWe uranium oxide fueled low power density LMR cores

International Nuclear Information System (INIS)

Orechwa, Y.

1987-04-01

The parametric behavior of some key neutronic performance parameters for low power density LMR cores fueled with uranium oxide is investigated. The results are compared to reference homogeneous and heterogeneous cores with normal fuel management and Pu fueling. It can be concluded that with respect to minimizing the initial fissile mass and thereby economizing on the inventory costs and carrying charges, the superior neutron economy of the LMR fuel cycle is best exploited through normal fuel management with Pu recycling. In the once-through mode the LMR fuel cycle has disadvantages due to a higher fissile inventory and is not competitive with the LWR fuel cycle

Computer programs for the in-core fuel management of power reactors

International Nuclear Information System (INIS)

1981-08-01

This document gives a survey of the presently tested and used computer programs applicable to the in-core fuel management of light and heavy water moderated nuclear power reactors. Each computer program is described (provided that enough information was supplied) such that the nature of the physical problem solved and the basic mathematical or calculational approach are evident. In addition, further information regarding computer requirements, up-to-date applications and experiences and specific details concerning implementation, staff requirements, etc., are provided. Program procurement conditions, possible program implementation assistance and commercial conditions (where applicable) are given. (author)

Burnup Estimation of Rhodium Self-Powered Neutron Detector Emitter in VVER Reactor Core Using Monte Carlo Simulations

OpenAIRE

Khrutchinsky, А. А.; Kuten, S. A.; Babichev, L. F.

2011-01-01

Estimation of burn-up in a rhodium-103 emitter of self-powered neutron detector in VVER-1000 reactor core has been performed using Monte Carlo simulations within approximation of a constant neutron flux.

C.N. Cofrentes power up-rate up to 110 %. A challenge for cycle 14 core design

International Nuclear Information System (INIS)

Gomez Bernal, M.I.; Lopez Carbonell, M.T.; Garcia Delgado, L.

2001-01-01

C.N.Cofrentes is a GE design BWR reactor with 624 bundles in the core, a rated power of 2894 MWt and it is currently operating Cycle 13 at 104.2 % power. Commercial operation started in 1984 with 12-month cycles at rated power. Both cycle length and thermal power have been increased since then. Power has been up-rated in two steps, first at 102 % in Cycle 4 and later in Cycle 11 at 104.2%. Cycle length has been extended from the original 12-month to the currently 18-month cycles. Next cycle, Cycle 14, will be an 18-month cycle operating at 110 % power. This goal is a challenge for the in-house nuclear design team. Start up for Cycle 14 is planned for the first quarter of 2002. (author)

3D thermal-hydraulic analysis on core of PWR nuclear power station

International Nuclear Information System (INIS)

Yao Zhaohui; Wang Xuefang; Shen Mengyu

1997-01-01

Thermal hydraulic analysis of core is of great importance in reactor safety analysis. A computer code, thermal hydraulic analysis porous medium analysis (THAPMA), has been developed to simulate the flow and heat transfer characteristics of reactor components. It has been proved reliable by several numerical tests. In the THAPMA code, a new difference scheme and solution method have been studied in developing the computer software. For the difference scheme, a second order accurate, high resolution scheme, called WSUC scheme, has been proposed. This scheme is total variation bounded and unconditionally stable in convective numeral stability. Numerical tests show that the WSUC is better in accuracy and resolution than the 1-st order upwind, 2-nd order upwind, SOUCUP by Zhu and Rodi. In solution method, a modified PISO algorithm is used, which is not only simpler but also more accurate and more rapid in convergence than the original PISO algorithm. Moreover, the modified PISO algorithm can effectively solve steady and transient state problem. Besides, with the THAPMA code, the flow and heat transfer phenomena in reactor core have been numerically simulated in the light of the design condition of Qinshan PWR nuclear power station (the second-term project). The simulation results supply a theoretical basis for the core design

3D Core Model for simulation of nuclear power plants: Simulation requirements, model features, and validation

International Nuclear Information System (INIS)

Zerbino, H.

1999-01-01

In 1994-1996, Thomson Training and Simulation (TT and S) earned out the D50 Project, which involved the design and construction of optimized replica simulators for one Dutch and three German Nuclear Power Plants. It was recognized early on that the faithful reproduction of the Siemens reactor control and protection systems would impose extremely stringent demands on the simulation models, particularly the Core physics and the RCS thermohydraulics. The quality of the models, and their thorough validation, were thus essential. The present paper describes the main features of the fully 3D Core model implemented by TT and S, and its extensive validation campaign, which was defined in extremely positive collaboration with the Customer and the Core Data suppliers. (author)

The domestic development of rhodium self-powered detector used in the core of Qinshan third nuclear power plant

International Nuclear Information System (INIS)

Xiong Weihua; Zhang Zhenhua; Yu Yijun; Zhang Yun; Wu Jun; Deng Peng

2009-01-01

This article introduced Qinshan third nuclear power plant's Vanadium detector's principle of work, the domestically development's earlier period preparation, the craft processing process, the domestically sample's experiment as well as the sample in core demonstration test. Elaborated process of manufacture's quality control request and the essential craft, and the factory manufacture experiment situation, and to the installation and trial run process, the modification factor and the test result has carried on the introduction and the analysis. (authors)

Verification Results of Safety-grade Optical Modem for Core Protection Calculator (CPC) in Korea Standard Nuclear Power Plant (KSNP)

Energy Technology Data Exchange (ETDEWEB)

Kim, Jangyeol; Son, Kwangseop; Lee, Youngjun; Cheon, Sewoo; Cha, Kyoungho; Lee, Jangsoo; Kwon, Keechoon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-05-15

We confirmed that the coverage criteria for a safety-grade optical modem of a Core Protection Calculator is satisfactory using a traceability analysis matrix between high-level requirements and lower-level system test case data set. This paper describes the test environment, test components and items, a traceability analysis, and system tests as a result of system verification and validation based on Software Requirement Specifications (SRS) for a safety-grade optical modem of a Core Protection Calculator (CPC) in a Korea Standard Nuclear Power Plant (KSNP), and Software Design Specifications (SDS) for a safety-grade optical modem of a CPC in a KSNP. All tests were performed according to the test plan and test procedures. Functional testing, performance testing, event testing, and scenario based testing for a safety-grade optical modem of a Core Protection Calculator in a Korea Standard Nuclear Power Plant as a thirty-party verifier were successfully performed.

Analysis the Response Function of the HTR Ex-core Neutron Detectors in Different Core Status

International Nuclear Information System (INIS)

Fan Kai; Li Fu; Zhou Xuhua

2014-01-01

Modular high temperature gas cooled reactor HTR-PM demonstration plant, designed by INET, Tsinghua University, is being built in Shidao Bay, Shandong province, China. HTR-PM adopts pebble bed concept. The harmonic synthesis method has been developed to reconstruct the power distributions on HTR-PM. The method based on the assumption that the neutron detector readings are mainly determined by the status of the core through the power distribution, and the response functions changed little when the status of the core changed. To verify the assumption, the influence factors to the ex-core neutron detectors are calculated in this paper, including the control rod position and the temperature of the core. The results shows that when the status of the core changed, the power distribution changed more remarkable than the response function, but the detector readings could change about 5% because of the response function changing. (author)

Results from core-edge experiments in high Power, high performance plasmas on DIII-D

Directory of Open Access Journals (Sweden)

T.W. Petrie

2017-08-01

Full Text Available Significant challenges to reducing divertor heat flux in highly powered near-double null divertor (DND hybrid plasmas, while still maintaining both high performance metrics and low enough density for application of RF heating, are identified. For these DNDs on DIII-D, the scaling of the peak heat flux at the outer target (q⊥P ∝ [PSOL x IP] 0.92 for PSOL= 8−19MW and IP= 1.0–1.4MA, and is consistent with standard ITPA scaling for single-null H-mode plasmas. Two divertor heat flux reduction methods were tested. First, applying the puff-and-pump radiating divertor to DIII-D plasmas may be problematical at high power and H98 (≥ 1.5 due to improvement in confinement time with deuterium gas puffing which can lead to unacceptably high core density under certain conditions. Second, q⊥P for these high performance DNDs was reduced by ≈35% when an open divertor is closed on the common flux side of the outer divertor target (“semi-slot” but also that heating near the slot opening is a significant source for impurity contamination of the core.

ETRR-2 in-core fuel management strategy

International Nuclear Information System (INIS)

Khalil, M.Y.; Amin, Esmat; Belal, M.G.

2005-01-01

The Egypt second research reactor has many irradiation channels, beam tubes and irradiation boxes, inside and outside the reactor core. The core reload configuration has great effect on the core performance and fluxes in the irradiation channels. This paper deals with the design and safety analysis that were performed for the determination of ETRR2 in-core fuel management strategy which fulfills neutronic design criteria, safety reactor operation, utility optimization and achieve the overall fuel management criteria. The core is divided into 8 zones, in order to obtain the minimum and adjacent fuel movement scheme that is recommended from the operational point of view. Then a search for the initial core using backward iteration, one get different initial cores, one initial core would assume the equilibrium core after 250 full power days of operation, while the other assumes equilibrium after 199 full power days, and shows a better performance of power peaking factor. (author)

Neutronic characteristics of FLWR in the transition phase changing from high conversion core to breeder core

International Nuclear Information System (INIS)

Akie, Hiroshi; Nakano, Yoshihiro; Okubo, Tsutomu

2009-01-01

Innovative Water Reactor for Flexible Fuel Cycle (FLWR) is a low moderation type boiling water reactor which can realize plutonium multiple recycling and breeding. For the introduction stage of FLWR, a high conversion (HC) type FLWR is proposed to keep technical continuity from current light water reactors. The HC core of FLWR has a less tight fuel lattice with lower coolant void fraction than the breeder (BR) type core. The HC type FLWR core is to be shifted to the BR core by only replacing the fuel assemblies of the same outer shape and size in the same reactor system. In the HC to BR transition phase of FLWR, there exist both types of fuel assemblies in the same core configuration. In the HC assembly, neutron spectrum is softer than in the BR assembly, and the axial fuel and blanket arrangement is different from the BR assembly. Due to these differences, there might appear a power peaking in the adjacent region between HC and BR assemblies. The power distribution in the HC + BR assemblies mixed core configuration is studied by performing assembly calculations and core calculations on a few assemblies local geometry and the whole core geometry. As a result, although a power peaking can be locally very large in the HC and BR assemblies adjacent regions, such local power peakings are shown to be effectively reduced by considering a rod-wise fuel enrichment distribution. In the whole core calculation, it seems possible to optimize the fuel assembly loading and shuffling pattern to avoid large power level mismatch between the assemblies. It is expected that FLWR can be shifted from HC type to BR type without major neutronic difficulties. (author)

EXPERIMENTAL DETERMINATION OF LONGITUDINAL COMPONENT OF MAGNETIC FLUX IN FERROMAGNETIC WIRE OF SINGLE-CORE POWER CABLE ARMOUR

Directory of Open Access Journals (Sweden)

I.A. Kostiukov

2014-12-01

Full Text Available A problem of determination of effective longitudinal magnetic permeability of single core power cable armour is defined. A technique for experimental determination of longitudinal component of magnetic flux in armour spiral ferromagnetic wire is proposed.

Core supervision methods and future improvements of the core master/presto system at KKB

International Nuclear Information System (INIS)

Lundberg, S.; Wenisch, J.; Teeffelen, W.V.

2000-01-01

Kernkraftwerk Brunsbuettel (KKB) is a KWU 806 MW e BWR located at the lower river Elbe, in Germany. The reactor has been in operation since 1976 and is now operating in its 14. cycle. The core supervision at KKB is performed with the ABB CORE MASTER system. This system mainly contains the 3-D simulator PRESTO supplied by Studsvik Scandpower A/S. The core supervision is performed by periodic PRESTO 3-D evaluations of the reactor operation state. The power distribution calculated by PRESTO is adapted with the ABB UPDAT program using the on-line LPRM readings. The thermal margins are based on this adapted power distribution. Related to core supervision, the function of the PRESTO/UPDAT codes is presented. The UPDAT method is working well and is capable of reproducing the true core power distribution. The quality of the 3-D calculation is, however, an important ingredient of the quality of the adapted power distribution. The adaptation method as such is also important for this quality. The data quality of this system during steady state and off-rate states (reactor manoeuvres) are discussed by presenting comparisons between PRESTO and UPDAT thermal margin utilisation from Cycle 13. Recently analysed asymmetries in the UPDAT evaluated MCPR values are also presented and discussed. Improvements in the core supervision such as the introduction of advanced modern nodal methods (PRESTO-2) are presented and an alternative core supervision philosophy is discussed. An ongoing project with the goal to update the data and result presentation interface (GUI) is also presented. (authors)

ORIGEN2.1 Cycle Specific Calculation of Krsko Nuclear Power Plant Decay Heat and Core Inventory

International Nuclear Information System (INIS)

Vukovic, J.; Grgic, D.; Konjarek, D.

2010-01-01

This paper presents ORIGEN2.1 computer code calculation of Krsko Nuclear Power Plant core for Cycle 24. The isotopic inventory, core activity and decay heat are calculated in one run for the entire core using explicit depletion and decay of each fuel assembly. Separate pre-ori application which was developed is utilized to prepare corresponding ORIGEN2.1 inputs. This application uses information on core loading pattern to determine fuel assembly specific depletion history using 3D burnup which is obtained from related PARCS computer code calculation. That way both detailed single assembly calculations as well as whole core inventory calculations are possible. Because of the immense output of the ORIGEN2.1, another application called post-ori is used to retrieve and plot any calculated property on the basis of nuclide, element, summary isotope or group of elements for activation products, actinides and fission products segments. As one additional possibility, with the post-ori application it is able to calculate radiotoxicity from calculated ORIGEN2.1 inventory. The results which are obtained using the calculation model of ORIGEN2.1 computer code are successfully compared against corresponding ORIGEN-S computer code results.(author).

Compact Reversed-Field Pinch Reactors (CRFPR): fusion-power-core integration study

International Nuclear Information System (INIS)

Copenhaver, C.; Krakowski, R.A.; Schnurr, N.M.

1985-08-01

Using detailed two-dimensional neutronics studies based on the results of a previous framework study (LA-10200-MS), the fusion-power-core (FPC) integration, maintenance, and radio-activity/afterheat control are examined for the Compact Reversed-Field Pinch Reactor (CRFPR). While maintaining as a base case the nominal 20-MW/m 2 neutron first-wall loading design, CRFPR(20), the cost and technology impact of lower-wall-loading designs are also examined. The additional detail developed as part of this follow-on study also allows the cost estimates to be refined. The cost impact of multiplexing lower-wall-loading FPCs into a approx. 1000-MWe(net) plant is also examined. The CRFPR(20) design remains based on a PbLi-cooled FPC with pressurized-water used as a coolant for first-wall, pumped-limiter, and structural-shield systems. Single-piece FPC maintenance of this steady-state power plant is envisaged and evaluated on the basis of a preliminary layout of the reactor building. This follow-on study also develops the groundwork for assessing the feasibility and impact of impurity/ash control by magnetic divertors as an alternative to previously considered pumped-limiter systems. Lastly, directions for future, more-detailed power-plant designs based on the Reversed-Field Pinch are suggested

KSI's Cross Insulated Core Transformer Technology

International Nuclear Information System (INIS)

Uhmeyer, Uwe

2009-01-01

Cross Insulated Core Transformer (CCT) technology improves on Insulated Core Transformer (ICT) implementations. ICT systems are widely used in very high voltage, high power, power supply systems. In an ICT transformer ferrite core sections are insulated from their neighboring ferrite cores. Flux leakage is present at each of these insulated gaps. The flux loss is raised to the power of stages in the ICT design causing output voltage efficiency to taper off with increasing stages. KSI's CCT technology utilizes a patented technique to compensate the flux loss at each stage of an ICT system. Design equations to calculate the flux compensation capacitor value are presented. CCT provides corona free operation of the HV stack. KSI's CCT based High Voltage power supply systems offer high efficiency operation, high frequency switching, low stored energy and smaller size over comparable ICT systems.

Design and test of the borosilicate glass burnable poison rod for Qinshan nuclear power plant core

International Nuclear Information System (INIS)

Huang Jinhua; Sun Hanhong

1988-08-01

Material for the burnable poison of Qinshan Nuclear Power Plant core is GG-17 borosilicate glass. The chemical composition and physico-chemical properties of GG-17 is very close to Pyrex-7740 glass used by Westinghouse. It is expected from the results of the experiments that the borosilicate glass burnable poison rod can be successfully used in Qinshan Nuclear Power Plant due to good physical, mechanical, corrosion-resistant and irradiaton properties for both GG-17 glass and cold-worked stainless steel cladding. Change of material for burnable poison from boron-bearing stainless steel to borosilicate glass will bring about much more economic benefit to Qinshan Naclear Power Plant

Improved core protection calculator system algorithm

International Nuclear Information System (INIS)

Yoon, Tae Young; Park, Young Ho; In, Wang Kee; Bae, Jong Sik; Baeg, Seung Yeob

2009-01-01

Core Protection Calculator System (CPCS) is a digitized core protection system which provides core protection functions based on two reactor core operation parameters, Departure from Nucleate Boiling Ratio (DNBR) and Local Power Density (LPD). It generates a reactor trip signal when the core condition exceeds the DNBR or LPD design limit. It consists of four independent channels which adapted a two out of four trip logic. CPCS algorithm improvement for the newly designed core protection calculator system, RCOPS (Reactor COre Protection System), is described in this paper. New features include the improvement of DNBR algorithm for thermal margin, the addition of pre trip alarm generation for auxiliary trip function, VOPT (Variable Over Power Trip) prevention during RPCS (Reactor Power Cutback System) actuation and the improvement of CEA (Control Element Assembly) signal checking algorithm. To verify the improved CPCS algorithm, CPCS algorithm verification tests, 'Module Test' and 'Unit Test', would be performed on RCOPS single channel facility. It is expected that the improved CPCS algorithm will increase DNBR margin and enhance the plant availability by reducing unnecessary reactor trips

Evaluation of BEACON-COLSS Core Monitoring System Benefits

International Nuclear Information System (INIS)

Kim, Joon Sung; Park, Young Ho; Morita, Toshio; Book, Michael A.

2005-01-01

In Korean Standard Nuclear Power Plant COLSS (Core Operating Limit Supervisory System) is used to monitor the DNBR Power Operating Limit (DNBRPOL) and Linear Heat Rate POL (KWPFPOL). Westinghouse and KNFC have developed an upgraded core monitoring system by combining the BEACON TM core monitoring system 1 (Best Estimate Analyzer for Core Operation . Nuclear) and COLSS into an integrated product that is called BEACON-COLSS. BEACON-COLSS generates the 3-D power distribution corrected by the in-core detectors measurements. The 3-D core power distribution methodology in BEACON-COLSS is significantly better than the synthesis methodology in COLSS. BEACONCOLSS uses the CETOP-D 2 thermal hydraulic code instead of CETOP-1. CETOP-D is a multi-channel thermal hydraulics code that will provide more accurate DNBR calculations than the DNBR calculators currently used in COLSS

Development of an inconel self powered neutron detector for in-core reactor monitoring

Science.gov (United States)

Alex, M.; Ghodgaonkar, M. D.

2007-04-01

The paper describes the development and testing of an Inconel600 (2 mm diameter×21 cm long) self-powered neutron detector for in-core neutron monitoring. The detector has 3.5 mm overall diameter and 22 cm length and is integrally coupled to a 12 m long mineral insulated cable. The performance of the detector was compared with cobalt and platinum detectors of similar dimensions. Gamma sensitivity measurements performed at the 60Co irradiation facility in 14 MR/h gamma field showed values of -4.4×10 -18 A/R/h/cm (-9.3×10 -24 A/ γ/cm 2-s/cm), -5.2×10 -18 A/R/h/cm (-1.133×10 -23 A/ γ/cm 2-s/cm) and 34×10 -18 A/R/h/cm (7.14×10 -23 A/ γ/cm 2-s/cm) for the Inconel, Co and Pt detectors, respectively. The detectors together with a miniature gamma ion chamber and fission chamber were tested in the in-core Apsara Swimming Pool type reactor. The ion chambers were used to estimate the neutron and gamma fields. With an effective neutron cross-section of 4b, the Inconel detector has a total sensitivity of 6×10 -23 A/nv/cm while the corresponding sensitivities for the platinum and cobalt detectors were 1.69×10 -22 and 2.64×10 -22 A/nv/cm. The linearity of the detector responses at power levels ranging from 100 to 200 kW was within ±5%. The response of the detectors to reactor scram showed that the prompt response of the Inconel detector was 0.95 while it was 0.7 and 0.95 for the platinum and cobalt self-powered detectors, respectively. The detector was also installed in the horizontal flux unit of 540 MW Pressurised Heavy Water Reactor (PHWR). The neutron flux at the detector location was calculated by Triveni code. The detector response was measured from 0.02% to 0.07% of full power and showed good correlation between power level and detector signals. Long-term tests and the dynamic response of the detector to shut down in PHWR are in progress.

Development of an inconel self powered neutron detector for in-core reactor monitoring

International Nuclear Information System (INIS)

Alex, M.; Ghodgaonkar, M.D.

2007-01-01

The paper describes the development and testing of an Inconel600 (2 mm diameterx21 cm long) self-powered neutron detector for in-core neutron monitoring. The detector has 3.5 mm overall diameter and 22 cm length and is integrally coupled to a 12 m long mineral insulated cable. The performance of the detector was compared with cobalt and platinum detectors of similar dimensions. Gamma sensitivity measurements performed at the 60 Co irradiation facility in 14 MR/h gamma field showed values of -4.4x10 -18 A/R/h/cm (-9.3x10 -24 A/γ/cm 2 -s/cm), -5.2x10 -18 A/R/h/cm (-1.133x10 -23 A/γ/cm 2 -s/cm) and 34x10 -18 A/R/h/cm (7.14x10 -23 A/γ/cm 2 -s/cm) for the Inconel, Co and Pt detectors, respectively. The detectors together with a miniature gamma ion chamber and fission chamber were tested in the in-core Apsara Swimming Pool type reactor. The ion chambers were used to estimate the neutron and gamma fields. With an effective neutron cross-section of 4b, the Inconel detector has a total sensitivity of 6x10 -23 A/nv/cm while the corresponding sensitivities for the platinum and cobalt detectors were 1.69x10 -22 and 2.64x10 -22 A/nv/cm. The linearity of the detector responses at power levels ranging from 100 to 200 kW was within ±5%. The response of the detectors to reactor scram showed that the prompt response of the Inconel detector was 0.95 while it was 0.7 and 0.95 for the platinum and cobalt self-powered detectors, respectively. The detector was also installed in the horizontal flux unit of 540 MW Pressurised Heavy Water Reactor (PHWR). The neutron flux at the detector location was calculated by Triveni code. The detector response was measured from 0.02% to 0.07% of full power and showed good correlation between power level and detector signals. Long-term tests and the dynamic response of the detector to shut down in PHWR are in progress

Neutronic evolution of SENA reactor during the first and second cycles. Comparison between the experimental power distributions obtained from the in-core instrumentation evaluation code CIRCE and the theoretical power values computed with the two-dimensional diffusion-evolution code EVOE

International Nuclear Information System (INIS)

Andrieux, Chantal

1976-03-01

The neutronic evolution of the reacteur Sena during the first and second cycles is presented. The experimental power distributions, obtained from the in-core instrumentation evaluation code CIRCE are compared with the theoretical powers calculated with the two-dimensional diffusion-evolution code EVOE. The CIRCE code allows: the study of the evolution of the principal parameters of the core, the comparison of the results of measured and theoretical estimates. Therefore this study has a great interest for the knowledge of the neutronic evolution of the core, as well as the validation of the refinement of theoretical estimation methods. The core calculation methods and requisite data for the evaluation of the measurements are presented after a brief description of the SENA core and its inner instrumentation. The principle of the in-core instrumentation evaluation code CIRCE, and calculation of the experimental power distributions and nuclear core parameters are then exposed. The results of the evaluation are discussed, with a comparison of the theoretical and experimental results. Taking account of the approximations used, these results, as far as the first and second cycles at SENA are concerned, are satisfactory, the deviations between theoretical and experimental power distributions being lower than 3% at the middle of the reactor and 9% at the periphery [fr

Studies on WWER core diagnostics

International Nuclear Information System (INIS)

Lunin, G.L.; Mitin, V.I.; Bulavin, V.V.

1987-01-01

The reliability and safety of nuclear power plants have decisive meaning under the situation that nuclear power generation steadily increases, and among various measures aiming at ensuring the reliability and safety in the operation of nuclear power plants, the countermeasures for protecting reactor core, main process equipment and high pressure circuits from damage have the important role, and the monitoring of condition and the organization of forecast, which are carried out continuously or periodically during the operation of nuclear power stations using the diagnostic expert system specially developed for the purpose, are included in them. Such monitoring enables the early detection of mechanical damage, increase of vibration, defects caused during operation and so on in reactor cores and primary and secondary circuits, and the continuous watching of defect developments. Also boiling in a core is detected, the place of abnormality occurrence is identified, and the intensity and characteristics of boiling are determined, thus the occurrence of dangerous condition is prevented. The developments of an in-core monitoring system and noise diagnostic systems are reported. (Kako, I.)

Characteristics of self-powered neutron detectors used in power reactors

International Nuclear Information System (INIS)

Todt, W.H.

1997-01-01

Self-Powered Neutron Detectors have been used effectively as in-core flux monitors for over twenty-five years in nuclear power reactors world-wide. The basic properties of these radiation sensors are described including their nuclear, electrical and mechanical characteristics. Recommendations are given for the proper choice of the self-powered detector emitter to provide the proper response time and radiation sensitivity desired for use in an effective in-core radiation monitoring system. Examples are shown of specific self-powered detector designs which are being effectively used in in-core instrumentation systems for pressurised water, heavy water and graphite moderated light water reactors. Examples are also shown of the mechanical configurations of in-core assemblies of self-powered detectors combined with in-core thermocouples presently used in pressurised water and heavy water reactors worldwide. This paper is a summary of a new IEC standard to be issued in 1996 describing the characteristics and test methods of self-powered detectors used in nuclear power reactors. (author)

Characteristics of self-powered neutron detectors used in power reactors

International Nuclear Information System (INIS)

Todt, William H. Sr.

1998-01-01

Self-powered neutron detectors have been used effectively as in-core flux monitors for over twenty-five years in nuclear power reactors worldwide. This paper describes the basic properties of these radiation sensors including their nuclear, electrical and mechanical characteristics. Recommendations are given for the proper choice of the self-powered detector emitter to provide the proper response time and radiation sensitivity desired for use in an effective in-core radiation monitoring system. Examples are shown of specific self-powered detector designs, which are being effectively, used in in-core instrumentation systems for pressurized water, heavy water and graphite moderated light water reactors. Also examples are shown of the mechanical configurations of in-core assemblies of self-powered detectors combined with in-core thermocouples presently used in pressurized water and heavy water reactors worldwide. (author)

BWR type reactor core

International Nuclear Information System (INIS)

Tatemichi, Shin-ichiro.

1981-01-01

Purpose: To eliminate the variation in the power distribution of a BWR type reactor core in the axial direction even if the flow rate is increased or decreased by providing a difference in the void coefficient between the upper part and the lower parts of the reactor core, and increasing the void coefficient at the lower part of the reactor core. Constitution: The void coefficient of the lower region from the center to the lower part along the axial direction of a nuclear fuel assembly is increased to decrease the dependence on the flow rate of the axial power distribution of the nuclear fuel assembly. That is, a water/fuel ratio is varied, the water in non-boiled region is increased or the neutron spectrum is varied so as to vary the void coefficient. In order to exemplify it, the rate of the internal pellets of the fuel rod of the nuclear fuel assembly or the shape of the channel box is varied. Accordingly, the power does not considerably vary even if the flow rate is altered since the power is varied in the power operation. (Yoshihara, H.)

Design of 50 MWe HTR-PBMR reactor core and nuclear power plant fuel using SRAC2006 programme

International Nuclear Information System (INIS)

Bima Caraka Putra; Yosaphat Sumardi; Yohannes Sardjono

2014-01-01

This research aims to assess the design of core and fuel of nuclear power plant type High Temperature Reactor-Pebble Bed Modular Reactor 50 MWe from the Beginning of Life (BOL) to Ending of life (EOL) with eight years operating life. The parameters that need to be analyzed in this research are the temperature distribution inside the core, quantity enrichment of U 235 , fuel composition, criticality, and temperature reactivity coefficient of the core. The research was conducted with a data set of core design parameters such as nuclides density, core and fuel dimensions, and the axial temperature distribution inside the core. Using SRAC2006 program package, the effective multiplication factor (k eff ) values obtained from the input data that has been prepared. The results show the value of the criticality of core is proportional to the addition of U 235 enrichment. The optimum enrichment obtained at 10.125% without the use of burnable poison with an excess reactivity of 3.1 2% at BOL. The addition Gd 2O3 obtained an optimum value of 12 ppm burnable poison with an excess reactivity 0.38 %. The use of Er 2O3 with an optimum value 290 ppm has an excess reactivity 1.24 % at BOL. The core temperature reactivity coefficient with and without the use of burnable poison has a negative values that indicates the nature of its inherent safety. (author)

Steady state thermal hydraulic analysis of a boiling water reactor core, for various power distributions, using computer code THABNA

International Nuclear Information System (INIS)

Venkat Raj, V.; Saha, D.

1976-01-01

The core of a boiling water reactor may see different power distributions during its operational life. How some of the typical power distributions affect some of the thermal hydraulic parameters such as pressure drop minimum critical heat flux ratio, void distribution etc. has been studied using computer code THABNA. The effect of an increase in the leakage flow has also been analysed. (author)

Code of practice for in-core instrumentation for neutron fluence rate (flux) measurements in power reactors

International Nuclear Information System (INIS)

Anon.

1982-01-01

This standard applies to in-core (on-line) neutron detectors and instrumentation which is designed for safety, information or control purposes. It also applies to components in so far as these components are contained within the primary envelope of the reactor. The detector types usually used are dc ionization chambers and self-powered neutron detectors

Lamination effects on a 3D model of the magnetic core of power transformers

Directory of Open Access Journals (Sweden)

Poveda-Lerma Antonio

2017-12-01

Full Text Available In this paper the lamination effect on the model of a power transformer’s core with stacked E-I structure is analyzed. The distribution of the magnetic flux in the laminations depends on the stacking method. In this work it is shown, using a 3D FEM model and an experimental prototype, that the non-uniform distribution of the flux in a laminated E-I core with alternate-lap joint stack increases substantially the average value of the magnetic flux density in the core, compared with a butt joint stack. Both the simulated model and the experimental tests show that the presence of constructive air-gaps in the E-I junctions gives rise to a zig-zag flux in the depth direction. This inter-lamination flux reduces the magnetic flux density in the I-pieces and increases substantially the magnetic flux density in the E-pieces, with highly saturated points that traditional 2D analysis cannot reproduce. The relation between the number of laminations included in the model, and the computational resourses needed to build it, is also evaluated in this work.

Reactor core performance calculating device

International Nuclear Information System (INIS)

Tominaga, Kenji; Bando, Masaru; Sano, Hiroki; Maruyama, Hiromi.

1995-01-01

The device of the present invention can calculate a power distribution efficiently at high speed by a plurality of calculation means while taking an amount of the reactor state into consideration. Namely, an input device takes data from a measuring device for the amount of the reactor core state such as a large number of neutron detectors disposed in the reactor core for monitoring the reactor state during operation. An input data distribution device comprises a state recognition section and a data distribution section. The state recognition section recognizes the kind and amount of the inputted data and information of the calculation means. The data distribution section analyzes the characteristic of the inputted data, divides them into a several groups, allocates them to each of the calculation means for the purpose of calculating the reactor core performance efficiently at high speed based on the information from the state recognition section. A plurality of the calculation means calculate power distribution of each of regions based on the allocated inputted data, to determine the power distribution of the entire reactor core. As a result, the reactor core can be evaluated at high accuracy and at high speed irrespective of the whole reactor core or partial region. (I.S.)

WWER-440 reactor thermal power increase. Up-to-date approaches to substantiation of the core heat-engineering reliability

International Nuclear Information System (INIS)

Vasilchenko, I.; Lushin, V.; Zubtsov, D.

2006-01-01

Increasing the Units power is an urgent problem for nuclear power plants with WWER-440 reactors. Improving the fuel assembly designs and calculated codes creates all prerequisites to fulfil this purpose. The decrease in the core power peaking is reached by using the profiled fuel assemblies, burnable absorber integrated into the fuel, the FA with the modernized interface attachment, modern calculated codes that allows to reduce conservatism of the RP safety substantiation. A wide spectrum of experimental study of behaviour of the fuel having reached burn-up (50-60) MW days / kg U under the transients and accident conditions was carried out, the post-irradiated examination of the fuel assemblies, fuel rods and fuel pellets with four and five annual operating fuel cycle were performed as well and confirmed the high reliability of the fuel, the presence of large margins of the fuel stack state that contributes to reactor operation at the increased power. The results of the carried out experiments on implementing the five and six annual fuel cycles show that the limiting fuel state as to its serviceability in the WWER-440 reactors is far from being reached. Presently there is an experience of the increased power operation of Kola NPP, Units 1, 2, 4 and Rovno NPP, Unit 2. The Loviisa NPP Units are operated at 109 % power. The Russian experts had gained an experience in substantiating the core operation at 108 % power for Paks NPP, Unit 4. In this paper the additional conditions for increasing the power of the Kola NPP, Units 1 and 2 and the main results of substantiation of increase in power of the Paks NPP, Unit 4 up to 1485 MW are presented in details

Advanced gadolinia core and Toshiba advanced reactor management system

International Nuclear Information System (INIS)

Miyamoto, Toshiki; Yoshioka, Ritsuo; Ebisuya, Mitsuo

1988-01-01

At the Hamaoka Nuclear Power Station, Unit No. 3, advanced core design and core management technology have been adopted, significantly improving plant availability, operability and reliability. The outstanding technologies are the advanced gadolinia core (AGC) which utilizes gadolinium for the axial power distribution control, and Toshiba advanced reactor management system (TARMS) which uses a three-dimensional core physics simulator to calculate the power distribution. Presented here are the effects of these advanced technologies as observed during field testing. (author)

Analysis of the documents about the core envelopment of nuclear reactor at the Laguna Verde U-1 power plant

International Nuclear Information System (INIS)

Zamora R, L.; Medina F, A.

1999-01-01

The degradation of internal components at BWR type reactors is an important subject to consider in the performance availability of the power plant. The Wuergassen nuclear reactor license was confiscated due to the presence of cracking in the core envelopment. In consequence it is necessary carrying out a detailed study with the purpose to avoid these problems in the future. This report presents a review and analysis of documents and technical information referring to the core envelopment of a BWR/5/6 and the Laguna Verde Unit 1 nuclear reactor in Mexico. In this document are presented design data, documents about fabrication processes, and manufacturing of core envelopment. (Author)

The core paradox.

Science.gov (United States)

Kennedy, G. C.; Higgins, G. H.

1973-01-01

Rebuttal of suggestions from various critics attempting to provide an escape from the seeming paradox originated by Higgins and Kennedy's (1971) proposed possibility that the liquid in the outer core was thermally stably stratified and that this stratification might prove a powerful inhibitor to circulation of the outer core fluid of the kind postulated for the generation of the earth's magnetic field. These suggestions are examined and shown to provide no reasonable escape from the core paradox.

Power control system in BWR type reactors

International Nuclear Information System (INIS)

Nishizawa, Yasuo.

1980-01-01

Purpose: To control the reactor power so that the power distribution can satisfy the limiting conditions, by regulating the reactor core flow rate while monitoring the power distribution in the reactor core of a BWR type reactor. Constitution: A power distribution monitor determines the power distribution for the entire reactor core based on the data for neutron flux, reactor core thermal power, reactor core flow rate and control rod pattern from the reactor and calculates the linear power density distribution. A power up ratio computing device computes the current linear power density increase ratio. An aimed power up ratio is determined by converting the electrical power up ratio transferred from a load demand input device into the reactor core thermal power up ratio. The present reactor core thermal power up ratio is subtracted from the limiting power up ratio and the difference is sent to an operation amount indicator and the reactor core flow rate is changed in a reactor core flow rate regulator, by which the reactor power is controlled. (Moriyama, K.)

LMFBR Ultra Long Life Cores

International Nuclear Information System (INIS)

Schmidt, J.E.; Doncals, R.A.; Porter, C.A.; Gundy, L.M.

1986-01-01

The Ultra Long Life Core is an attractive and innovative design approach with several extremely beneficial attributes. Long Life cores are applicable to the full range of LMR plant sizes resulting in lifetimes up to 30 years. Core life is somewhat limited for smaller plant sizes, however significant benefits of this approach still exist for all plant sizes. The union of long life cores and the complementary inherent safety technology offer a means of utilizing the well-proven oxide fuel in a system with unsurpassed safety capability. A further benefit is that the uranium fuel cycle can be used in long life cores, especially for initial LMR plant deployment, thereby eliminating the need for reprocessing prior to starting LMR plant construction in the U.S. Finally the long life core significantly reduces power costs. With inherent safety capability designed into an LMR and with the ULLC fuel cycle, power costs competitive with light water plants are achievable while offering improved operational flexibility derived through extending refueling intervals

The PWR cores management

International Nuclear Information System (INIS)

Barral, J.C.; Rippert, D.; Johner, J.

2000-01-01

During the meeting of the 25 january 2000, organized by the SFEN, scientists and plant operators in the domain of the PWR debated on the PWR cores management. The five first papers propose general and economic information on the PWR and also the fast neutron reactors chains in the electric power market: statistics on the electric power industry, nuclear plant unit management, the ITER project and the future of the thermonuclear fusion, the treasurer's and chairman's reports. A second part offers more technical papers concerning the PWR cores management: performance and optimization, in service load planning, the cores management in the other countries, impacts on the research and development programs. (A.L.B.)

77 FR 30435 - In-core Thermocouples at Different Elevations and Radial Positions in Reactor Core

Science.gov (United States)

2012-05-23

... NUCLEAR REGULATORY COMMISSION 10 CFR Part 50 [Docket No. PRM-50-105; NRC-2012-0056] In-core Thermocouples at Different Elevations and Radial Positions in Reactor Core AGENCY: Nuclear Regulatory Commission... of operating licenses for nuclear power plants (``NPP'') to operate NPPs with in-core thermocouples...

Development of an inconel self powered neutron detector for in-core reactor monitoring

Energy Technology Data Exchange (ETDEWEB)

Alex, M. [Electronics Division, BARC, Mumbai (India)]. E-mail: maryalex@barc.gov.in; Ghodgaonkar, M.D. [Electronics Division, BARC, Mumbai (India)

2007-04-21

The paper describes the development and testing of an Inconel600 (2 mm diameterx21 cm long) self-powered neutron detector for in-core neutron monitoring. The detector has 3.5 mm overall diameter and 22 cm length and is integrally coupled to a 12 m long mineral insulated cable. The performance of the detector was compared with cobalt and platinum detectors of similar dimensions. Gamma sensitivity measurements performed at the {sup 60}Co irradiation facility in 14 MR/h gamma field showed values of -4.4x10{sup -18} A/R/h/cm (-9.3x10{sup -24} A/{gamma}/cm{sup 2}-s/cm), -5.2x10{sup -18} A/R/h/cm (-1.133x10{sup -23} A/{gamma}/cm{sup 2}-s/cm) and 34x10{sup -18} A/R/h/cm (7.14x10{sup -23} A/{gamma}/cm{sup 2}-s/cm) for the Inconel, Co and Pt detectors, respectively. The detectors together with a miniature gamma ion chamber and fission chamber were tested in the in-core Apsara Swimming Pool type reactor. The ion chambers were used to estimate the neutron and gamma fields. With an effective neutron cross-section of 4b, the Inconel detector has a total sensitivity of 6x10{sup -23} A/nv/cm while the corresponding sensitivities for the platinum and cobalt detectors were 1.69x10{sup -22} and 2.64x10{sup -22} A/nv/cm. The linearity of the detector responses at power levels ranging from 100 to 200 kW was within {+-}5%. The response of the detectors to reactor scram showed that the prompt response of the Inconel detector was 0.95 while it was 0.7 and 0.95 for the platinum and cobalt self-powered detectors, respectively. The detector was also installed in the horizontal flux unit of 540 MW Pressurised Heavy Water Reactor (PHWR). The neutron flux at the detector location was calculated by Triveni code. The detector response was measured from 0.02% to 0.07% of full power and showed good correlation between power level and detector signals. Long-term tests and the dynamic response of the detector to shut down in PHWR are in progress.

Reactor physics data for safety analysis of CANFLEX-NU CANDU-6 core

Energy Technology Data Exchange (ETDEWEB)

Jeong, Chang Joon; Suk, Ho Chun

2001-08-01

This report contains the reactor physics data for safety analysis of CANFLEX-NU fuel CANDU-6 core. First, the physics parameters for time-average core have been described, which include the channel power and maximum bundle power map, channel axial power shape and bundle burnup. And, next the data for fuel performance such as relative ring power distribution and bundle burnup conversion ratio are represented. The transition core data from 0 to 900 full power day are represented by 100 full power day interval. Also, the data for reactivity devices of time-average core and 300 full power day of transition core are given.

JOYO MK-II core characteristics database

International Nuclear Information System (INIS)

Ohkawachi, Yasushi; Maeda, Shigetaka; Sekine, Takashi; Aoyama, Takafumi

2003-04-01

The 'JOYO' MK-II core characteristics database was compiled and published in 1998. Comments and requests from many users led to the creation of a revised edition. The revisions include changes to the MAGI calculation code system to use the 70 group JFS-3-J3.2 constant set processed from the JENDL-3.2 library. Total control rod worth, reactor kinetic parameters and the MK-II core performance test results were included per user's requests. The core characteristics obtained from the 32 nd to 35 th operational cycles, which were conducted in the MK-III transition core, were newly added in this revised version. The MK-II core management data and core characteristics data were recorded to CD-ROM for user convenience. The Configuration Data' include the core arrangement and refueling record for each operational cycle. The 'Subassembly Library Data' include the atomic number density, neutron fluence, burn-up, integral power of 362 driver fuel subassemblies and 69 irradiation test subassemblies. The 'Output Data' contain the calculated neutron flux, gamma flux, power density, linear heat rate, coolant and fuel temperature distribution of all the fuel subassemblies at the beginning and end of each operational cycle. The 'Core Characteristics Data' include the measured excess reactivity, control rod worth calibration curve, and reactivity coefficients of temperature, power and burn-up. (author)

Core status computing system

International Nuclear Information System (INIS)

Yoshida, Hiroyuki.

1982-01-01

Purpose: To calculate power distribution, flow rate and the like in the reactor core with high accuracy in a BWR type reactor. Constitution: Total flow rate signals, traverse incore probe (TIP) signals as the neutron detector signals, thermal power signals and pressure signals are inputted into a process computer, where the power distribution and the flow rate distribution in the reactor core are calculated. A function generator connected to the process computer calculates the absolute flow rate passing through optional fuel assemblies using, as variables, flow rate signals from the introduction part for fuel assembly flow rate signals, data signals from the introduction part for the geometrical configuration data at the flow rate measuring site of fuel assemblies, total flow rate signals for the reactor core and the signals from the process computer. Numerical values thus obtained are given to the process computer as correction signals to perform correction for the experimental data. (Moriyama, K.)

Optical Bistability in Graded Core-Shell Granular Composites

International Nuclear Information System (INIS)

Wu Ya-Min; Chen Guo-Qing; Xue Si-Zhong; Zhu Zhuo-Wei; Ma Chao-Qun

2012-01-01

The intrinsic optical bistability (OB) of graded core-shell granular composites is investigated. The coated particles are made of cores with gradient dielectric function in c (r) = A(r/a) k and nonlinear shells. In view of the exponential distribution of the core dielectric constant, the potential functions of each region are obtained by solving the Maxwell equations, and the mathematical expressions of electric field in the shells and cores are determined. Numerical study reveals that the optical bistable threshold and the threshold width of the composite medium are dependent on the shell thickness, core dielectric exponent, and power function coefficient. The optical bistable width increases with the decreasing shell thickness and the power exponent and with the increasing power function coefficient

Electromechanical modeling of a honeycomb core integrated vibration energy converter with increased specific power for energy harvesting applications

Science.gov (United States)

Chandrasekharan, Nataraj

especially if the application imposes a space/size constraint. Moreover, the bimorph with increased thickness will now require a larger mechanical force to deform the structure which can fall outside the input ambient excitation amplitude range. In contrast, the honeycomb core bimorph offers an advantage in terms of preserving the global geometric dimensions. The natural frequency of the honeycomb core bimorph can be altered by manipulating honeycomb cell design parameters, such as cell angle, cell wall thickness, vertical cell height and inclined cell length. This results in a change in the mass and stiffness properties of the substrate and hence the bimorph, thereby altering the natural frequency of the harvester. Design flexibility of honeycomb core bimorphs is demonstrated by varying honeycomb cell parameters to alter mass and stiffness properties for power harvesting. The influence of honeycomb cell parameters on power generation is examined to evaluate optimum design to attain highest specific power. In addition, the more compliant nature of a honeycomb core bimorph decreases susceptibility towards fatigue and can increase the operating lifetime of the harvester. The second component of this dissertation analyses an uncoupled equivalent circuit model for piezoelectric energy harvesting. Open circuit voltage developed on the piezoelectric materials can be easily computed either through analytical or finite element models. The efficacy of a method to determine power developed across a resistive load, by representing the coupled piezoelectric electromechanical problem with an external load as an open circuit voltage driven equivalent circuit, is evaluated. The lack of backward feedback at finite resistive loads resulting from such an equivalent representation is examined by comparing the equivalent circuit model to the governing equations of a fully coupled circuit model for the electromechanical problem. It is found that the backward feedback is insignificant for weakly

Development of on-line core performance evaluation system for 'FUGEN'

International Nuclear Information System (INIS)

Natori, Hisahide; Kaneto, Kunikazu; Oteru, Shigeru.

1982-01-01

An on-line core performance evaluation system ATROPOS has been developed in order to carry out safe and efficient reactor operation of ''FUGEN''(a heavy water moderated, boiling light water cooled, pressure tube type reactor). This system offers detailed and useful information on such items of core performance as core thermal power, power distribution and thermal operation limits. The power distribution is calculated first by using a three-dimensional nodal coupling model, employing such process data as control rod position and 10 B concentration in the D 2 O moderator. Then the calculated power distribution is corrected by local power monitor readings. An axial one-dimensional nodal coupling model, which considers radial power distribution, and a localized three-dimensional nodal coupling model are used to predict the core thermal power and the power distribution for the region surrounding the control rods respectively, within a short time in advance of control rod operation. The methods employed in this system are verified by comparison with start-up test data from the FUGEN initial core. The estimated power distribution and channel flow agree with values measured by the power calibration monitor and with channel flow converted from measured values of pressure drop, within 3 and 5% respectively, in their root mean square values. The difference in core thermal power between the predicted value and the value measured by the total power monitor is about 1% for control rod operation. (author)

On-line calculation of 3-D power distribution

International Nuclear Information System (INIS)

Park, Y. H.; In, W. K.; Park, J. R.; Lee, C. C.; Auh, G. S.

1996-01-01

The 3-D power distribution synthesis scheme was implemented in Totally Integrated Core Operation Monitoring System (TICOMS), which is under development as the next generation core monitoring system. The on-line 3-D core power distribution obtained from the measured fixed incore detector readings is used to construct the hot pin power as well as the core average axial power distribution. The core average axial power distribution and the hot pin power of TICOMS were compared with those of the current digital on-line core monitoring system, COLSS, which construct the core average axial power distribution and the pseudo hot pin power. The comparison shows that TICOMS results in the slightly more accurate core average axial power distribution and the less conservative hot pin power. Therefore, these results increased the core operating margins. In addition, the on-line 3-D power distribution is expected to be very useful for the core operation in the future

First order study for an iron core OH system for TNS

International Nuclear Information System (INIS)

Ballou, J.K.; Schultz, J.

1977-01-01

A simple comparison has been made between an air core and an iron core ohmic heating system for a particular device, and it was shown that the peak power requirements can be substantially reduced by the use of an iron core to power levels handled by industry today. It was also shown that for an ohmic heating system initiated plasma that the cost of the iron core ohmic heating power system (iron core, dual rectifier, and DC switch) is less than the cost for a subset of the power system for an air core system (dual rectifier and DC switch). There is considerable work being done on other methods of initiating the plasma none of which seem to be incompatible with the use of an iron core system

Calculation of ex-core detector responses

Energy Technology Data Exchange (ETDEWEB)

Wouters, R. de; Haedens, M. [Tractebel Engineering, Brussels (Belgium); Baenst, H. de [Electrabel, Brussels (Belgium)

2005-07-01

The purpose of this work carried out by Tractebel Engineering, is to develop and validate a method for predicting the ex-core detector responses in the NPPs operated by Electrabel. Practical applications are: prediction of ex-core calibration coefficients for startup power ascension, replacement of xenon transients by theoretical predictions, and analysis of a Rod Drop Accident. The neutron diffusion program PANTHER calculates node-integrated fission sources which are combined with nodal importance representing the contribution of a neutron born in that node to the ex-core response. These importance are computed with the Monte Carlo program MCBEND in adjoint mode, with a model of the whole core at full power. Other core conditions are treated using sensitivities of the ex-core responses to water densities, computed with forward Monte Carlo. The Scaling Factors (SF), or ratios of the measured currents to the calculated response, have been established on a total of 550 in-core flux maps taken in four NPPs. The method has been applied to 15 startup transients, using the average SF obtained from previous cycles, and to 28 xenon transients, using the SF obtained from the in-core map immediately preceding the transient. The values of power (P) and axial offset (AOi) reconstructed with the theoretical calibration agree well with the measured values. The ex-core responses calculated during a rod drop transient have been successfully compared with available measurements, and with theoretical data obtained by alternative methods. In conclusion, the method is adequate for the practical applications previously listed. (authors)

JOYO MK-II core characteristics database

International Nuclear Information System (INIS)

Tabuchi, Shiro; Aoyama, Takafumi; Nagasaki, Hideaki; Kato, Yuichi

1998-12-01

The experimental fast reactor JOYO served as the MK-II irradiation bed core for testing fuel and material for FBR development for 15 years from 1982 to 1997. During the MK-II operation, extensive data were accumulated from the core characteristics tests conducted in thirty-one duty operations and thirteen special test operations. These core management data and core characteristics data were compiled into a database. The code system MAGI has been developed and used for core management of JOYO MK-II, and the core characteristics and the irradiation test conditions were calculated using MAGI on the basis of three dimensional diffusion theory with seven neutron energy groups. The core management data include extensive data, which were recorded on CD-ROM for user convenience. The data are specifications and configurations of the core, and for about 300 driver fuel subassemblies and about 60 uninstrumented irradiation subassemblies are core composition before and after irradiation, neutron flux, neutron fluences, fuel and control rod burn-up, and temperature and power distributions. MK-II core characteristics and test conditions were stored in the database for post analysis. Core characteristics data include excess reactivities, control rod worths, and reactivity coefficients, e.g., temperature, power and burn-up. Test conditions include both measured and calculated data for irradiation conditions. (author)

Review of the Shoreham Nuclear Power Station Probabilistic Risk Assessment: internal events and core damage frequency

International Nuclear Information System (INIS)

Ilberg, D.; Shiu, K.; Hanan, N.; Anavim, E.

1985-11-01

A review of the Probabilistic Risk Assessment of the Shoreham Nuclear Power Station was conducted with the broad objective of evaluating its risks in relation to those identified in the Reactor Safety Study (WASH-1400). The scope of the review was limited to the ''front end'' part, i.e., to the evaluation of the frequencies of states in which core damage may occur. Furthermore, the review considered only internally generated accidents, consistent with the scope of the PRA. The review included an assessment of the assumptions and methods used in the Shoreham study. It also encompassed a reevaluation of the main results within the scope and general methodological framework of the Shoreham PRA, including both qualitative and quantitative analyses of accident initiators, data bases, and accident sequences which result in initiation of core damage. Specific comparisons are given between the Shoreham study, the results of the present review, and the WASH-1400 BWR, for the core damage frequency. The effect of modeling uncertainties was considered by a limited sensitivity study so as to show how the results would change if other assumptions were made. This review provides an independently assessed point value estimate of core damage frequency and describes the major contributors, by frontline systems and by accident sequences. 17 figs., 81 tabs

Two and three dimensional core power distribution monitor and display

International Nuclear Information System (INIS)

Impink, A.J. Jr.; Grobmyer, L.R.

1988-01-01

This patent describes a sensor monitoring system for displaying a profile of fractional deviations in relative coolant enthalpy rise over a defined area comprising at least a part of a core of a nuclear reactor, which system comprises: core exit coolant temperature sensors positioned to monitor at least a portion of the defined area; an inlet temperature sensor outside the core which monitors the temperature of core coolant at an inlet to the reactor means, responsive to the outputs from both the core exit temperature sensors and the inlet temperature sensor, for generating corresponding representative values of actual coolant enthalpy rise and corresponding values of relative enthalpy rise at each location in the defined area at which a core exit coolant temperature sensor is available; means, responsive to the generated values of relative enthalpy rise and to reference values of relative enthalpy rise at corresponding locations in the defined area, for generating values of the fractional deviation of the measured values of relative enthalpy rise from the corresponding values; means for interpolating the generated values of fractional deviation in relative enthalpy rise to provide interpolated values of fractional deviation in relative enthalpy rise at locations in the defined area of the core other than those at which core exit coolant temperature sensors are available; and means for multidimensionally displaying the generated and interpolated values

Nuclear power reactor core melt accidents. Current State of Knowledge

International Nuclear Information System (INIS)

Jacquemain, Didier; Cenerino, Gerard; Corenwinder, Francois; Raimond, Emmanuel IRSN; Bentaib, Ahmed; Bonneville, Herve; Clement, Bernard; Cranga, Michel; Fichot, Florian; Koundy, Vincent; Meignen, Renaud; Corenwinder, Francois; Leteinturier, Denis; Monroig, Frederique; Nahas, Georges; Pichereau, Frederique; Van-Dorsselaere, Jean-Pierre; Couturier, Jean; Debaudringhien, Cecile; Duprat, Anna; Dupuy, Patricia; Evrard, Jean-Michel; Nicaise, Gregory; Berthoud, Georges; Studer, Etienne; Boulaud, Denis; Chaumont, Bernard; Clement, Bernard; Gonzalez, Richard; Queniart, Daniel; Peltier, Jean; Goue, Georges; Lefevre, Odile; Marano, Sandrine; Gobin, Jean-Dominique; Schwarz, Michel; Repussard, Jacques; Haste, Tim; Ducros, Gerard; Journeau, Christophe; Magallon, Daniel; Seiler, Jean-Marie; Tourniaire, Bruno; Durin, Michel; Andreo, Francois; Atkhen, Kresna; Daguse, Thierry; Dubreuil-Chambardel, Alain; Kappler, Francois; Labadie, Gerard; Schumm, Andreas; Gauntt, Randall O.; Birchley, Jonathan

2015-11-01

For over thirty years, IPSN and subsequently IRSN has played a major international role in the field of nuclear power reactor core melt accidents through the undertaking of important experimental programmes (the most significant being the Phebus-FP programme), the development of validated simulation tools (the ASTEC code that is today the leading European tool for modelling severe accidents), and the coordination of the SARNET (Severe Accident Research Network) international network of excellence. These accidents are described as 'severe accidents' because they can lead to radioactive releases outside the plant concerned, with serious consequences for the general public and for the environment. This book compiles the sum of the knowledge acquired on this subject and summarises the lessons that have been learnt from severe accidents around the world for the prevention and reduction of the consequences of such accidents, without addressing those from the Fukushima accident, where knowledge of events is still evolving. The knowledge accumulated by the Institute on these subjects enabled it to play an active role in informing public authorities, the media and the public when this accident occurred, and continues to do so to this day. Following the introduction, which describes the structure of this book and highlights the objectives of R and D on core melt accidents, this book briefly presents the design and operating principles (Chapter 2) and safety principles (Chapter 3) of the reactors currently in operation in France, as well as the main accident scenarios envisaged and studied (Chapter 4). The objective of these chapters is not to provide exhaustive information on these subjects (the reader should refer to the general reference documents listed in the corresponding chapters), but instead to provide the information needed in order to understand, firstly, the general approach adopted in France for preventing and mitigating the consequences of core melt

Evaluation report on CCTF core-II reflood test C2-6 (Run 64)

International Nuclear Information System (INIS)

Akimoto, Hajime; Iguchi, Tadashi; Sugimoto, Jun; Okubo, Tsutomu; Murao, Yoshio; Okabe, Kazuharu.

1985-03-01

In order to evaluate the effect of the radial power profile on the system behavior and the core thermal hydraulic behavior during the reflood phase of a PWR LOCA, a test was performed using the Cylindrical Core Test Facility(CCTF) with the flat radial power profile. The test was conducted with the same total core power as that of the steep radial power test C2-5(Run 63). Through the comparisons of the results from these two tests, the following conclusions were obtained: (1) The radial power profile in the core has weak effect on the thermal hydraulic behavior in the primary system except the core. (2) Almost the same differential pressure was observed at various elevations in the periphery of the core regardless of different radial power profile. The result suggests that the core differential pressure is determined mainly by the total power and the total stored energy rather than by the local power and the local stored energy. (3) The test results support the single channel core model with the average power rod used in the reactor safety analysis codes such as REFLA-1DS, WREM for the evaluation of the overall system behavior. (4) In the steep radial power test, the heat transfer coefficient in the central(high power) region was higher than that in the peripheral(low power) region. The tendency was not explained by the estimation with the heat transfer correlation developed by Murao and Sugimoto assuming that the void fraction was uniform in a horizontal cross section. It is necessary to study more the dependency of core heat transfer on the radial power profile in the wide core. (author)

A Systematic Approach to Design Low-Power Video Codec Cores

Directory of Open Access Journals (Sweden)

Corporaal Henk

2007-01-01

Full Text Available The higher resolutions and new functionality of video applications increase their throughput and processing requirements. In contrast, the energy and heat limitations of mobile devices demand low-power video cores. We propose a memory and communication centric design methodology to reach an energy-efficient dedicated implementation. First, memory optimizations are combined with algorithmic tuning. Then, a partitioning exploration introduces parallelism using a cyclo-static dataflow model that also expresses implementation-specific aspects of communication channels. Towards hardware, these channels are implemented as a restricted set of communication primitives. They enable an automated RTL development strategy for rigorous functional verification. The FPGA/ASIC design of an MPEG-4 Simple Profile video codec demonstrates the methodology. The video pipeline exploits the inherent functional parallelism of the codec and contains a tailored memory hierarchy with burst accesses to external memory. 4CIF encoding at 30 fps, consumes 71 mW in a 180 nm, 1.62 V UMC technology.

A Systematic Approach to Design Low-Power Video Codec Cores

Directory of Open Access Journals (Sweden)

Kristof Denolf

2007-05-01

Full Text Available The higher resolutions and new functionality of video applications increase their throughput and processing requirements. In contrast, the energy and heat limitations of mobile devices demand low-power video cores. We propose a memory and communication centric design methodology to reach an energy-efficient dedicated implementation. First, memory optimizations are combined with algorithmic tuning. Then, a partitioning exploration introduces parallelism using a cyclo-static dataflow model that also expresses implementation-specific aspects of communication channels. Towards hardware, these channels are implemented as a restricted set of communication primitives. They enable an automated RTL development strategy for rigorous functional verification. The FPGA/ASIC design of an MPEG-4 Simple Profile video codec demonstrates the methodology. The video pipeline exploits the inherent functional parallelism of the codec and contains a tailored memory hierarchy with burst accesses to external memory. 4CIF encoding at 30 fps, consumes 71 mW in a 180 nm, 1.62 V UMC technology.

In core reload design for cycle 4 of Daya Bay nuclear power station both units

International Nuclear Information System (INIS)

Zhang Zongyao; Liu Xudong; Xian Chunyu; Li Dongsheng; Zhang Hong; Liu Changwen; Rui Min; Wang Yingming; Zhao Ke; Zhang Hong; Xiao Min

1998-01-01

The basic principles and the contents of the reload design for Daya Bay nuclear power station are briefly introduced. The in core reload design results, and the comparison between the calculated values and the measured values of both units the fourth cycle are also given. The reload design results of the two units satisfy all the economic requirements and safety criteria. The experimented results shown that the predicated values are tally good with all the measurement values

Modeling and analysis of the disk MHD generator component of a gas core reactor/MHD Rankine cycle space power system

International Nuclear Information System (INIS)

Welch, G.E.; Dugan, E.T.; Lear, W.E. Jr.; Appelbaum, J.G.

1990-01-01

A gas core nuclear reactor (GCR)/disk magnetohydrodynamic (MHD) generator direct closed Rankine space power system concept is described. The GCR/disk MHD generator marriage facilitates efficient high electric power density system performance at relatively high operating temperatures. The system concept promises high specific power levels, on the order of 1 kW e /kg. An overview of the disk MHD generator component magnetofluiddynamic and plasma physics theoretical modeling is provided. Results from a parametric design analysis of the disk MHD generator are presented and discussed

Oyster Creek fuel thermal margin during core thermal-hydraulic oscillations

International Nuclear Information System (INIS)

Dougher, J.D.

1990-01-01

The Oyster Creek nuclear facility, a boiling water reactor (BWR)-2 plant type, has never experienced core thermal-hydraulic instability. Power oscillations, however, have been observed in other BWR cores both domestically and internationally. Two modes of oscillations have been observed, core wide and regional half-core. During core wide oscillations, the neutron flux in the core oscillates in the radial fundamental mode. During regional half-core oscillations, higher order harmonics in the radial plane result in out-of-phase oscillations with the neutron flux in one half of the core oscillating 180 deg out-of-phase with the neutron flux in the other half of the core. General Design Criteria 12 requires either prevention or detection and suppression of power oscillations which could result in violations of fuel design limits. Analyses performed by General Electric have demonstrated that for large-magnitude oscillations the potential exists for violation of the safety limit minimum critical power ratio (MCPR). However, for plants with a flow-biased neutron flux scram automatic mitigation of oscillations may be provided at an oscillation magnitude below that at which the safety limit is challenged. Plant-specific analysis for Oyster Creek demonstrates that the existing average power range monitor (APRM) system will sense and suppress power oscillations prior to violation of any safety limits

Development of JOYO MK-II core characteristics database

International Nuclear Information System (INIS)

Tabuchi, Shiro; Aoyama, Takafumi

2000-01-01

The MK-II core of the experimental fast reactor JOYO served as the irradiation bed for testing fuels and materials for FBR development since 1982 for 15 years. During the MK-II operation, extensive data were accumulated from the core management calculations and characteristics tests conducted in thirty-one duty operations and thirteen special test operations. These core management data and core characteristics data were compiled into a database recorded on CD-ROM for user convenience. The calculated core management data are the text style data. The 'Configuration Data' include the history of the fuel exchange and core arrangement for each cycle. The Subassembly Library Data' include the atomic number density, neutron fluence, burn-up, integral power of about 300 fuel subassemblies, and 60 irradiation subassemblies. The 'Output Data' include the neutron fluxes, gamma fluxes, power density, linear heat rates, coolant and fuel temperature distributions of each core position at the beginning and end of each cycle. The measured core characteristics data, such as the excess reactivity, control rod worths, temperature coefficient, power coefficient, and burn-up coefficient are also included along with the measurement conditions. (J.P.N.)

Metal Amorphous Nanocomposite (MANC) Alloy Cores with Spatially Tuned Permeability for Advanced Power Magnetics Applications

Science.gov (United States)

Byerly, K.; Ohodnicki, P. R.; Moon, S. R.; Leary, A. M.; Keylin, V.; McHenry, M. E.; Simizu, S.; Beddingfield, R.; Yu, Y.; Feichter, G.; Noebe, R.; Bowman, R.; Bhattacharya, S.

2018-06-01

Metal amorphous nanocomposite (MANC) alloys are an emerging class of soft magnetic materials showing promise for a range of inductive components targeted for higher power density and higher efficiency power conversion applications including inductors, transformers, and rotating electrical machinery. Magnetization reversal mechanisms within these alloys are typically determined by composition optimization as well as controlled annealing treatments to generate a nanocomposite structure composed of nanocrystals embedded in an amorphous precursor. Here we demonstrate the concept of spatially varying the permeability within a given component for optimization of performance by using the strain annealing process. The concept is realized experimentally through the smoothing of the flux profile from the inner to outer core radius achieved by a monotonic variation in tension during the strain annealing process. Great potential exists for an extension of this concept to a wide range of other power magnetic components and more complex spatially varying permeability profiles through advances in strain annealing techniques and controls.

Development of CANDU core monitoring system

International Nuclear Information System (INIS)

Yoon, M. Y.; Yeam, C. S.; Kwon, O. H.; Kim, K. H.

2003-01-01

The research was performed to develop a CANDU Core Monitoring System(CCMS) that enables operators to have efficient core management by monitoring core power distribution, burnup distribution, and the other important core variables and managing the past core history for Wolsong Nuclear Power Plant(NPP) No. 1. CCMS uses RFSP(Reactor Fueling Simulation Program) for continuous core calculation by integrating the algorithm and assumptions validated and uses the information taken from DCC(Digital Control Computer) for the purpose of producing basic input data. CCMS could be largely divided into two modules; CCMS server program and CCMS client program. CCMS server program plays the role in automatic and continuous RFSP run and management of the past output data resulting from the run using Data Base Management System(DBMS). CCMS client program enables users to monitor current and past core status with GUI(Graphic-User Interface) environment predefined. The effectiveness of CCMS was verified by comparing the data resulted from field-test of the system for about 43 hours with the data used in the field of Wolsong NPP No. 1

Development of CANDU core monitoring system

Energy Technology Data Exchange (ETDEWEB)

Yoon, M. Y.; Yeam, C. S.; Kwon, O. H.; Kim, K. H. [Institute for Advanced Engineering, Yongin (Korea, Republic of)

2003-07-01

The research was performed to develop a CANDU Core Monitoring System(CCMS) that enables operators to have efficient core management by monitoring core power distribution, burnup distribution, and the other important core variables and managing the past core history for Wolsong Nuclear Power Plant(NPP) No. 1. CCMS uses RFSP(Reactor Fueling Simulation Program) for continuous core calculation by integrating the algorithm and assumptions validated and uses the information taken from DCC(Digital Control Computer) for the purpose of producing basic input data. CCMS could be largely divided into two modules; CCMS server program and CCMS client program. CCMS server program plays the role in automatic and continuous RFSP run and management of the past output data resulting from the run using Data Base Management System(DBMS). CCMS client program enables users to monitor current and past core status with GUI(Graphic-User Interface) environment predefined. The effectiveness of CCMS was verified by comparing the data resulted from field-test of the system for about 43 hours with the data used in the field of Wolsong NPP No. 1.

Improvement of Cycle Dependent Core Model for NPP Simulator

International Nuclear Information System (INIS)

Song, J. S.; Koo, B. S.; Kim, H. Y. and others

2003-11-01

The purpose of this study is to establish automatic core model generation system and to develop 4 cycle real time core analysis methodology with 5% power distribution and 500 pcm reactivity difference criteria for nuclear power plant simulator. The standardized procedure to generate database from ROCS and ANC, which are used for domestic PWR core design, was established for the cycle specific simulator core model generation. An automatic data interface system to generate core model also established. The system includes ARCADIS which edits group constant and DHCGEN which generates interface coupling coefficient correction database. The interface coupling coefficient correction method developed in this study has 4 cycle real time capability and accuracies of which the maximum differences between core design results are within 103 pcm reactivity, 1% relative power distribution and 6% control rod worth. A nuclear power plant core simulation program R-MASTER was developed using the methodology and applied by the concept of distributed client system in simulator. The performance was verified by site acceptance test in Simulator no. 2 in Kori Training Center for 30 initial condition generation and 27 steady state, transient and postulated accident situations

Improvement of Cycle Dependent Core Model for NPP Simulator

Energy Technology Data Exchange (ETDEWEB)

Song, J. S.; Koo, B. S.; Kim, H. Y. and others

2003-11-15

The purpose of this study is to establish automatic core model generation system and to develop 4 cycle real time core analysis methodology with 5% power distribution and 500 pcm reactivity difference criteria for nuclear power plant simulator. The standardized procedure to generate database from ROCS and ANC, which are used for domestic PWR core design, was established for the cycle specific simulator core model generation. An automatic data interface system to generate core model also established. The system includes ARCADIS which edits group constant and DHCGEN which generates interface coupling coefficient correction database. The interface coupling coefficient correction method developed in this study has 4 cycle real time capability and accuracies of which the maximum differences between core design results are within 103 pcm reactivity, 1% relative power distribution and 6% control rod worth. A nuclear power plant core simulation program R-MASTER was developed using the methodology and applied by the concept of distributed client system in simulator. The performance was verified by site acceptance test in Simulator no. 2 in Kori Training Center for 30 initial condition generation and 27 steady state, transient and postulated accident situations.

Web-based Core Design System Development

International Nuclear Information System (INIS)

Moon, So Young; Kim, Hyung Jin; Yang, Sung Tae; Hong, Sun Kwan

2011-01-01

The selection of a loading pattern is one of core design processes in the operation of a nuclear power plant. A potential new loading pattern is identified by selecting fuels that to not exceed the major limiting factors of the design and that satisfy the core design conditions for employing fuel data from the existing loading pattern of the current operating cycle. The selection of a loading pattern is also related to the cycle plan of an operating nuclear power plant and must meet safety and economic requirements. In selecting an appropriate loading pattern, all aspects, such as input creation, code runs and result processes are processed as text forms manually by a designer, all of which may be subject to human error, such as syntax or running errors. Time-consuming results analysis and decision-making processes are the most significant inefficiencies to avoid. A web-based nuclear plant core design system was developed here to remedy the shortcomings of an existing core design system. The proposed system adopts the general methodology of OPR1000 (Korea Standard Nuclear Power Plants) and Westinghouse-type plants. Additionally, it offers a GUI (Graphic User Interface)-based core design environment with a user-friendly interface for operators. It reduces human errors related to design model creation, computation, final reload core model selection, final output confirmation, and result data validation and verification. Most significantly, it reduces the core design time by more than 75% compared to its predecessor

Xenon oscillation tests in four-loop PWR cores

International Nuclear Information System (INIS)

Aoki, Norihiko; Osaka, Kenichi; Shimada, Shoichiro; Tochihara, Hiroshi; Machii, Seigo

1980-01-01

The Kansai Electric Power Co.'s OHI Unit 1 and 2 are the first 4-loop PWRs in Japan which use 17 x 17 fuel assemblies and have essentially the same plant parameters. A 4-loop core has larger core radius and higher power density than those of 2- or 3-loop cores, and is less stable for Xe oscillation. It is therefore important to confirm that Xe oscillations in radial direction are sufficiently stable in a 4-loop core. Radial and axial Xe oscillation tests were performed during the startup physics tests of OHI Unit 1 and 2; Xe oscillation was induced by perturbation of control rods and the Xe effect on power distribution observed periodically. The test results show that the transverse Xe oscillation in the 4-loop core is sufficiently stable and that the agreement between the measurement and the calculated prediction is good. (author)

Modular approach to LWR in-core fuel management

International Nuclear Information System (INIS)

Urli, N.; Pevec, D.; Coffou, E.; Petrovic, B.

1980-01-01

The most important methods in the LWR in-core fuel management are reviewed. A modular approach and optimization by use of infinite multiplication factor and power form-factor are favoured. A computer program for rotation of fuel assemblies at reloads has been developed which improves further fuel economy and reliability of nuclear power plants. The program has been tested on the PWR core and showed to decrease the power form-factors and flatten the radial power distribution. (author)

Strategic plan for the development of core technologies for the Korean advanced nuclear power reactor for export

International Nuclear Information System (INIS)

Moon, Joo Hyun; Cho, Young Ho

2010-01-01

With the soaring oil price and worsening global warming, nuclear power has attracted considerable attention on a global scale and a new large market of nuclear power plants (NPPs) is expected. The Korean government aims to export up to 10 NPPs by 2012, based on the successful export of 2 NPPs to the UAE in 2009. It is also going to develop a follow-up model of the Advanced Power Reactor (APR) 1400, and join the world's NPP market under the banner of Korea's original reactor type. For this, it promulgated the strategic plan, NuTech 2012, a technology development plan intended for the early acquisition of core technologies for the Korean advanced NPP design and domestic production of the main components in NPP. This paper introduces the strategic plan of NuTech 2012. (orig.)

Experience with mixed cores in the IRR-1

International Nuclear Information System (INIS)

Gilat, J.; Hirshfeld, H.; Wiener, A.

1985-01-01

Over twenty mixed core configurations composed of 'old' (18 curved plate) and 'new' 23 flat plate) MTR type fuel elements were irradiated in the IRR-1 swimming pool reactor. The number of 'new' fuel elements in the core varied from one to twenty. To establish the safety of these configurations, thermohydraulic calculations were carried out to derive the maximum allowed hot channel power, determined by the onset of flow instabilities. A core is considered safe if its hot channel power, obtained from a two-dimensional neutronic calculation of power distribution in the core, does not exceed the maximum allowed value. The conservative nature of the assumptions used in the above safety evaluation procedure was verified by measurements of pressure drops vs. coolant flow rates as well as of temperature and neutron flux distributions. (author)

Distribution of equilibrium burnup for an homogeneous core with fuel elements of slightly enriched uranium (0.85% U-235) at Atucha I nuclear power plant

International Nuclear Information System (INIS)

Sidelnik, J.I.; Perez, R.A.; Salom, G.F.

1987-01-01

At Atucha I, the present fuel management with natural uranium comprises three burnup areas and one irradiation path, sometimes performing four steps in the reactor core, according to the requirements. The discharge burnup is 6.0 Mw d/kg U for a waste reactivity of 6.5 m k and a heavy water purity of 99.75%. This is a preliminary study to obtain the distribution of equilibrium burnup of an homogeneous core with slightly enriched uranium (0.85% by weight U-235), using the time-averaged method implemented in the code PUMA and a representative model of one third of core and fixed rod position. It was found a strategy of three areas and two paths that agrees with the present limits of channel power and specific power in fuel rod. The discharge burnup obtained is 11.6 Mw d/kg U. This strategy is calculated with the same method and a full core representation model is used to verify the obtained results. (Author)

Ultra-short pulse delivery at high average power with low-loss hollow core fibers coupled to TRUMPF's TruMicro laser platforms for industrial applications

Science.gov (United States)

Baumbach, S.; Pricking, S.; Overbuschmann, J.; Nutsch, S.; Kleinbauer, J.; Gebs, R.; Tan, C.; Scelle, R.; Kahmann, M.; Budnicki, A.; Sutter, D. H.; Killi, A.

2017-02-01

Multi-megawatt ultrafast laser systems at micrometer wavelength are commonly used for material processing applications, including ablation, cutting and drilling of various materials or cleaving of display glass with excellent quality. There is a need for flexible and efficient beam guidance, avoiding free space propagation of light between the laser head and the processing unit. Solid core step index fibers are only feasible for delivering laser pulses with peak powers in the kW-regime due to the optical damage threshold in bulk silica. In contrast, hollow core fibers are capable of guiding ultra-short laser pulses with orders of magnitude higher peak powers. This is possible since a micro-structured cladding confines the light within the hollow core and therefore minimizes the spatial overlap between silica and the electro-magnetic field. We report on recent results of single-mode ultra-short pulse delivery over several meters in a lowloss hollow core fiber packaged with industrial connectors. TRUMPF's ultrafast TruMicro laser platforms equipped with advanced temperature control and precisely engineered opto-mechanical components provide excellent position and pointing stability. They are thus perfectly suited for passive coupling of ultra-short laser pulses into hollow core fibers. Neither active beam launching components nor beam trackers are necessary for a reliable beam delivery in a space and cost saving packaging. Long term tests with weeks of stable operation, excellent beam quality and an overall transmission efficiency of above 85 percent even at high average power confirm the reliability for industrial applications.

In-core fuel management via perturbation theory

International Nuclear Information System (INIS)

Mingle, J.O.

1975-01-01

A two-step procedure is developed for the optimization of in-core nuclear fuel management using perturbation theory to predict the effects of various core configurations. The first procedure is a cycle cost minimization using linear programming with a zoned core and discrete burnup groups. The second program utilizes an individual fuel assembly shuffling sequence to minimize the maldistribution of power generation. This latter quantity is represented by a figure of merit or by an assembly power peaking factor. A pressurized water reactor example calculation is utilized. 24 references

Multi-core System Architecture for Safety-critical Control Applications

DEFF Research Database (Denmark)

Li, Gang

and size, and high power consumption. Increasing the frequency of a processor is becoming painful now due to the explosive power consumption. Furthermore, components integrated into a single-core processor have to be certified to the highest SIL, due to that no isolation is provided in a traditional single...... certification cost. Meanwhile, hardware platforms with improved processing power are required to execute the applications of larger size. To tackle the two issues mentioned above, the state of the art approaches are using more Electronic Control Units (ECU) in a federated architecture or increasing......-core processor. A promising alternative to improve processing power and provide isolation is to adopt a multi-core architecture with on-chip isolation. In general, a specific multi-core architecture can facilitate the development and certification of safety-related systems, due to its physical isolation between...

Active core profile and transport modification by application of Ion Bernstein Wave power in PBX-M

International Nuclear Information System (INIS)

LeBlanc, B.; Bell, R.

1995-01-01

Application of Ion Bernstein Wave Heating (IBWH) into the Princeton Beta Experiment-Modification (PBX-M) tokamak stabilizes sawtooth oscillations and generates peaked density profiles. A transport barrier, spatially correlated with the IBWH power deposition profile, is observed in the core of IBWH assisted neutral beam injection (NBI) discharges. A precursor to the fully developed barrier is seen in the soft x-ray data during edge localized mode (ELM) activity. Sustained IBWH operation is conducive to a regime where the barrier supports large triangledown n e , triangledown T e , triangledown v phi , and triangledown T i , delimiting the confinement zone. This regime is reminiscent of the H(high)-mode but with a confinement zone moved inwards. The core region has better than H-mode confinement while the peripheral region is L(low)-mode-like. The peaked profile enhanced NBI core deposition and increases nuclear reactivity. An increase in central T i results from χ i reduction (compared to H-mode) and better beam penetration. Bootstrap current fractions of up to 0.32--0.35 locally and 0.28 overall were obtained when an additional NBI burst is applied to this plasma

5  W output power from a double-clad hybrid fiber with Yb-doped phosphate core and silicate cladding.

Science.gov (United States)

Wang, Longfei; He, Dongbing; Zhang, Lei; Yu, Chunlei; Feng, Suya; Wang, Meng; Chen, Danping; Hu, Lili

2017-08-01

For the first time, to the best of our knowledge, we report on the realization of a laser from a Yb-doped phosphate core/silicate cladding double-clad hybrid fiber. 5 W output power was extracted with 14.6% slope efficiency and a laser spectrum of a 1027 nm central wavelength from a 20 cm long single-mode fiber with a ∼10  μm core diameter in a 20%-4% laser cavity. The laser efficiency can be significantly enhanced by correspondingly adjusting and optimizing the laser oscillator.

Severe Accident Mitigation by using Core Catcher applicable for Korea standard nuclear power plant

Energy Technology Data Exchange (ETDEWEB)

Park, Hae Kyun; Kim, Sang Nyung [Kyung Hee Univ., Yongin (Korea, Republic of)

2013-10-15

Nuclear power plants have been designed and operated in order to prevent severe accident because of their risk that contains tremendous radioactive materials that are potentially hazardous. Moreover, the government requested the nuclear industry to implement a severe accident management strategy for existing reactors to mitigate the risk of potential severe accidents. However, Korea standard nuclear power plant(APR-1400 and OPR-1000) are much more vulnerable for severe accident management than that of developed countries. Due to the design feature of reactor cavity in Korea standard nuclear power plant, inequable and serious Molten Core-Concrete Interaction(MCCI) may cause considerable safety problem to the reactor containment liner. At worst, it brings the release of radioactive materials to the environment. This accident applies to the fourth level of defense in depth(IAEA 1996), 'severe accident'. This study proposes and designs the 'slope' to secure reactor containment liner integrity when the corium spreads out from the destroyed reactor vessel to the reactor cavity due to the core melting accident. For this, make the initial corium distribution evenly exploit the 'slope' on the basis of the study of Ex-vessel corium behavior to prevent inequable and serious MCCI, in order to mitigate severe accident. The viscosity has a dominant position in the calculation. According to the result, the spread out distance on the slope is 10.7146841m, considering the rough surface of the concrete(slope) and margin of reactor cavity end(under 11m). Easy to design, production and economic feasibility are the advantage of the designed slope in this study. However, the slope design may unsuitable when the sequences of the accidents did not satisfy the assumptions as mentioned. Despite of those disadvantages, the slope will show a great performance to mitigate the severe accident. As mentioned in assumption, the corium releasing time property was

Severe Accident Mitigation by using Core Catcher applicable for Korea standard nuclear power plant

International Nuclear Information System (INIS)

Park, Hae Kyun; Kim, Sang Nyung

2013-01-01

Nuclear power plants have been designed and operated in order to prevent severe accident because of their risk that contains tremendous radioactive materials that are potentially hazardous. Moreover, the government requested the nuclear industry to implement a severe accident management strategy for existing reactors to mitigate the risk of potential severe accidents. However, Korea standard nuclear power plant(APR-1400 and OPR-1000) are much more vulnerable for severe accident management than that of developed countries. Due to the design feature of reactor cavity in Korea standard nuclear power plant, inequable and serious Molten Core-Concrete Interaction(MCCI) may cause considerable safety problem to the reactor containment liner. At worst, it brings the release of radioactive materials to the environment. This accident applies to the fourth level of defense in depth(IAEA 1996), 'severe accident'. This study proposes and designs the 'slope' to secure reactor containment liner integrity when the corium spreads out from the destroyed reactor vessel to the reactor cavity due to the core melting accident. For this, make the initial corium distribution evenly exploit the 'slope' on the basis of the study of Ex-vessel corium behavior to prevent inequable and serious MCCI, in order to mitigate severe accident. The viscosity has a dominant position in the calculation. According to the result, the spread out distance on the slope is 10.7146841m, considering the rough surface of the concrete(slope) and margin of reactor cavity end(under 11m). Easy to design, production and economic feasibility are the advantage of the designed slope in this study. However, the slope design may unsuitable when the sequences of the accidents did not satisfy the assumptions as mentioned. Despite of those disadvantages, the slope will show a great performance to mitigate the severe accident. As mentioned in assumption, the corium releasing time property was conservatively calculated

An improved one-and-a-half group BWR core simulator for a new-generation core management system

International Nuclear Information System (INIS)

Iwamoto, Tatsuya; Yamamoto, Munenari

2000-01-01

An improved one-and-a-half group core simulator method for a next-generation BWR core management system is presented. In the improved method, intranodal spectral index (thermal to fast flux ratio) is expanded with analytic solutions to the diffusion equation, and the nodal power density and the interface net current are calculated, taking the intranodal flux shape into consideration. A unique method was developed for assembly heterogeneity correction. Thus eliminating the insufficiencies of the conventional one-and-a-half group method, we can have accurate power distributions as well as local peaking factors for cores having large spectral mismatch between fuel assemblies. The historical effects of spectral mismatch are also considered in both nodal power and local peaking calculations. Although reflectors are not solved explicitly, there is essentially no need for core dependent adjustable parameters, since boundary conditions are derived in the same manner as in the interior nodes. Calculation time for nodal solutions is comparable to that for the conventional method, and is less than 1/10 of a few-group nodal simulator. Verifications of the present method were made by comparing the results with those obtained by heterogeneous fine-mesh multi-group core depletion calculations, and the accuracy was shown to be fairly good. (author)

Reactor power measuring device

International Nuclear Information System (INIS)

Izumi, Mikio; Sano, Yuji; Seki, Eiji; Yoshida, Toshifumi; Ito, Toshiaki.

1993-01-01

The present invention provides a self-powered long detector having a sensitivity over the entire length of a reactor core as an entire control rod withdrawal range of a BWR type reactor, and a reactor power measuring device using a gamma ray thermometer which scarcely causes sensitivity degradation. That is, a hollow protection pipe is disposed passing through the reactor core from the outside of a reactor pressure vessel. The self-powered long detectors and the gamma ray thermometers are inserted and installed in the protection pipe. An average reactor power in an axial direction of the reactor relative to a certain position in the horizontal cross section of the reactor core is determined based on the power of the self-powered long detector over the entire length of the reactor core. Since the response of the self-powered detector relative to a local power change is rapid, the output is used as an input signal to a safety protection device of the reactor core. Further, a gamma ray thermometer secured in the reactor and having scarce sensitivity degradation is used instead of an incore travelling neutron monitor used for relative calibration of an existent neutron monitor secured in the reactor. (I.S.)

New Sodium Cooled Long-Life Cores with Axially Multi-Driver Regions

International Nuclear Information System (INIS)

Hyun, Hae Ri; Hong, Ser Gi

2014-01-01

In this concept of long-life core (they are sometimes called B-B (Breed and Burn)), tall blanket is placed above the relatively short driver fuel. In the initial stage of burning, the power by fission is mostly generated in the driver region and it moves into the blanket region. The power and flux distributions that are highly peaked in the axial direction propagates slowly from the driver into the blanket region. This concept of long-life core fully utilizes the breeding of blanket in the fast spectra and it can achieve very high burnup of fuel. In this work, we introduce new sodium cooled longlife cores rating 600MWe (1800MWt). In these cores, the driver regions are heterogeneously placed into blanket region so as to achieve stabilized and less peaked axial power distribution as depletion proceeds. At present, our study is focused on only two axial driver regions but this concept can be easily extended onto the multi-driver region concept. The cores designed in this paper have two axial driver regions so as to have stabilized and less peaked axial power distributions as depletion proceeds. The results of the core design and analyses show that the cores have very long-lives longer than -49EFPYs and high discharge burnup higher than 200GWD/kg. Additionally, we considered a long-life core having no blanket. As expected, it was shown that these cores have stabilized and less peaked axial power distribution as the fuel depletes. However, the study shows that the cores having two driver regions still show high initial peaking of the axial power distributions and the core can be optimized by changing the driver fuel height

Power distribution monitor in a nuclear reactor

International Nuclear Information System (INIS)

Uematsu, Hitoshi

1983-01-01

Purpose: To enable accurate monitoring for the reactor power distribution within a short time in a case where abnormality occurs in in-core neutron monitors or in a case where the reactor core state changes after the calibration for the neutron monitors. Constitution: The power distribution monitor comprises a power distribution calculator adapted to be inputted counted values from a reactor core present state data instruments and calculate the neutron flux distribution in the reactor core and the power distribution based on previously incorporated physical models, an RCF calculator adapted to be inputted with the counted values from the in-core neutron monitors and the neutron flux distribution and the power distribution calculated in the power distribution calculator and compensate the counted errors included in the counted values form the in-core neutron monitors and the calculation errors included in the power distribution calculated in the power distribution calculator to thereby calculate the power distribution within the reactor core, and an input/output device for the input of the data required for said power distribution calculator and the display for the calculation result calculated in the RCF calculator. (Ikeda, J.)

Optimization of fuel core loading pattern design in a VVER nuclear power reactors using Particle Swarm Optimization (PSO)

International Nuclear Information System (INIS)

Babazadeh, Davood; Boroushaki, Mehrdad; Lucas, Caro

2009-01-01

The two main goals in core fuel loading pattern design optimization are maximizing the core effective multiplication factor (K eff ) in order to extract the maximum energy, and keeping the local power peaking factor (P q ) lower than a predetermined value to maintain fuel integrity. In this research, a new strategy based on Particle Swarm Optimization (PSO) algorithm has been developed to optimize the fuel core loading pattern in a typical VVER. The PSO algorithm presents a simple social model by inspiration from bird collective behavior in finding food. A modified version of PSO algorithm for discrete variables has been developed and implemented successfully for the multi-objective optimization of fuel loading pattern design with constraints of keeping P q lower than a predetermined value and maximizing K eff . This strategy has been accomplished using WIMSD and CITATION calculation codes. Simulation results show that this algorithm can help in the acquisition of a new pattern without contravention of the constraints.

Evaluation of CANDU6 PCR (power coefficient of reactivity) with a 3-D whole-core Monte Carlo Analysis

International Nuclear Information System (INIS)

Motalab, Mohammad Abdul; Kim, Woosong; Kim, Yonghee

2015-01-01

Highlights: • The PCR of the CANDU6 reactor is slightly negative at low power, e.g. <80% P. • Doppler broadening of scattering resonances improves noticeably the FTC and make the PCR more negative or less positive in CANDU6. • The elevated inlet coolant condition can worsen significantly the PCR of CANDU6. • Improved design tools are needed for the safety evaluation of CANDU6 reactor. - Abstract: The power coefficient of reactivity (PCR) is a very important parameter for inherent safety and stability of nuclear reactors. The combined effect of a relatively less negative fuel temperature coefficient and a positive coolant temperature coefficient make the CANDU6 (CANada Deuterium Uranium) PCR very close to zero. In the original CANDU6 design, the PCR was calculated to be clearly negative. However, the latest physics design tools predict that the PCR is slightly positive for a wide operational range of reactor power. It is upon this contradictory observation that the CANDU6 PCR is re-evaluated in this work. In our previous study, the CANDU6 PCR was evaluated through a standard lattice analysis at mid-burnup and was found to be negative at low power. In this paper, the study was extended to a detailed 3-D CANDU6 whole-core model using the Monte Carlo code Serpent2. The Doppler broadening rejection correction (DBRC) method was implemented in the Serpent2 code in order to take into account thermal motion of the heavy uranium nucleus in the neutron-U scattering reactions. Time-average equilibrium core was considered for the evaluation of the representative PCR of CANDU6. Two thermal hydraulic models were considered in this work: one at design condition and the other at operating condition. Bundle-wise distributions of the coolant properties are modeled and the bundle-wise fuel temperature is also considered in this study. The evaluated nuclear data library ENDF/B-VII.0 was used throughout this Serpent2 evaluation. In these Monte Carlo calculations, a large number

Analysis of criticality safety of coupled fast-thermal core 'HERBE'

International Nuclear Information System (INIS)

Pesic, M.

1991-01-01

Power excursion during possible fast core flooding is analyzed as serious accident. Model gives short filling time of fast zone with moderator after break of fast core tank. Reactivity increase is determined by computer codes and verified in specific experiments. Measurements of safety rods drop time and reactivity worth are performed. Coupled core kinetics parameters are determined according to model of Avery. Power excursion study, depending on power level threshold and safety instrumentation response time is performed. It was shown that safety system can shut-down reactor safely even in case of highly set power thresholds and partially failure of safety chain. (author)

Sensitivity Analysis of Core Damage from Reactor Coolant Pump Seal Leakage during Extended Loss of All AC Power

Energy Technology Data Exchange (ETDEWEB)

Park, Da Hee; Kim, Min Gi; Lee, Kyung Jin; Hwang, Su hyun; Lee, Byung Chul [FNC Technology Co. Ltd., Yongin (Korea, Republic of); Yoon, Duk Joo; Lee, Seung Chan [Korea Hydro and Nuclear Power Co. Ltd., Daejeon (Korea, Republic of)

2015-10-15

In this study, in order to comprehend the Fukushima accident, the sensitivity analysis was performed to analyze the behavior of Reactor Coolant System (RCS) during ELAP using the RELAP5/MOD3.3 code. The Fukushima accident was caused by tsunami resulted in Station Black Out (SBO) followed by the reactor core melt-down and release of radioactive materials. After the accident, the equipment and strategies for the Extended Loss of All AC Power (ELAP) were recommended strongly. In this analysis, sensitivity studies for the RCP seal failure of the OPR1000 type NPP were performed by using RELAP5/MOD3.3 code. Six cases with different leakage rate of RCP seal were studied for ELAP with operator action or not. The main findings are summarized as follows: (1) Without the operator action, the core uncovery time is determined by the leakage rate of RCP seal. When the leakage rate per RCP seal are 5 gpm, 50 gpm, and 300 gpm respectively, the core uncovery time are 1.62 hr, 1.58 hr, and 1.29 hr respectively. Namely, If the leakage rate of RCP seal was much bigger, the uncover time of core would be shorter. (2) In case that the cooling by SG secondary side was performed using the TDAFP and SG ADV, the core uncovery time was significantly extended.

Development of coring, consolidating, subterrene penetrators

International Nuclear Information System (INIS)

Murphy, H.D.; Neudecker, J.W.; Cort, G.E.; Turner, W.C.; McFarland, R.D.; Griggs, J.E.

1976-02-01

Coring penetrators offer two advantages over full face-melting penetrators, i.e., formation of larger boreholes with no increase in power and the production of glass-lined, structurally undisturbed cores which can be recovered with conventional core-retrieval systems. These cores are of significant value in geological exploratory drilling programs. The initial design details and fabrication features of a 114-mm-diam coring penetrator are discussed; significant factors for design optimization are also presented. Results of laboratory testing are reported and compared with performance predictions, and an initial field trial is described

Silicon Nanophotonics for Many-Core On-Chip Networks

Science.gov (United States)

Mohamed, Moustafa

Number of cores in many-core architectures are scaling to unprecedented levels requiring ever increasing communication capacity. Traditionally, architects follow the path of higher throughput at the expense of latency. This trend has evolved into being problematic for performance in many-core architectures. Moreover, the trends of power consumption is increasing with system scaling mandating nontraditional solutions. Nanophotonics can address these problems, offering benefits in the three frontiers of many-core processor design: Latency, bandwidth, and power. Nanophotonics leverage circuit-switching flow control allowing low latency; in addition, the power consumption of optical links is significantly lower compared to their electrical counterparts at intermediate and long links. Finally, through wave division multiplexing, we can keep the high bandwidth trends without sacrificing the throughput. This thesis focuses on realizing nanophotonics for communication in many-core architectures at different design levels considering reliability challenges that our fabrication and measurements reveal. First, we study how to design on-chip networks for low latency, low power, and high bandwidth by exploiting the full potential of nanophotonics. The design process considers device level limitations and capabilities on one hand, and system level demands in terms of power and performance on the other hand. The design involves the choice of devices, designing the optical link, the topology, the arbitration technique, and the routing mechanism. Next, we address the problem of reliability in on-chip networks. Reliability not only degrades performance but can block communication. Hence, we propose a reliability-aware design flow and present a reliability management technique based on this flow to address reliability in the system. In the proposed flow reliability is modeled and analyzed for at the device, architecture, and system level. Our reliability management technique is

Responses of platinum, vanadium and cobalt self-powered flux detectors near simulated booster rods in a ZED-2 mockup of a Bruce reactor core

International Nuclear Information System (INIS)

French, P.M.; Shields, R.B.; Kroon, J.C.

1978-02-01

The static responses of Pt, V and Co self-powered detectors have been compared with copper-foil neutron activation profiles in reference and perturbed Bruce reactor core mockups assembled in the ZED-2 test reactor at Chalk River Nuclear Laboratories. The results indicate that the normalized response of each self-powered detector is an accurate measure of the thermal-neutron flux at locations greater than one lattice pitch from either a booster rod or the core boundary. They indicate that, in the Bruce booster/detector configuration, the normalized static Pt response overestimates the neutron flux by less than 3.5% upon full booster-rod insertion. (author)

Analysis with SCDAP/RELAP5 of reflooding of an overheated core in Forsmark 3 BWR after loss of electric power

International Nuclear Information System (INIS)

Nilsson, Lars.

1993-01-01

In foregoing SIK-2.2 project two severe accident sequences for Forsmark 3 comprising loss of electric power (Total Blackout, TB) without recovery actions (e.g. emergency cooling) were analysed with SCDAP/RELAP5. Code version MOD2.5/V3f was applied and a single core channel input model was used. Present analysis was done with the same initiating events as in SIK-2.2, but assuming recovery of auxiliary feed water and/or emergency core cooling systems to take place after heat-up of the core, in compliance with the plans of the SIK-2.3 project. A new test version of the SCDAP/RELAP5 code, version MOD3/V7af, was used here. The geometrical input model was extended from having one, into five parallel core zones, still with ten axial core sub volumes. Calculations were performed for three TB cases, one without cooling recovery, and two with injection of cold water at a rate of 45 kg/s, beginning when the maximum cladding temperature has reached 1522 K. The results show that a considerably more efficient cooling was obtained spraying the water to the top of the core, compared to the case where the cooling water was introduced into the downcomer, leading to bottom reflooding

In-core fuel management activities in China

International Nuclear Information System (INIS)

Ruan Keqiang; Chen Renji; Hu Chuanwen

1990-01-01

The development of nuclear power in China has reached such a stage that PWR in-core fuel management becomes an urgent problem. At present the main effort is concentrated on solving the Qinshan nuclear power plant and Daya Bay nuclear power plant fuel management problems. For the Qinshan PWR (300 MWe) two packages of in-core fuel management code were developed, one with simplified nodal diffusion method and the other uses advanced Green's function nodal method. Both were used in the PWR core design. With the help of the two code packages first two cycles of the Qinshan PWR core burn-up were calculated. Besides, several research works are under way in the following areas: improvement of the nodal diffusion method and other coarse mesh method in terms of computing speed and accuracy; backward diffusion technique for fuel management application; optimization technique in the fuel loading pattern searching. As for the Daya Bay PWR plant (twin 900 MWe unit), the problem about using what kind of code package for in-core fuel management is still under discussion. In principle the above mentioned code packages are also applicable to it. Besides PWR, in-core fuel management research works are also under way for research reactors, for example, heavy water research reactor and high flux research reactor in some institutes in China. China also takes active participation in international in-core fuel management activities. (author). 19 refs

Protection set-points lines for the reactor core and considerations about power distribution and peak factors

International Nuclear Information System (INIS)

Furieri, E.B.

1981-01-01

In order to assure the reactor core integrity during the slow operational transients (power excursion above the nominal value and the high coolant temperature), the formation of a steam film (DNB-Departure from Nucleate Boiling) in the control rods must be avoided. The protection set points lines presents the points where DNBR (relation between critical heat flux-q sub(DNB) and the local heat flux-q' sub(local) is equal to 1.30, corrected by peak factors and uncertainty in function of ΔTr and T sub(R), respectively coolant elevation and medium coolant temperature in reactor pressure vessel. The curve set-points were determined using a new version of COBRA-IIIF (CUPRO) computer code, implemented with new subroutines and linearized convergence scheme. Pratical results for Angra-1 core were obtained and its were compared with the results from the fabricator. (E.G.) [pt

Core Design Concept and Core Structural Material Development for a Prototype SFR

International Nuclear Information System (INIS)

Chang, Jinwook

2013-01-01

• Core design Concept: – Initial core is Uranium metal fueled core, then it will evolve into TRU core; – Tight pressure drop constraint lowers power density; – Trade-off studies with relaxed pressure drop constraint (~0.4MPa) are on-going; – Major feature will be finalized this year. • KAERI is developing advanced cladding for high burnup fuel in Ptototype SFR: – Advanced cladding materials are now developing, which shows superior high temperature mechanical property to the conventional material; – Processing technologies related to tube making process are now developed to enhance high temperature mechanical propertyl – Preliminary HT9 cladding tube was manufactured and out-of pile mechanical properties were evaluated. Advanced cladding tube is now being developed and being prepared for irradiation test

Thermal-hydraulic analysis of the OSURR pool for power upgrade with natural convection core cooling

International Nuclear Information System (INIS)

Ha, J.J.; Aldemir, T.

1988-01-01

Natural convection mode core cooling will be maintained in the LEU conversion/power upgrade of The Ohio State University Research Reactor (OSURR) to 250-500 kW. The pool water will be cooled by a water-glycol-air and a water-water heat exchanger. A plume disperser will be installed in the pool to minimize evaporation from the pool top and to maintain the dose rate due to N-16 activity within allowable levels. The minimization of the pool heat removal system operation costs necessitates maximizing the inlet temperature to the water-glycol-air heat exchanger. For the maximization process, the change in the pool temperature and velocity fields have to be investigated as a function of: location and orientation of the heat removal system components and the plume disperser in the pool; mass flow rate through the plume disperser. The velocity and temperature fields in the pool are determined using COMMIX-1A. The computational system model accounts for the presence of all the pool components (i.e. core, thermal column, beam ports, ion chamber, guide tubes, rabbit, neutron source etc.). The results show that: (1) Both the heat removal system inlet point and the plume disperser have to be located close to the top of the core. (2) Using a disperser system consisting of several pipes may be more feasible than a single unit. (3) For high disperser flow, the disperser jet has to be almost parallel to the top of the core to prevent flow reversal in coolant channels. (4) More than one disperser system may be necessary to create an inversion layer in the pool

Core design concepts for high performance light water reactors

International Nuclear Information System (INIS)

Schulenberg, T.; Starflinger, J.

2007-01-01

Light water reactors operated under supercritical pressure conditions have been selected as one of the promising future reactor concepts to be studied by the Generation IV International Forum. Whereas the steam cycle of such reactors can be derived from modern fossil fired power plants, the reactor itself, and in particular the reactor core, still need to be developed. Different core design concepts shall be described here to outline the strategy. A first option for near future applications is a pressurized water reactor with 380 .deg. C core exit temperature, having a closed primary loop and achieving 2% pts. higher net efficiency and 24% higher specific turbine power than latest pressurized water reactors. More efficiency and turbine power can be gained from core exit temperatures around 500 .deg. C, which require a multi step heat up process in the core with intermediate coolant mixing, achieving up to 44% net efficiency. The paper summarizes different core and assembly design approaches which have been studied recently for such High Performance Light Water Reactors

Development of Pipeline Database and CAD Model for Selection of Core Security Zone in Nuclear Power Plants

International Nuclear Information System (INIS)

Choi, Seong Soo; Kwon, Tae Gyun; Baek, Hun Hyun; Kwon, Min Jin

2008-07-01

The objective of the project is to develop the pipeline database which can be used for selection of core security zones considering safety significance of pipes and to develop CAD model for 3-dimensional visualization of core security zones, for the purpose of minimizing damage and loss, enforcing security and protection on important facilities, and improving plant design preparing against emergency situations such as physical terrors in nuclear power plants. In this study, the pipeline database is developed for selection of core security zones considering safety significance of safety class 1 and 2 pipes. The database includes the information on 'pipe-room information-surrogate component' mapping, initiating events which may occur and accident mitigation functions which may be damaged by the pipe failure, and the drawing information related to 2,270 pipe segments of 30 systems. For the 3-dimensional visualization of core security zones, the CAD models on the containment building and the auxiliary building are developed using 3-D MAX tool and the demo program which can visualize the direct-X model converted from the 3-D MAX model is also developed. In addition to this, the coordinate information of all the buildings and their rooms is generated using AUTO CAD tool in order to be used as an input for 3-dimensional browsing of the VIP program

Some aspects of optimising the reactor core for a 600 MW(e) high temperature helium turbine power plant

Energy Technology Data Exchange (ETDEWEB)

Hansen, U; Presser, W

1972-04-24

For the HHT 600 MW(e) power plant a core design with Triagonal blocks containing 24 channels with directly cooled fuel pins was considered. The design was found to require a low HM loading in the fuel zone to achieve favourable economic merits. For low HM densities a strong incentive exists to aim for burn-ups between 80 and 100 GWd/t. At the present an average discharge irradiation of 80 GWd/t was thought feasible and a reference design with a HM density of 0.6 g/cm {sup 3} in both core zones was chosen. The optimisation is not likely to be upset by local hot channel effects as a special investigation into the influence of safety margins found no changes in fuel cycle economics.

PANTHER solution to the NEA-NSC 3-D PWR core transient benchmark. Uncontrolled withdrawal of control rods at zero power

Energy Technology Data Exchange (ETDEWEB)

Kuijper, J.C.

1994-10-01

This report contains the results of PANTHER calculations for the ``NEA-NSC 3-D PWR Core Transient Benchmark: Uncontrolled Withdrawal of Control Rods at Zero Power``. PANTHER was able to model the benchmark problems without modifications to the code. All the calculations were performed in 3-D. (orig.).

Final PANTHER solution to the NEA-NSC3-DPWR core transient benchmark. Uncontrolled withdrawal of control rods at zero power

International Nuclear Information System (INIS)

Kuijper, J.C.

1996-10-01

This report contains the final results of PANTHER calculations for the 'NEA-NSC 3-D PWR Core Transient Benchmark: Uncontrolled Withdrawal of Control Rods at Zero Power'. PANTHER was able to model the benchmark problems without modifications to the code. All the calculations were performed in 3-D. (orig.)

PANTHER solution to the NEA-NSC 3-D PWR core transient benchmark. Uncontrolled withdrawal of control rods at zero power

International Nuclear Information System (INIS)

Kuijper, J.C.

1994-10-01

This report contains the results of PANTHER calculations for the ''NEA-NSC 3-D PWR Core Transient Benchmark: Uncontrolled Withdrawal of Control Rods at Zero Power''. PANTHER was able to model the benchmark problems without modifications to the code. All the calculations were performed in 3-D. (orig.)

Final PANTHER solution to the NEA-NSC3-DPWR core transient benchmark. Uncontrolled withdrawal of control rods at zero power

Energy Technology Data Exchange (ETDEWEB)

Kuijper, J.C.

1996-10-01

This report contains the final results of PANTHER calculations for the `NEA-NSC 3-D PWR Core Transient Benchmark: Uncontrolled Withdrawal of Control Rods at Zero Power`. PANTHER was able to model the benchmark problems without modifications to the code. All the calculations were performed in 3-D. (orig.).

Core calculational techniques and procedures

International Nuclear Information System (INIS)

Romano, J.J.

1977-10-01

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

Surry Power Station, Units 1 and 2. Annual operating report for 1976, section 2: core performance and startup physics test reports

International Nuclear Information System (INIS)

1977-01-01

The analyses of four core performance indicators including burnup distribution, reactivity depletion, power distribution, and primary coolant activity are discussed. Information is presented concerning fuel densification monitoring; start-up physics testing; and James River temperature, salinity, fish, biota, and plankton entrainment monitoring

Characteristic test of initial HTTR core

International Nuclear Information System (INIS)

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

2004-01-01

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

TerraPower, Bill Gates' reactor

International Nuclear Information System (INIS)

Guidez, J.

2016-01-01

TerraPower is a traveling wave reactor, it means that the reactor gradually converts non fissile material into the fuel it needs and the active part of the core progressively moves through the core leaving spent fuel behind. The last design of the TerraPower shows that it will use depleted uranium as fuel and that its core will need reloading every 10 years. Re-arrangement of the nuclear fuel will have to be made every 18 months to keep the core reactive. Metallic nuclear fuels will be used as they allow the highest breeding rates. It appears that apart from the very specific configuration of the core, the TerraPower is a reactor very similar to sodium-cooled fast reactors. Neutron transport inside traveling wave reactor core is complex and simulations show that the piling-up of fission product tends to kill the chain reaction and a continuous neutron addition may be necessary to keep the reactor going. A large part of the TerraPower feasibility studies concerns neutron transport inside its core. (A.C.)

Neutronic calculations of PARR-1 cores using LEU silicide fuel

International Nuclear Information System (INIS)

Arshad, M.; Bakhtyar, S.; Hayat, T.; Salahuddin, A.

1991-08-01

Detailed neutronic calculations have been carried out for different PARR-1 cores utilizing low enriched uranium (LEU) silicide fuel and operating at an upgraded power of 9 MW. The calculations include the search for critical loadings in open and stall ends of the pool, neutronic analysis of the first full equilibrium core and calculations cores. The burnup study of inventory have also been carried out. Further, the reactivity coefficients of the first full power operation core are evaluated for use in the accident analysis. 14 figs. (author)

In-core monitoring detectors

International Nuclear Information System (INIS)

Mitelman, M.G.

2001-01-01

The main task of in-core monitoring consists in securing observability of the reactor installation in all possible operation modes (normal, transient, accident and post-accident). Operation safety at acceptable cost can be achieved by optimized measurement errors. The range of sensors applied as in-core detectors for operative measurements in the industry is very limited in number. Among them might be cited self powered neutron detectors (SPND) and thermocouples. Sensors are incorporated in the in-core detectors assemblies (SVRD). The presentation makes an effort to touch upon the problems of assuring and increasing quality of in-core on-line measurements. So we do not consider systems using movable detectors, as the latter do not assure on-line measurements. (Authors)

Design study on metal fuel FBR cores

International Nuclear Information System (INIS)

Yokoo, T.; Tanaka, Y.; Ogata, T.

1991-01-01

A design approach for metal fuel FBR core to maintain fuel integrity during transient events by limiting eutectic/liquid phase formation is proposed based on the current status of metallic fuel development. Its impact as the limitation on the core outlet temperature is assessed through its application to two of CRIEPI's core concepts, high linear power 1000 MWe homogeneous design and medium linear power 300 MWe radially heterogeneous design. SESAME/SALT code is used in this study to analyze steady state and transient fuel behavior. SE2-FA code is developed based on SUPERENERGY-2 and used to analyze core thermal-hydraulics with uncertainties. As the result, the core outlet temperatures of both designs are found to be limited to ≤500degC if it is required to prevent eutectic/liquid phase formation during operational transients in order to guarantee the fuel integrity. Additional assessment is made assuming an advanced limiting condition that allows small liquid phase formation based on the liquid phase penetration rate derived from existing experimental results. The result indicates possibility of raising core outlet temperature to ∼ 530degC. Also, it is found that core design technology improvements such as hot spot factors reduction can contribute to the core outlet temperature extension by 10 ∼ 20degC. (author)

Monitoring device for the reactor power distribution

International Nuclear Information System (INIS)

Uematsu, Hitoshi; Tsuiki, Makoto

1982-01-01

Purpose: To enable accurate monitoring for the power distribution in a short time, as well as independent detection for in-core neutron flux detectors in abnormal operation due to failures or like other causes to thereby surely provide reliable substitute values. Constitution: Counted values are inputted from a reactor core present status data detector by a power distribution calculation device to calculate the in-core neutron flux density and the power distribution based on previously stored physical models. While on the other hand, counted value from the in-core neutron detectors and the neutron flux distribution and the power distribution calculated from the power distribution calculation device are inputted from a BCF calculation device to compensate the counting errors incorporated in the counted value from the in-core neutron flux detectors and the calculation errors incorporated in the power distribution calculated in the power distribution calculation device respectively and thereby calculate the power distribution in the reactor core. Further, necessary data are inputted to the power distribution calculation device by an input/output device and the results calculated in the BCF calculation device are displayed. (Aizawa, K.)

Hybrid Magnetics and Power Applications

DEFF Research Database (Denmark)

Mo, Wai Keung; Paasch, Kasper

2017-01-01

A hybrid magnetic approach, merging two different magnetic core properites such as ferrite and iron powder cores, is an effective solution for power converter applications. It can offer similar magnetic properties to that of magnetic powder cores but showing less copper loss than powder cores....... In order to prevent ferrite core saturation, placing an effective air gap within the ferrite core is a key method to obtain optimum hybrid magnetic performance. Furthermore, a relatively large inductance at low loading current is an excellent way to minimze power loss in order to achieve high efficiency...

Nuclear Power Reactor Core Melt Accidents. Current State of Knowledge

International Nuclear Information System (INIS)

Bentaib, Ahmed; Bonneville, Herve; Clement, Bernard; Cranga, Michel; Fichot, Florian; Koundy, Vincent; Meignen, Renaud; Corenwinder, Francois; Leteinturier, Denis; Monroig, Frederique; Nahas, Georges; Pichereau, Frederique; Van-Dorsselaere, Jean-Pierre; Cenerino, Gerard; Jacquemain, Didier; Raimond, Emmanuel; Ducros, Gerard; Journeau, Christophe; Magallon, Daniel; Seiler, Jean-Marie; Tourniaire, Bruno

2013-01-01

For over thirty years, IPSN and subsequently IRSN has played a major international role in the field of nuclear power reactor core melt accidents through the undertaking of important experimental programmes (the most significant being the Phebus- FP programme), the development of validated simulation tools (the ASTEC code that is today the leading European tool for modelling severe accidents), and the coordination of the SARNET (Severe Accident Research Network) international network of excellence. These accidents are described as 'severe accidents' because they can lead to radioactive releases outside the plant concerned, with serious consequences for the general public and for the environment. This book compiles the sum of the knowledge acquired on this subject and summarises the lessons that have been learnt from severe accidents around the world for the prevention and reduction of the consequences of such accidents, without addressing those from the Fukushima accident, where knowledge of events is still evolving. The knowledge accumulated by the Institute on these subjects enabled it to play an active role in informing public authorities, the media and the public when this accident occurred, and continues to do so to this day

Core reset system design for linear induction accelerator

International Nuclear Information System (INIS)

Durga Praveen Kumar, D.; Mitra, S.; Sharma, Archana; Nagesh, K.V.; Chakravarthy, D.P.

2006-01-01

A repetitive pulsed power system based Linear Induction Accelerator (LIA-200) is being developed at BARC to get an electron beam of 200keV, 5kA, 50ns, 10-100 Hz. Amorphous core is the heart of these accelerators. It serves various functions in different subsystems viz. pulse power modulator, pulse transformer, magnetic switches and induction cavities. One of the factors that make the magnetic components compact is utilization of the total flux swing available in the core. In the present system, magnetic switches, pulse transformers, and induction cavity are designed to avail the full flux swing available in the core. For achieving this objective, flux density in the core has to be kept at the reverse saturation, before the main pulse is applied. The electrical circuit which makes it possible is called the core reset system. In this paper the details of core reset system designed for LIA-200 are described. (author)

Operation method and operation control device for emergency core cooling system

Energy Technology Data Exchange (ETDEWEB)

Kinoshita, Shoichiro; Takahashi, Toshiyuki; Fujii, Tadashi [Hitachi Ltd., Tokyo (Japan); Mizutani, Akira

1996-05-07

The present invention provides a method of reducing continuous load capacity of an emergency cooling system of a BWR type reactor and a device reducing a rated capacity of an emergency power source facility. Namely, the emergency core cooling system comprises a first cooling system having a plurality of power source systems based on a plurality of emergency power sources and a second cooling system having a remaining heat removing function. In this case, when the first cooling system is operated the manual starting under a predetermined condition that an external power source loss event should occur, a power source division different from the first cooling system shares the operation to operate the secondary cooling system simultaneously. Further, the first cooling system is constituted as a high pressure reactor core water injection system and the second cooling system is constituted as a remaining heat removing system. With such a constitution, a high pressure reactor core water injection system for manual starting and a remaining heat removing system of different power source division can be operated simultaneously before automatic operation of the emergency core cooling system upon loss of external power source of a nuclear power plant. (I.S.)

Development of Core Design Technology for LMR

Energy Technology Data Exchange (ETDEWEB)

Kim, Yeong Il; Hong, S. G.; Jang, J. W. (and others)

2007-06-15

This report describes the contents of core design technology and computer code system development performed during 2005 and 2006 on the objects of nuclear proliferation resistant core and nuclear fuel basic key technology development security. Also, it is including the future application plans for the results and the developed methodology, important information and the materials acquired in this period. Two core designs with single enrichment were considered for the KALIMER-600 during the first year : 1) the first core uses the non-fuel rods such as B4C, ZrH1.8, and dummy rods, 2) the core using different cladding thickness for each core region (inner, middle, and outer cores) without non-fuel rods to flatten the power distribution. In particular, the latter design was intended to simplify the fuel assembly design by eliminating the heterogeneity. It was found that the proposed design satisfy all of the Gen IV SFR design goals on the cycle length longer than 18 EFPM, fuel discharge burnup larger than 80GWd/t, sodium void worth, conversion ratio, reactivity burnup swing and so on. For this object reactor, the structure integrity outside of reactor is confirmed for the radiation exposure during the plant life according to the result of shielding design and evaluation. The transmutation capability and the core characteristics of sodium cooled fast reactor was also evaluated according to the change of MA amount. The reactivity coefficients for the BN-600 reactor with MA fueled are calculated and the results are compared and evaluated with other participants results. Even though the discrepancies between the results of participants are somewhat large but the K-CORE results are close to the average within a standard deviation. To have the capability of 3-dimensional core dynamic analysis such as analyzing power distribution and reactivity variations according to the asymmetric insertion/withdrawal of control rods, the calculation module for core dynamic parameters was

Fort St. Vrain core performance

International Nuclear Information System (INIS)

McEachern, D.W.; Brown, J.R.; Heller, R.A.; Franek, W.J.

1977-07-01

The Fort St. Vrain High Temperature Gas Cooled Reactor core performance has been evaluated during the startup testing phase of the reactor operation. The reactor is graphite moderated, helium cooled, and uses coated particle fuel and on-line flow control to each of the 37 refueling regions. Principal objectives of startup testing were to determine: core and control system reactivity, radial power distribution, flow control capability, and initial fission product release. Information from the core demonstrates that Technical Specifications are being met, performance of the core and fuel is as expected, flow and reactivity control are predictable and simple for the operator to carry out

Nanostructured core-shell electrode materials for electrochemical capacitors

Science.gov (United States)

Jiang, Long-bo; Yuan, Xing-zhong; Liang, Jie; Zhang, Jin; Wang, Hou; Zeng, Guang-ming

2016-11-01

Core-shell nanostructure represents a unique system for applications in electrochemical energy storage devices. Owing to the unique characteristics featuring high power delivery and long-term cycling stability, electrochemical capacitors (ECs) have emerged as one of the most attractive electrochemical storage systems since they can complement or even replace batteries in the energy storage field, especially when high power delivery or uptake is needed. This review aims to summarize recent progress on core-shell nanostructures for advanced supercapacitor applications in view of their hierarchical architecture which not only create the desired hierarchical porous channels, but also possess higher electrical conductivity and better structural mechanical stability. The core-shell nanostructures include carbon/carbon, carbon/metal oxide, carbon/conducting polymer, metal oxide/metal oxide, metal oxide/conducting polymer, conducting polymer/conducting polymer, and even more complex ternary core-shell nanoparticles. The preparation strategies, electrochemical performances, and structural stabilities of core-shell materials for ECs are summarized. The relationship between core-shell nanostructure and electrochemical performance is discussed in detail. In addition, the challenges and new trends in core-shell nanomaterials development have also been proposed.

Advanced core monitoring technology for WWER reactors

International Nuclear Information System (INIS)

Nguyen, T.Q.; Casadei, A.L.; Doshi, P.K.

1993-01-01

The Westinghouse BEACON online monitoring system has been developed to provide continuous core monitoring and operational support for pressurized water reactor using movable detectors (fission chamber) and core thermocouples. The basic BEACON core monitoring methodology is described. Traditional WWER reactors use rhodium fixed in-core detectors as the means to provide detailed core power distribution for surveillance purposes. An adapted version of the BEACON advanced core monitoring and support system is described which seems to be, due to the different demand/response requirements, the optimal solution (for routine surveillance and anomaly detection) for WWER reactors with existing fixed in-core detectors. (Z.S.) 4 refs

Power generator in BWR type reactors

International Nuclear Information System (INIS)

Yoshida, Kenji.

1984-01-01

Purpose: To enable to perform stable and dynamic conditioning operation for nuclear fuels in BWR type reactors. Constitution: The conditioning operation for the nuclear fuels is performed by varying the reactor core thermal power in a predetermined pattern by changing the predetermined power changing pattern of generator power, the rising rate of the reactor core thermal power and the upper limit for the rising power of the reactor core thermal power are calculated and the power pattern for the generator is corrected by a power conditioning device such that the upper limit for the thermal power rising rate and the upper limit for the thermal power rising rate are at the predetermined levels. Thus, when the relation between the reactor core thermal power and the generator electrical power is fluctuated, the fluctuation is detected based on the variation in the thermal power rising rate and the limit value for the thermal power rising rate, and the correction is made to the generator power changing pattern so that these values take the predetermined values to thereby perform the stable conditioning operation for the nuclear fuels. (Moriyama, K.)

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

International Nuclear Information System (INIS)

Ishida, Masayoshi; Kawada, Kenichi; Niwa, Hajime

2003-07-01

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

Core design aspects of SNR 2

International Nuclear Information System (INIS)

Wehmann, U.K.

1987-01-01

The paper describes in its first part the main characteristics of the core of the SNR 2 fast breeder reactor which is being planned within the European collaboration on fast breeder reactors. In the second part some core design aspects are discussed. The fuel element management with an inwards shuffling after each cycle is illustrated which offers advantages with respect to linear rating, steel damage and average discharge burnup. For this management, the full three-dimensional power and burnup history has been calculated and some typical results are presented. The shutdown requirements and the capabilities of the two shutdown systems of SNR 2 are discussed. The necessity for a reliable surveillance of the power distribution is demonstrated by the pronounced power tilts in case of the unintentional withdrawal of an absorber rod. Finally, a short review of the main nuclear design methods and their validation with help of the evaluation of experiments in zero power facilities and power reactors is given

Noise analysis of Forsmark 1 data to investigate BWR core local instability

International Nuclear Information System (INIS)

Oguma, R.

1998-04-01

BWR core local instability was experienced at Forsmark 1 (F1) during reactor operation in cycle 16. The event has been studied by applying noise analysis and stability calculations to get insight into the event as well as to identify the cause of local instability. The present report is concerned with noise analysis of data collected during start-up in cycle 17. The results of the current study indicates: The F1 core is quite stable in cycle 17. The max. decay ratio (DR) value of 0.37 was obtained from the stability evaluation of an APRM (average power range monitor) and LPRM (local power range monitor) signals measured at 66% (APRM) of reactor power and 4252 Kg/s (SA-HC) of core flow. Compared with the power profile in cycle 17 (as well as in reactor F2), the core in cycle 16 had an extreme power profile with high power and bottom-shifted axial peak in the core periphery esp. at the four quadrant corners. Such a profile decreases the stability margin in the region. It is a common observation that the DR obtained from APRM tends to be higher than that from LPRM if the global instability mechanism is dominant in the core, and vice versa. The comparison of global and local DR values should be an effective method for detecting local instability during the reactor operation. In order to detect the local instability it is important to evaluate the core stability with sufficient number of LPRMs so as to cover the whole core cross section together with APRMs

Distribution of radionuclides in a marine sediment core off the waterspout of the nuclear power plants in Daya Bay, northeastern South China Sea

International Nuclear Information System (INIS)

Zhou, Peng; Li, Dongmei; Li, Haitao; Fang, Hongda; Huang, Chuguang; Zhang, Yusheng; Zhang, Hongbiao; Zhao, Li; Zhou, Junjie; Wang, Hua; Yang, Jie

2015-01-01

A sediment core was collected and dated using 210 Pb ex dating method off the waterspout of nuclear power base of Daya Bay, northeastern South China Sea. The γ-emitting radionuclides were analyzed using HPGe γ spectrometry, gross alpha and beta radioactivity as well as other geochemical indicators were deliberated to assess the impact of nuclear power plants (NPP) operation and to study the past environment changes. It suggested that NPP provided no new radioactivity source to sediment based on the low specific activity of 137 Cs. Two broad peaks of TOC, TC and LOI accorded well with the commercial operations of Daya Bay NPP (1994.2 and 1994.5) and LNPP Phase I (2002.5 and 2003.3), implying that the mass input of cooling water from NPP may result into a substantial change in the ecological environment and Daya Bay has been severely impacted by human activities. - Graphical abstract: A sediment core was collected and dated using 210 Pb ex dating method, off the waterspout of nuclear power base of Daya Bay, northeastern South China Sea. The γ-emitting radionuclides analyzed using the HPGe γ spectrometry, gross alpha and beta radioactivity as well as other geochemical indicators were deliberated to study the past environment changes and assess the impact of nuclear power plants (NPP) operation. NPP provided no new radioactivity source to sediment, but the mass input of cooling water from nuclear power plants may result into a substantial change in the ecological environment and Daya Bay has been severely impacted by human activities. - Highlights: • A sediment core collected from Daya Bay, South China Sea was dated using 210 Pb ex method. • The γ-emitting radionuclides, gross α and β, TOC, TIC, TC, LOI were deliberated to assess the impact of nuclear power plants (NPP) operation. • The low activity of 137 Cs in sediment suggested NPP provided no new radioactivity source. • Two peaks of TOC, TC and LOI implied that the mass input of NPP cooling water may

State space modeling of reactor core in a pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Ashaari, A.; Ahmad, T.; M, Wan Munirah W. [Department of Mathematical Science, Faculty of Science, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Shamsuddin, Mustaffa [Institute of Ibnu Sina, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Abdullah, M. Adib [Swinburne University of Technology, Faculty of Engineering, Computing and Science, Jalan Simpang Tiga, 93350 Kuching, Sarawak (Malaysia)

2014-07-10

The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

Sensitivity of reactivity feedback due to core bowing in a metallic-fueled core

International Nuclear Information System (INIS)

Nakagawa, Masatoshi; Kawashima, Masatoshi; Endo, Hiroshi; Nishimura, Tomohiro

1991-01-01

A sensitivity study has been carried out on negative reactivity feedback caused by core bowing to assess the potential effectiveness of FBR passive safety features in regard to withstanding an anticipated transient without scram (ATWS). In the present study, an analysis has been carried to obtain the best material and geometrical conditions concerning the core restraint system out for several power to flow rates (P/F), up to 2.0 for a 300 MWe metallic-fueled core. From this study, it was clarified that the pad stiffness at an above core loading pads (ACLP) needs to be large enough to ensure negative reactivity feedback against ATWS. It was also clarified that there is an upper limit for the clearances between ducts at ACLP. A new concept, in regard to increasing the absolute value for negative reactivity feedback due to core bowing at ATWS, is proposed and discussed. (author)

Westinghouse power distribution monitoring experience at Duke Power's McGuire Unit 1

International Nuclear Information System (INIS)

Grobmyer, L.R.; Cash, M.T.; Kitlan, M.S.; Impink, A.J. Jr.

1987-01-01

In the evolution of the Westinghouse methodology of assuring safe core power distributions, emphasis was placed on analysis and not on continuous detailed core monitoring. Power distribution monitoring is currently achieved by periodic surveillances using the movable in-core detector system (MIDS) and by continuous observations of the two-section excore power range detectors. Control of the power distribution is regulated by limits on the indications from these systems, by limits on control rod insertion, and by operational constraints on the position indication systems. As more plants come on line and as more utilities take over the fuel design function for themselves, the desire for better core monitoring becomes evident. Also, the need and desire by the utilities to have more control over their operating margin has motivated the industry to offer and/or upgrade core monitoring systems. Westinghouse and Duke Power are participants in a joint development program to finalize the development of the core on-line surveillance monitoring and operations system (COSMOS). This final stage of development consists of prototype field trials at the McGuire Nuclear Plant. The purpose of the prototype program is to determine how well the design objectives are met and how to improve the system based on the operating experience at McGuire. Another purpose of this prototype program is to generate the necessary experience and information to develop a topical report for the US Nuclear Regulatory Commission to obtain a licensing basis for technical specification relaxation

United States Department of Energy's reactor core protection evaluation methodology for fires at RBMK and VVER nuclear power plants. Revision 1

International Nuclear Information System (INIS)

1997-06-01

This document provides operators of Soviet-designed RBMK (graphite moderated light water boiling water reactor) and VVER (pressurized light water reactor) nuclear power plants with a systematic Methodology to qualitatively evaluate plant response to fires and to identify remedies to protect the reactor core from fire-initiated damage

On the effective parallel programming of multi-core processors

NARCIS (Netherlands)

Varbanescu, A.L.

2010-01-01

Multi-core processors are considered now the only feasible alternative to the large single-core processors which have become limited by technological aspects such as power consumption and heat dissipation. However, due to their inherent parallel structure and their diversity, multi-cores are

FBR type reactor core

International Nuclear Information System (INIS)

Tamiya, Tadashi; Kawashima, Katsuyuki; Fujimura, Koji; Murakami, Tomoko.

1995-01-01

Neutron reflectors are disposed at the periphery of a reactor core fuel region and a blanket region, and a neutron shielding region is disposed at the periphery of them. The neutron reflector has a hollow duct structure having a sealed upper portion, a lower portion opened to cooling water, in which a gas and coolants separately sealed in the inside thereof. A driving pressure of a primary recycling pump is lowered upon reduction of coolant flow rate, then the liquid level of coolants in the neutron reflector is lowered due to imbalance between the driving pressure and a gas pressure, so that coolants having an effect as a reflector are eliminated from the outer circumference of the reactor core. Therefore, the amount of neutrons leaking from the reactor core is increased, and negative reactivity is charged to the reactor core. The negative reactivity of the neutron reflector is made greater than a power compensation reactivity. Since this enables reactor scram by using an inherent performance of the reactor core, the reactor core safety of an LMFBR-type reactor can be improved. (I.N.)

Toward full MOX core design

International Nuclear Information System (INIS)

Rouviere, G.; Guillet, J.L.; Bruna, G.B.; Pelet, J.

1999-01-01

This paper presents a selection of the main preliminary results of a study program sponsored by COGEMA and currently carried out by FRAMATOME. The objective of this study is to investigate the feasibility of full MOX core loading in a French 1300 MWe PWR, a recent and widespread standard nuclear power plant. The investigation includes core nuclear design, thermal hydraulic and systems aspects. (authors)

Fuel assembly and reactor core

International Nuclear Information System (INIS)

Yuchi, Yoko; Aoyama, Motoo; Haikawa, Katsumasa; Yamanaka, Akihiro; Koyama, Jun-ichi.

1996-01-01

In a fuel assembly of a BWR type reactor, a region substantially containing burnable poison is divided into an upper region and a lower region having different average concentrations of burnable poison along a transverse cross section perpendicular to the axial direction. The ratio of burnable poison contents of both regions is determined to not more than 80%, and the average concentration of the burnable poison in the lower region is determined to not less than 9% by weight. An infinite multiplication factor at an initial stage of the burning of the fuel assembly is controlled effectively by the burnable poisons. Namely, the ratio of the axial power can be controlled by the distribution of the enrichment degree of uranium fuels and the distribution of the burnable poison concentration in the axial direction. Since the average enrichment degree of the reactor core has to be increased in order to provide an initially loaded reactor core at high burnup degree. Distortion of the power distribution in the axial direction of the reactor core to which fuel assemblies at high enrichment degree are loaded is flattened to improve thermal margin, to extend continuous operation period and increase a burnup degree upon take-out thereby improving fuel economy without worsening the reactor core characteristics of the initially loaded reactor core. (N.H.)

Power distribution changes caused by subcooled nucleate boiling at Callaway Nuclear Power Plant

International Nuclear Information System (INIS)

Konya, M.J.; Bryant, K.R.; Hopkins, D.L.

1993-01-01

This paper reports the results of an evaluation undertaken by Union Electric (UE) and Westinghouse to explain anomalous behavior of the core axial power distribution at the Callaway Nuclear Power Plant. The behavior was characterized by a gradual unexpected power shift toward the bottom of the core and was first detected during cycle 4 at a core average burnup of approximately 7,000 MWD/MTU. Once started, the power shift continued until burnup effects became dominant and caused power to shift back to the top of the core at the end of the cycle. In addition to the anomalous power distribution, UE observed that estimated critical control rod position (ECP) deviations increased to over 500 pcm (0.5%Δk/k) during Cycles 4 and 5. ECPs for plant restarts that occurred early in each cycle agreed well with measured critical conditions. However, this agreement disappeared for restarts that occurred later in core life. After analyzing relevant data, performing scoping calculations and reviewing industry experience, the authors concluded that the power distribution anomaly was most likely caused by subcooled nucleate boiling. Crud deposition on the fuel was believed to enhance the subcooled boiling. The ECP deviations were a secondary effect of the power shift, since void fraction, axial burnup and xenon distributions departed design predictions during a substantial portion of the fuel cycles. Significant evidence supporting these conclusions include incore detector indications of flux depressions between intermediate flow mixing (IFM) and structural grids. In addition, visual exam results show the presence of crud deposits on fuel pins

A study on the core characteristics of the fact breed reactor

International Nuclear Information System (INIS)

Cho, M.; Ryu, K.G.; Soh, D.S.; Kim, Y.C.; Lee, K.W.; Kim, Y.I.; Lim, J.C.; Oh, K.B.

1983-01-01

In order to investigate equilibrium core characteristics of a large LMFBR, nuclear-thermal hydraulic and safety-related characteristics are considered for the equilbrium core of Super-Phenix 1. Using the nuclear computational system for a large LMFBR (KAERI-26G cross section library/1DX/2DB), bias factor of the effective multiplication factor, refueling pattern and enrichments of refueled fuel assemblies are determined. Nuclear characteristics such as criticality, power distribution, and breeding gain are also obtained for the initial core to the equilibrium core. Based on the assembly power distribution of the equilibrium core, coolant flow distributions are determined with the use of THI3D code. Temperature fields within the core assemblies and pressure drop across the core assemblies are also analyzed. In addition, the subchannel analysis of the hottest assembly under the steady state is performed with COBRA-IV-I code. Finally, transient behaviors of reactivity, power, and temperature due to the loss of flow accident without scram are considered and the boiling initiation time of coolant in the fuel assembly is determined for the equilibrium core with SACO code. (Author)

Optimal burnable poison utilization in PWR core reload design

International Nuclear Information System (INIS)

Downar, T.J.

1986-01-01

A method was developed for determining the optimal distribution and depletion of burnable poisons in a Pressurized Water Reactor core. The well-known Haling depletion technique is used to achieve the end-of-cycle core state where the fuel assembly arrangement is configured in the absence of all control poison. The soluble and burnable poison required to control the core reactivity and power distribution are solved for as unknown variables while step depleting the cycle in reverse with a target power distribution. The method was implemented in the NRC approved licensing code SIMULATE

Reactor core in FBR type reactor

International Nuclear Information System (INIS)

Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

1989-01-01

In a reactor core in FBR type reactors, a portion of homogenous fuels constituting the homogenous reactor core is replaced with multi-region fuels in which the enrichment degree of fissile materials is lower nearer to the axial center. This enables to condition the composition such that a reactor core having neutron flux distribution either of a homogenous reactor core or a heterogenous reactor core has substantially identical reactivity. Accordingly, in the transfer from the homogenous reactor core to the axially heterogenous reactor core, the average reactivity in the reactor core is substantially equal in each of the cycles. Further, by replacing a portion of the homogenous fuels with a multi-region fuels, thereby increasing the heat generation near the axial center, it is possiable to reduce the linear power output in the regions above and below thereof and, in addition, to improve the thermal margin in the reactor core. (T.M.)

Preliminary core design of IRIS-50

International Nuclear Information System (INIS)

Petrovic, Bojan; Franceschini, Fausto

2009-01-01

IRIS-50 is a small, 50 MWe, advanced PWR with integral primary system. It evolved employing the same design principles as the well known medium size (335 MWe) IRIS. These principles include the 'safety-by-design' philosophy, simple and robust design, and deployment flexibility. The 50 MWe design addresses the needs of specific applications (e.g., power generation in small regional grids, water desalination and biodiesel production at remote locations, autonomous power source for special applications, etc.). Such applications may favor or even require longer refueling cycles, or may have some other specific requirements. Impact of these requirements on the core design and refueling strategy is discussed in the paper. Trade-off between the cycle length and other relevant parameters is addressed. A preliminary core design is presented, together with the core main reactor physics performance parameters. (author)

Analysis of core uncovery time in Kuosheng station blackout transient with MELCOR

International Nuclear Information System (INIS)

Wang, S.J.; Chien, C.S.

1996-01-01

The MELCOR code, developed by the Sandia National Laboratories, is capable of simulating severe accident phenomena of nuclear power plants. Core uncovery time is an important parameter in the probabilistic risk assessment. However, many MELCOR users do not generate the initial conditions in a station blackout (SBO) transient analysis. Thus, achieving reliable core uncovery time is difficult. The core uncovery time for the Kuosheng nuclear power plant during an SBO transient is analyzed. First, full-power steady-state conditions are generated with the application of a developed self-initialization algorithm. Then the response of the SBO transient up to core uncovery is simulated. The effects of key parameters including the initialization process and the reactor feed pump (RFP) coastdown time on the core uncovery time are analyzed. The initialization process is the most important parameter that affects the core uncovery time. Because SBO transient analysis, the correct initial conditions must be generated to achieve a reliable core uncovery time. The core uncovery time is also sensitive to the RFP coastdown time. A correct time constant is required

Fast reactor core monitoring device

International Nuclear Information System (INIS)

Sanda, Toshio; Inoue, Kotaro; Azekura, Kazuo.

1982-01-01

Purpose: To enable the rapid and accurate on-line identification of the state of a fast reactor core by effectively utilizing the measured data on the temperature and flow rate of the coolant. Constitution: The spacial power distribution and average assembly power are quickly calculated using an approximate calculating method, the measured values and the calculated values of the inlet and outlet temperature difference, flow rate and coolant physical values of an assembly are combined and are individually obtained, the most definite respective values and their errors are obtained by a least square method utilizing a formula of the relation between these values, and the power distribution and the temperature distribution of a reactor core are estimated in this manner. Accordingly, even when the measuring accuracy and the calculating accuracy are equal as in a fast reactor, the power distribution and the temperature distribution can be accurately estimated on-line at a high speed in a nuclear reactor, information required for the operator is provided, and the reactor can thus be safely and efficiently operated. (Yoshihara, H.)

Thermal hydraulics model for Sandia's annular core research reactor

International Nuclear Information System (INIS)

Rao, Dasari V.; El-Genk, Mohamed S.; Rubio, Reuben A.; Bryson, James W.; Foushee, Fabian C.

1988-01-01

A thermal hydraulics model was developed for the Annular Core Research Reactor (ACRR) at Sandia National Laboratories. The coupled mass, momentum and energy equations for the core were solved simultaneously using an explicit forward marching numerical technique. The model predictions of the temperature rise across the central channel of the ACRR core were within ± 10 percent agreement with the in-core temperature measurements. The model was then used to estimate the coolant mass flow rate and the axial distribution of the cladding surface temperature in the central and average channels as functions of the operating power and the water inlet subcooling. Results indicated that subcooled boiling occurs at the cladding surface in the central channels of the ACRR at power levels in excess of 0.5 MW. However, the high heat transfer coefficient due to subcooled boiling causes the cladding temperature along most of the active fuel rod region to be quite uniform and to increase very little with the reactor power. (author)

TRACE analysis of Phenix core response to an increase of the core inlet sodium temperature

Energy Technology Data Exchange (ETDEWEB)

Chenu, A., E-mail: aurelia.chenu@psi.ch [Paul Scherrer Inst., Villigen PSI (Switzerland); Ecole Polytechnique Federale (Switzerland); Mikityuk, K., E-mail: konstantin.mikityuk@psi.ch [Paul Scherrer Inst., Villigen PSI (Switzerland); Adams, R., E-mail: robert.adams@psi.ch [Paul Scherrer Inst., Villigen PSI (Switzerland); Eidgenossische Technische Hochschule, Zurich (Switzerland); Chawla, R., E-mail: rakesh.chawla@epfl.ch [Paul Scherrer Inst., Villigen PSI (Switzerland); Ecole Polytechnique Federale (Switzerland)

2011-07-01

This work presents the analysis, using the TRACE code, of the Phenix core response to an inlet sodium temperature increase. The considered experiment was performed in the frame of the Phenix End-Of-Life (EOL) test program of the CEA, prior to the final shutdown of the reactor. It corresponds to a transient following a 40°C increase of the core inlet temperature, which leads to a power decrease of 60%. This work focuses on the first phase of the transient, prior to the reactor scram and pump trip. First, the thermal-hydraulic TRACE model of the core developed for the present analysis is described. The kinetic parameters and feedback coefficients for the point kinetic model were first derived from a 3D static neutronic ERANOS model developed in a former study. The calculated kinetic parameters were then optimized, before use, on the basis of the experimental reactivity in order to minimize the error on the power calculation. The different reactivity feedbacks taken into account include various expansion mechanisms that have been specifically implemented in TRACE for analysis of fast-neutron spectrum systems. The point kinetic model has been used to study the sensitivity of the core response to the different feedback effects. The comparison of the calculated results with the experimental data reveals the need to accurately calculate the reactivity feedback coefficients. This is because the reactor response is very sensitive to small reactivity changes. This study has enabled us to study the sensitivity of the power change to the different reactivity feedbacks and define the most important parameters. As such, it furthers the validation of the FAST code system, which is being used to gain a more in-depth understanding of SFR core behavior during accidental transients. (author)

Soliton formation in hollow-core photonic bandgap fibers

DEFF Research Database (Denmark)

Lægsgaard, Jesper

2009-01-01

of an approximate scaling relation is tested. It is concluded that compression of input pulses of several ps duration and sub-MW peak power can lead to a formation of solitons with ∼100 fs duration and multi-megawatt peak powers. The dispersion slope of realistic hollow-core fibers appears to be the main obstacle......The formation of solitons upon compression of linearly chirped pulses in hollow-core photonic bandgap fibers is investigated numerically. The dependence of soliton duration on the chirp and power of the input pulse and on the dispersion slope of the fiber is investigated, and the validity...

Ultrasound guided core biopsy of suspicious mammographic calcifications using high frequency and power Doppler ultrasound

International Nuclear Information System (INIS)

Teh, W.L.; Wilson, A.R.M; Evans, A.J.; Burrell, H.; Pinder, S.E.; Ellis, I.O.

2000-01-01

AIM: The pre-operative diagnosis of suspicious mammographic microcalcifications usually requires stereotactic needle biopsy. The aim of this study was to evaluate if high frequency 13 MHz ultrasound (HFUS) and power Doppler (PD) can aid visualization and biopsy of microcalcifications. MATERIALS AND METHODS: Forty-four consecutive patients presenting with microcalcifications without associated mammographic or palpable masses were examined with HFUS and PD. Ultrasound-guided core biopsy (USCB) was performed where possible. Stereotactic biopsy was carried out when US-guided biopsy was unsuccessful. Surgery was performed if a diagnosis of malignancy was made on core biopsy or if the repeat core biopsy was non-diagnostic. RESULTS: Forty-one patients (93%) had ultrasound abnormalities corresponding to mammographic calcification. USCB was performed on 37 patients. In 29/37, USCB obtained a definitive result (78.4%). USCB was non-diagnostic in 4/9 benign (44.4%) and 4/28 (14.3%) malignant lesions biopsied. The complete and absolute sensitivities for malignancy using USCB were 85.7% (24/28) and 81% (23/28), respectively. USCB correctly identified invasive disease in 12/23 (52.2%) cases. There was no significant difference in the presence of abnormal flow on PD between benign and malignant lesions. However, abnormal PD vascularity was present in 43.5% of invasive cancer and was useful in directing successful biopsy in eight cases. CONCLUSION: The combination of high frequency US with PD is useful in the detection and guidance of successful needle biopsy of microcalcifications particularly where there is an invasive focus within larger areas of DCIS. Teh, W.L. (2000)

Modelling of reactor control and protection systems in the core simulator program GARLIC

International Nuclear Information System (INIS)

Beraha, D.; Lupas, O.; Ploegert, K.

1984-01-01

For analysis of the interaction between control and limitation systems and the power distribution in the reactor core, a valuable tool is provided by the joint simulation of the core and the interacting systems. To this purpose, the core simulator GARLIC has been enhanced by models of the systems for controlling and limiting the reactor power and the power distribution in the core as well as by modules for calculating safety related core parameters. The computer-based core protection system, first installed in the Grafenrheinfeld NPP, has been included in the simulation. In order to evaluate the accuracy of GARLIC-simulations, the code has been compared with a design code in the train of a verification phase. The report describes the program extensions and the results of the verification. (orig.) [de

Impact of advanced BWR core physics method on BWR core monitoring

Energy Technology Data Exchange (ETDEWEB)

Moon, H; Wells, A [Siemens Power Corporation, Richland (United States)

2000-07-01

Siemens Power Corporation recently initiated development of POWERPLEX{sup TM}-III for delivery to the Grand Gulf Nuclear Power Station. The main change introduced in POWERPLEX{sup TM}-III as compared to its predecessor POWERPLEX{sup TM}-II is the incorporation of the advances BWR core simulator MICROBURN-B2. A number of issues were identified and evaluated relating to the implementation of MICROBURN-B2 and its impact on core monitoring. MICROBURN-B2 demands about three to five times more memory and two to three times more computing time than its predecessor MICROBURN-B in POWERPLEX {sup TM}-II. POWERPLEX{sup TM}-III will improve thermal margin prediction accuracy and provide more accurate plant operating conditions to operators than POWERPLEX{sup TM}-II due to its improved accuracy in predicted TIP values and critical k-effective. The most significant advantage of POWERPLEX{sup TM}-III is its capability to monitor a relaxed rod sequence exchange operation. (authors)

Thermal hydraulic design of a hydride-fueled inverted PWR core

International Nuclear Information System (INIS)

Malen, J.A.; Todreas, N.E.; Hejzlar, P.; Ferroni, P.; Bergles, A.

2009-01-01

An inverted PWR core design utilizing U(45%, w/o)ZrH 1.6 fuel (here referred to as U-ZrH 1.6 ) is proposed and its thermal hydraulic performance is compared to that of a standard rod bundle core design also fueled with U-ZrH 1.6 . The inverted design features circular cooling channels surrounded by prisms of fuel. Hence the relative position of coolant and fuel is inverted with respect to the standard rod bundle design. Inverted core designs with and without twisted tape inserts, used to enhance critical heat flux, were analyzed. It was found that higher power and longer cycle length can be concurrently achieved by the inverted core with twisted tape relative to the optimal standard core, provided that higher core pressure drop can be accommodated. The optimal power of the inverted design with twisted tape is 6869 MW t , which is 135% of the optimally powered standard design (5080 MW t -determined herein). Uncertainties in this design regarding fuel and clad dimensions needed to accommodate mechanical loads and fuel swelling are presented. If mechanical and neutronic feasibility of these designs can be confirmed, these thermal assessments imply significant economic advantages for inverted core designs.

Reactor core with rod-shaped fuel cells

International Nuclear Information System (INIS)

Dworak, A.

1976-01-01

The proposal refers to the optimization of the power distribution in a reactor core which is provided with several successive rod-shaped fuel cells. A uniform power output - especially in radial direction - is aimed at. This is achieved by variation of the dwelling periods of the fuel cells, which have, for this purpose, a fuel mixture changing from layer to layer. The fuel cells with the shortest dwelling period are arranged near the coolant inlet side of the reactor core. The dwelling periods of the fuel cells are adapted to the given power distribution. As neighboring cells have equal dwelling periods, the exchange can be performed much easier then with the composition currently known. (UWI) [de

Increasing the neutron flux study for the TRR-II core design

International Nuclear Information System (INIS)

Chen, C.-H.; Yang, J.-T.; Chou, Y.-C.

1999-01-01

The maximum unperturbed thermal flux of the originally proposed core design, which is a 6x6 square arrangement with power level of 20 MW and has been presented at the 6th Meeting of IGORR, for the TRR-II reactor is about 2.0x10 14 n/cm 2 -sec. However, it is no longer satisfied the user's requirement, that is, it must reach at least 2.5x10 14 n/cm 2 -sec. In order to enhance the thermal neutron flux, one of the most effective ways is to increase the average power density. Therefore, two new designs with more compact cores are then proposed and studied. One is 5x6 rectangular arrangement with power of 20 MW; the other one is 5x5 square arrangement with power of 16 MW. It is for sure that both core designs can satisfy thermal hydraulic safety limits. The designed parameters related to neutronics are listed and compared fundamentally. According to our calculation, although both cores have similar average power density, the results show that the 5x6/20 MW design has the maximum unperturbed thermal flux in the D 2 O region about 2.7x10 14 n/cm 2 -sec, and the 5x5/16 MW design has 2.5x10 14 n/cm 2 -sec. The maximum thermal flux in the neighborhood of the longer side of the 5x6 core is about 7% higher than the one in the neighborhood of any side of the 5x5 core. This 'long-side effect' gives the 5x6/20 MW core design an advantage of the utilization of the thermal neutron flux in the D 2 O region. In addition, the 5x5 core is also more sensitive to the reactivity change on account of in-core irradiation test facilities. Therefore, under overall considerations the 5x6/20 MW core design is chosen for further detailed design. (author)

Reconstruction of core inlet temperature distribution by cold leg temperature measurements

International Nuclear Information System (INIS)

Saarinen, S.; Antila, M.

2010-01-01

The reduced core of Loviisa NPP contains 33 thermocouple measurements measuring the core inlet temperature. Currently, these thermocouple measurements are not used in determining the inlet temperature distribution. The average of cold leg temperature measurements is used as inlet temperature for each fuel assembly. In practice, the inlet temperature distribution is not constant. Thus, using a constant inlet temperature distribution induces asymmetries in the measured core power distribution. Using a more realistic inlet temperature distribution would help us to reduce virtual asymmetries of the core power distribution and increase the thermal margins of the core. The thermocouples at the inlet cannot be used directly to measure the inlet temperature accurately because the calibration of the thermocouples that is done at hot zero power conditions is no longer valid at full power, when there is temperature change across the core region. This is due to the effect of neutron irradiation on the Seebeck coefficient of the thermocouple wires. Therefore, we investigate in this paper a method to determine the inlet temperature distribution based on the cold leg temperature measurements. With this method we rely on the assumption that although the core inlet thermocouple measurements do not measure the absolute temperature accurately they do measure temperature changes with sufficient accuracy particularly in big disturbances. During the yearly testing of steam generator safety valves we observe a large temperature increase up to 12 degrees in the cold leg temperature. The change in the temperature of one of the cold legs causes a local disturbance in the core inlet temperature distribution. Using the temperature changes observed in the inlet thermocouple measurements we are able to fit six core inlet temperature response functions, one for each cold leg. The value of a function at an assembly inlet is determined only by the corresponding cold leg temperature disturbance

Gas core reactors for actinide transmutation and breeder applications. Annual report

International Nuclear Information System (INIS)

Clement, J.D.; Rust, J.H.

1978-01-01

This work consists of design power plant studies for four types of reactor systems: uranium plasma core breeder, uranium plasma core actinide transmuter, UF6 breeder and UF6 actinide transmuter. The plasma core systems can be coupled to MHD generators to obtain high efficiency electrical power generation. A 1074 MWt UF6 breeder reactor was designed with a breeding ratio of 1.002 to guard against diversion of fuel. Using molten salt technology and a superheated steam cycle, an efficiency of 39.2% was obtained for the plant and the U233 inventory in the core and heat exchangers was limited to 105 Kg. It was found that the UF6 reactor can produce high fluxes (10 to the 14th power n/sq cm-sec) necessary for efficient burnup of actinide. However, the buildup of fissile isotopes posed severe heat transfer problems. Therefore, the flux in the actinide region must be decreased with time. Consequently, only beginning-of-life conditions were considered for the power plant design. A 577 MWt UF6 actinide transmutation reactor power plant was designed to operate with 39.3% efficiency and 102 Kg of U233 in the core and heat exchanger for beginning-of-life conditions

Experimentation of a fixed in-core-based system for core limiting conditions of operation (LCO) monitoring

International Nuclear Information System (INIS)

Piguet, F.; Carrasco, M.; Mourlevat, J.L.; Rio, G.; Verneret, C.

2006-01-01

In order to comply with the needs of Utilities for improvements in the economic competitiveness of nuclear energy, one of the solutions proposed is to reduce the cost of the fuel cycle. To this aim, increasing the lifetime of cycles by introducing so-called 'low leakage' fuel loading patterns to the reactor is a rather promising solution. However, these loading patterns lead to an increase in the core hotspot factors and therefore to a reduction in the operating margins with respect to the core operating limits also called 'Limiting Conditions of Operations (LCO)'. For many years FRAMATOME-ANP has developed and proposed solutions aiming at increasing and therefore restoring these margins, namely: the improvement in design methods based on three-dimensional modelling of the core, on kinetic representation of transients and on neutron-thermohydraulic coupling or the improvement in the fuel with the introduction of intermediate grids. A complementary approach is to improve the core instrumentation associated with the system for monitoring the core operating margins to the LCO thresholds. The core operating limits monitoring function calls on real-time knowledge of the current power distribution in the core. If we take the French 1300 MWe units as an example, this knowledge is based on the measurement of the mean axial power distribution made by six sections neutron detectors, located outside the pressure vessel and equipped with a fast neutron filtering device. The results of this measurement are combined with pre-tabulated radial hotspot factors (Fxy), in order to calculate the total hotspot factor (FQ) of the core, the minimum Departure from Nucleate Boiling Ratio (DNBR) and, consequently, the margins with respect to the core operating limits. The limitations of a measurement made outside the vessel, and those of the 1D/2D modelling adopted, mean that these margins calculations have a high potential for improving the level of their accuracy. This is the reason why

Development of Structural Core Components for Breeder Reactors

International Nuclear Information System (INIS)

Saibaba, N.

2013-01-01

Core structural materials: • The desire is to have only fuel in the core, structural material form 25% of the total core: – To support and to retain the fuel in position; – Provide necessary ducts to make coolant flow through & transfer/remove heat. • For 500 MWe FBR with Oxide fuel (Peak Linear Power 450 W/cm), total fuel pins required in the core are of the order 39277 pins (both inner & outer core Fuel SA); • Considering 217 pins/Fuel SA there are 181 Fuel SA wrapper tubes • These structural materials see hostile core with max temperature and neutron flux

Solid core dipoles and switching power supplies: lower cost light sources?

Science.gov (United States)

Benesch, J.; Philip, S.

2015-05-01

As a result of improvements in power semiconductors, moderate frequency switching supplies can now provide the hundreds of amps typically required by accelerators with zero-to-peak noise in the kHz region ~ 0.06% in current or voltage mode. Modeling was undertaken using a finite electromagnetic program to determine if eddy currents induced in the solid steel of CEBAF magnets and small supplemental additions would bring the error fields down to the 5ppm level needed for beam quality. The expected maximum field of the magnet under consideration is 0.85 T and the DC current required to produce that field is used in the calculations. An additional 0.1% current ripple is added to the DC current at discrete frequencies 360 Hz, 720 Hz or 7200 Hz. Over the region of the pole within 0.5% of the central integrated BdL the resulting AC field changes can be reduced to less than 1% of the 0.1% input ripple for all frequencies, and a sixth of that at 7200 Hz. Doubling the current, providing 1.5 T central field, yielded the same fractional reduction in ripple at the beam for the cases checked. A small dipole was measured at 60, 120, 360 and 720 Hz in two conditions and the results compared to the larger model for the latter two frequencies with surprisingly good agreement. For light sources with aluminum vacuum vessels and full energy linac injection, the combination of solid core dipoles and switching power supplies may result in significant cost savings. The work may also be used to guide retrofit of existing machines to reduce the level of ripple in the particle beam path.

Study of core support barrel vibration monitoring using ex-core neutron noise analysis and fuzzy logic algorithm

International Nuclear Information System (INIS)

Christian, Robby; Song, Seon Ho; Kang, Hyun Gook

2015-01-01

The application of neutron noise analysis (NNA) to the ex-core neutron detector signal for monitoring the vibration characteristics of a reactor core support barrel (CSB) was investigated. Ex-core flux data were generated by using a nonanalog Monte Carlo neutron transport method in a simulated CSB model where the implicit capture and Russian roulette technique were utilized. First and third order beam and shell modes of CSB vibration were modeled based on parallel processing simulation. A NNA module was developed to analyze the ex-core flux data based on its time variation, normalized power spectral density, normalized cross-power spectral density, coherence, and phase differences. The data were then analyzed with a fuzzy logic module to determine the vibration characteristics. The ex-core neutron signal fluctuation was directly proportional to the CSB's vibration observed at 8Hz and15Hzin the beam mode vibration, and at 8Hz in the shell mode vibration. The coherence result between flux pairs was unity at the vibration peak frequencies. A distinct pattern of phase differences was observed for each of the vibration models. The developed fuzzy logic module demonstrated successful recognition of the vibration frequencies, modes, orders, directions, and phase differences within 0.4 ms for the beam and shell mode vibrations.

Core rotational dynamics and geological events

Science.gov (United States)

Greff-Lefftz; Legros

1999-11-26

A study of Earth's fluid core oscillations induced by lunar-solar tidal forces, together with tidal secular deceleration of Earth's axial rotation, shows that the rotational eigenfrequency of the fluid core and some solar tidal waves were in resonance around 3.0 x 10(9), 1.8 x 10(9), and 3 x 10(8) years ago. The associated viscomagnetic frictional power at the core boundaries may be converted into heat and would destabilize the D" thermal layer, leading to the generation of deep-mantle plumes, and would also increase the temperature at the fluid core boundaries, perturbing the core dynamo process. Such phenomena could account for large-scale episodes of continental crust formation, the generation of flood basalts, and abrupt changes in geomagnetic reversal frequency.

Shaping of the axial power density distribution in the core to minimize the vapor volume fraction at the outlet of the VVER-1200 fuel assemblies

Energy Technology Data Exchange (ETDEWEB)

Savander, V. I.; Shumskiy, B. E., E-mail: borisshumskij@yandex.ru [National Research Nuclear University MEPhI (Russian Federation); Pinegin, A. A. [National Research Center Kurchatov Institute (Russian Federation)

2016-12-15

The possibility of decreasing the vapor fraction at the VVER-1200 fuel assembly outlet by shaping the axial power density field is considered. The power density field was shaped by axial redistribution of the concentration of the burnable gadolinium poison in the Gd-containing fuel rods. The mathematical modeling of the VVER-1200 core was performed using the NOSTRA computer code.

Prospects for future uranium savings through LWRs with high performance cores

International Nuclear Information System (INIS)

Mochida, T.; Yamamoto, T.; Sasaki, M.; Matsuura, H.; Ueji, M.; Murata, T.; Kanda, K.; Oka, Y.; Kondo, S.

1995-01-01

Since 1986, Nuclear Power Engineering Cooperation (NUPEC) has been studying four types of LWR high performance core concepts (i.e., the uranium saving core I (USC-I), the uranium saving core II (USC-II), the high moderation core (HMC) and the low moderation core (LMC)), which aim at improvement of uranium and plutonium utilization. After the evaluation of fundamental core performance and uranium and plutonium material balance for each reactor, potential uranium savings with different reactor strategies are evaluated for the Japanese scenario with assumption of the growth of future nuclear power plant generation, annual reprocessing capacity and schedules for the introduction of high performance core. At 2030, about 3-6% savings in uranium demand are expected by USC-I or USC-II strategy, while about 14% savings by HMC strategy and about 8% by LMC strategy. (author)

One Core Phase Shifting Transformer for Control of the Power Flow Distribution in Electric Networks

Directory of Open Access Journals (Sweden)

Golub I.V.

2016-08-01

Full Text Available This paper presents the variant of phase shifting transformer that is made, unlike from traditional technology, on the basis of only one magnetic core. The paper describes the methodology related to the analysis of operation modes of device and its components. Additionally it presents a mathematical model of device with determines the relationship between input and output electric quantities as well as own longitudinal and transverse parameters of an equivalent circuit of phase shifting transformer (PST. Proposed configuration of PST is interesting from an economic and operational consideration; enable continuous control of power flow distribution in electric networks as a result of regulation a phase shift angle between input and output voltages of device.

Spontaneous stabilization of HTGRs without reactor scram and core cooling—Safety demonstration tests using the HTTR: Loss of reactivity control and core cooling

Energy Technology Data Exchange (ETDEWEB)

Takamatsu, Kuniyoshi, E-mail: takamatsu.kuniyoshi@jaea.go.jp; Yan, Xing L.; Nakagawa, Shigeaki; Sakaba, Nariaki; Kunitomi, Kazuhiko

2014-05-01

It is well known that a High-Temperature Gas-cooled Reactor (HTGR) has superior safety characteristics; for example, an HTGR has a self-control system that uses only physical phenomena against various accidents. Moreover, the large heat capacity and low power density of the core result in very slow temperature transients. Therefore, an HTGR serves inherently safety features against loss of core cooling accidents such as the Tokyo Electric Power Co., Inc. (TEPCO)’s Fukushima Daiichi Nuclear Power Station (NPS) disaster. Herein we would like to demonstrate the inherent safety features using the High-Temperature Engineering Test Reactor (HTTR). The HTTR is the first HTGR in Japan with a thermal power of 30 MW and a maximum reactor outlet coolant temperature of 950 °C; it was built at the Oarai Research and Development Center of Japan Atomic Energy Agency (JAEA). In this study, an all-gas-circulator trip test was analyzed as a loss of forced cooling (LOFC) test with an initial reactor power of 9 MW to demonstrate LOFC accidents. The analytical results indicate that reactor power decreases from 9 MW to 0 MW owing to the negative reactivity feedback effect of the core, even if the reactor shutdown system is not activated. The total reactivity decreases for 2–3 h and then gradually increases in proportion to xenon reactivity; therefore, the HTTR achieves recritical after an elapsed time of 6–7 h, which is different from the elapsed time at reactor power peak occurrence. After the reactor power peak occurs, the total reactivity oscillates several times because of the negative reactivity feedback effect and gradually decreases to zero. Moreover, the new conclusions are as follows: the greater the amount of residual heat removed from the reactor core, the larger the stable reactor power after recriticality owing to the heat balance of the reactor system. The minimum reactor power and the reactor power peak occurrence are affected by the neutron source. The greater the

Effects of sintering temperature on properties of toroid cores using NiZnCu ferrites for power applications at >1 MHz

Science.gov (United States)

Liu, Junchang; Mei, Yunhui; Liu, Wen; Li, Xin; Hou, Feng; Lu, Guo-Quan

2018-05-01

The microstructures, magnetic and electronic performance of NiZnCu ferrites have been investigated at temperature from 850 °C to 1000 °C. X-ray diffraction (XRD) patterns showed that only single phase with spinel structure existed. Scanning electron microscopy (SEM) results showed that grain size increased with enhancement of sintering temperature and the most homogeneous, compact microstructure was obtained at 950 °C. Magnetic properties measurements revealed that both complex permeability and saturation magnetization increased with increasing of sintering temperature. The initial permeability was approximately linear within the scope of 850-1000 °C as well as the resonance frequency decreased from 70 MHz to 30 MHz. Power loss density tests demonstrated that the core sintered at 950 °C instead of the one sintered at 1000 °C had the lower power loss density at both 5 mT and 10 mT and the higher inductance under a certain exciting direct current at 1 MHz. Also the inductance of the sample sintered at the higher temperature dropped faster than that at the lower temperature. The results showed that the core sintered at 950 °C had better electrical performance and was suitable for wide usage.

Thermal neutron flux measurement using self-powered neutron detector (SPND) at out-core locations of TRIGA PUSPATI Reactor (RTP)

Science.gov (United States)

Ali, Nur Syazwani Mohd; Hamzah, Khaidzir; Mohamad Idris, Faridah; Hairie Rabir, Mohamad

2018-01-01

The thermal neutron flux measurement has been conducted at the out-core location using self-powered neutron detectors (SPNDs). This work represents the first attempt to study SPNDs as neutron flux sensor for developing the fault detection system (FDS) focusing on neutron flux parameters. The study was conducted to test the reliability of the SPND’s signal by measuring the neutron flux through the interaction between neutrons and emitter materials of the SPNDs. Three SPNDs were used to measure the flux at four different radial locations which located at the fission chamber cylinder, 10cm above graphite reflector, between graphite reflector and tank liner and fuel rack. The measurements were conducted at 750 kW reactor power. The outputs from SPNDs were collected through data acquisition system and were corrected to obtain the actual neutron flux due to delayed responses from SPNDs. The measurements showed that thermal neutron flux between fission chamber location near to the tank liner and fuel rack were between 5.18 × 1011 nv to 8.45 × 109 nv. The average thermal neutron flux showed a good agreement with those from previous studies that has been made using simulation at the same core configuration at the nearest irradiation facilities with detector locations.

Reactor core of FBR type reactor

International Nuclear Information System (INIS)

Hayashi, Hideyuki; Ichimiya, Masakazu.

1994-01-01

A reactor core is a homogeneous reactor core divided into two regions of an inner reactor core region at the center and an outer reactor core region surrounding the outside of the inner reactor core region. In this case, the inner reactor core region has a lower plutonium enrichment degree and less amount of neutron leakage in the radial direction, and the outer reactor core region has higher plutonium enrichment degree and greater amount of neutron leakage in the radial direction. Moderator materials containing hydrogen are added only to the inner reactor core fuels in the inner reactor core region. Pins loaded with the fuels with addition of the moderator materials are inserted at a ratio of from 3 to 10% of the total number of the fuel pins. The moderator materials containing hydrogen comprise zirconium hydride, titanium hydride, or calcium hydride. With such a constitution, fluctuation of the power distribution in the radial direction along with burning is suppressed. In addition, an absolute value of the Doppler coefficient can be increased, and a temperature coefficient of coolants can be reduced. (I.N.)

Conceptual design of PFBR core

International Nuclear Information System (INIS)

Lee, S.M.; Govindarajan, S.; Indira, R.; John, T.M.; Mohanakrishnan, P.; Shankar Singh, R.; Bhoje, S.B.

1996-01-01

The design options selected for the core of the 500 MWe Prototype Fast Breeder Reactor are presented. PFBR has a conventional mixed oxide fuel core of homogeneous type with two enrichment zones for power flattening and with radial and axial blankets to make the reactor self-sustaining in fissile material. Pin diameter has been selected for minimization of fissile inventory. Considerations for the choice of number of pins per subassembly, integrated versus separate axial blankets, and other pin and subassembly parameters are discussed. As the core size is moderate, no special schemes for reducing the maximum positive sodium voiding coefficient is envisages. Two independent, diverse fast acting shutdown systems working in fail-safe mode are selected. The number of absorber rods has been minimized by choosing a layout for maximum antishadow effect. Nine control and safety rods are distributed in two rods for power flattening by differential insertion. Three Diverse Safety Rods, are also provided which are normally fully withdrawn. The optimization of layout of radial and axial shielding and adequacy of flux at detector location are also discussed. (author). 2 figs

Development of core design and analyses technology for integral reactor

Energy Technology Data Exchange (ETDEWEB)

Zee, Sung Quun; Lee, C. C.; Song, J. S. and others

1999-03-01

Integral reactors are developed for the applications such as sea water desalination, heat energy for various industries, and power sources for large container ships. In order to enhance the inherent and passive safety features, low power density concept is chosen for the integral reactor SMART. Moreover, ultra-longer cycle and boron-free operation concepts are reviewed for better plant economy and simple design of reactor system. Especially, boron-free operation concept brings about large difference in core configurations and reactivity controls from those of the existing large size commercial nuclear power plants and also causes many differences in the safety aspects. The ultimate objectives of this study include detailed core design of a integral reactor, development of the core design system and technology, and finally acquisition of the system design certificate. The goal of the first stage is the conceptual core design, that is, to establish the design bases and requirements suitable for the boron-free concept, to develop a core loading pattern, to analyze the nuclear, thermal and hydraulic characteristics of the core and to perform the core shielding design. Interface data for safety and performance analyses including fuel design data are produced for the relevant design analysis groups. Nuclear, thermal and hydraulic, shielding design and analysis code systems necessary for the core conceptual design are established through modification of the existing design tools and newly developed methodology and code modules. Core safety and performance can be improved by the technology development such as boron-free core optimization, advaned core monitoring and operational aid system. Feasiblity study on the improvement of the core protection and monitoring system will also contribute toward core safety and performance. Both the conceptual core design study and the related technology will provide concrete basis for the next design phase. This study will also

Development of core design and analyses technology for integral reactor

International Nuclear Information System (INIS)

Zee, Sung Quun; Lee, C. C.; Song, J. S. and others

1999-03-01

Integral reactors are developed for the applications such as sea water desalination, heat energy for various industries, and power sources for large container ships. In order to enhance the inherent and passive safety features, low power density concept is chosen for the integral reactor SMART. Moreover, ultra-longer cycle and boron-free operation concepts are reviewed for better plant economy and simple design of reactor system. Especially, boron-free operation concept brings about large difference in core configurations and reactivity controls from those of the existing large size commercial nuclear power plants and also causes many differences in the safety aspects. The ultimate objectives of this study include detailed core design of a integral reactor, development of the core design system and technology, and finally acquisition of the system design certificate. The goal of the first stage is the conceptual core design, that is, to establish the design bases and requirements suitable for the boron-free concept, to develop a core loading pattern, to analyze the nuclear, thermal and hydraulic characteristics of the core and to perform the core shielding design. Interface data for safety and performance analyses including fuel design data are produced for the relevant design analysis groups. Nuclear, thermal and hydraulic, shielding design and analysis code systems necessary for the core conceptual design are established through modification of the existing design tools and newly developed methodology and code modules. Core safety and performance can be improved by the technology development such as boron-free core optimization, advaned core monitoring and operational aid system. Feasiblity study on the improvement of the core protection and monitoring system will also contribute toward core safety and performance. Both the conceptual core design study and the related technology will provide concrete basis for the next design phase. This study will also

Performance Evaluation of the Concept of Hybrid Heat Pipe as Passive In-core Cooling Systems for Advanced Nuclear Power Plant

International Nuclear Information System (INIS)

Jeong, Yeong Shin; Kim, Kyung Mo; Kim, In Guk; Bang, In Cheol

2015-01-01

As an arising issue for inherent safety of nuclear power plant, the concept of hybrid heat pipe as passive in-core cooling systems was introduced. Hybrid heat pipe has unique features that it is inserted in core directly to remove decay heat from nuclear fuel without any changes of structures of existing facilities of nuclear power plant, substituting conventional control rod. Hybrid heat pipe consists of metal cladding, working fluid, wick structure, and neutron absorber. Same with working principle of the heat pipe, heat is transported by phase change of working fluid inside metal cask. Figure 1 shows the systematic design of the hybrid heat pipe cooling system. In this study, the concept of a hybrid heat pipe was introduced as a Passive IN-core Cooling Systems (PINCs) and demonstrated for internal design features of heat pipe containing neutron absorber. Using a commercial CFD code, single hybrid heat pipe model was analyzed to evaluate thermal performance in designated operating condition. Also, 1-dimensional reactor transient analysis was done by calculating temperature change of the coolant inside reactor pressure vessel using MATLAB. As a passive decay heat removal device, hybrid heat pipe was suggested with a concept of combination of heat pipe and control rod. Hybrid heat pipe has distinct feature that it can be a unique solution to cool the reactor when depressurization process is impossible so that refueling water cannot be injected into RPV by conventional ECCS. It contains neutron absorber material inside heat pipe, so it can stop the reactor and at the same time, remove decay heat in core. For evaluating the concept of hybrid heat pipe, its thermal performance was analyzed using CFD and one-dimensional transient analysis. From single hybrid heat pipe simulation, the hybrid heat pipe can transport heat from the core inside to outside about 18.20 kW, and total thermal resistance of hybrid heat pipe is 0.015 .deg. C/W. Due to unique features of long heat

TMI-2 [Three Mile Island Nuclear Power Station] fuel canister and core sample handling equipment used in INEL hot cells

International Nuclear Information System (INIS)

McConnell, J.W. Jr.; Shurtliff, W.T.; Lynch, R.J.; Croft, K.M.; Whitmill, L.J.; Allen, S.M.

1987-01-01

This paper describes the specialized remote handling equipment developed and used at the Idaho National Engineering Laboratory (INEL) to handle samples obtained from the core of the damaged Unit 2 reactor at Three Mile Island Nuclear Power Station (TM-2). Samples of the core were removed, placed in TMI-2 fuel canisters, and transported to the INEL. Those samples will be examined as part of the analysis of the TMI-2 accident. The equipment described herein was designed for removing sample materials from the fuel canisters, assisting with initial examination, and processing samples in preparation for detailed examinations. The more complex equipment used microprocessor remote controls with electric motor drives providing the required force and motion capabilities. The remaining components were unpowered and manipulator assisted

Evaluation report on SCTF Core-III test S3-06

International Nuclear Information System (INIS)

Iwamura, Takamichi; Iguchi, Tadashi; Akimoto, Hajime; Okubo, Tsutomu; Ohnuki, Akira; Adachi, Hiromichi; Murao, Yoshio; Minato, Akihiko; Sakaki, Isao.

1988-10-01

In order to investigate the effect of radial power distribution on the thermal-hydraulic characteristics during the reflood phase of a PWR-LOCA with a combined injection type ECCS, a core cooling separate effect test S3-06 and a combined injection test S3-16-Phase 2 were performed using the Slab Core Test Facility (SCTF) Core-III. The radial power distributions in these two tests simulated a reference distribution for a PWR with a combined injection type ECCS and a steep distribution for a PWR with a cold leg injection type ECCS, respectively. Under the radial power distribution of a PWR with a combined injection type ECCS, the radial power distribution had little effect on the thermal-hydraulic behavior in the two-phase up-flow region due to the approximately flat power distribution in this region (power ratio = 1.04 ∼ 1.08). The overall fluid behavior in the pressure vessel was also little affected by the radial power distribution. On the other hand, under the steep radial power distribution (peak power ratio = 1.36), the degree of heat transfer enhancement in high power bundles in the two-phase up-flow region was dominated by the bundlewise radial power ratio as in the case of a PWR with a cold leg injection type ECCS. (author)