Bundle 13 position verification tool description and on-reactor use

Energy Technology Data Exchange (ETDEWEB)

Onderwater, T G [Canadian General Electric Co. Ltd., Peterborough, ON (Canada)

1997-12-31

To address the Power Pulse problem, Bruce B uses Gap: a comprehensive monitoring program by the station to maintain the gap between the fuel string and the upstream shield plug. The gap must be maintained within a band. The gap must not be so large as to allow excessive reactivity increases or cause high impact forces during reverse flow events. It should also not be so small as to cause crushed fuel during rapid, differential reactor/fuel string cool downs. Rapid cool downs are infrequent. The Bundle 13 Position Verification Tool (BPV tool) role is to independently measure the position of the upstream bundle of the fuel string. The measurements are made on-reactor, on-power and will allow verification of the Gap Management system`s calculated fuel string position. This paper reviews the reasons for developing the BPV tool. Design issues relevant to safe operation in the fuelling machine, fuel channel and fuel handling equipment are also reviewed. Tests ensuring no adverse effects on channel pressure losses are described and actual on-reactor, on-power results are discussed. (author). 4 figs.

Bundle 13 position verification tool description and on-reactor use

International Nuclear Information System (INIS)

Onderwater, T.G.

1996-01-01

To address the Power Pulse problem, Bruce B uses Gap: a comprehensive monitoring program by the station to maintain the gap between the fuel string and the upstream shield plug. The gap must be maintained within a band. The gap must not be so large as to allow excessive reactivity increases or cause high impact forces during reverse flow events. It should also not be so small as to cause crushed fuel during rapid, differential reactor/fuel string cool downs. Rapid cool downs are infrequent. The Bundle 13 Position Verification Tool (BPV tool) role is to independently measure the position of the upstream bundle of the fuel string. The measurements are made on-reactor, on-power and will allow verification of the Gap Management system's calculated fuel string position. This paper reviews the reasons for developing the BPV tool. Design issues relevant to safe operation in the fuelling machine, fuel channel and fuel handling equipment are also reviewed. Tests ensuring no adverse effects on channel pressure losses are described and actual on-reactor, on-power results are discussed. (author). 4 figs

MDEP Common Position CP-VICWG-01. Common Position: Establishment of Common QA/QM Criteria for the Multinational Vendor Inspection CP-VICWG-01

International Nuclear Information System (INIS)

2015-01-01

This document provides a set of common positions for harmonizing inspection criteria called 'Common QA/QM Criteria' which will be used in Multinational Vendor Inspections. This document was prepared by the Vendor Inspection Co-operation Working Group (VICWG) of the Multinational Design Evaluation Program (MDEP). The 'Common QA/QM Criteria' provides the basic areas for consideration when performing Vendor Inspections. The criteria have been developed in conformity with International Codes and Standards such as IAEA, ISO, etc. that MDEP member countries have adopted

Positioning means for circumferentially locating inspection apparatus in a nuclear reactor vessel

International Nuclear Information System (INIS)

Burns, D.C.

1980-01-01

There is provided for a reactor vessel inspection device a support ring sized to accommodate the circular path defined by three or more guide studs extending upwardly from the vessel. The support ring has at least three movably mounted guide stud bushings which can be positionally adjusted to align each bushing with one of the studs. When engaged, the guide studs and bushings yield a coarse positioning of the inspection device relative to the reactor vessel. Also provided are three support legs which are clamped to the support ring and dimensioned to an appropriate length. Two of the support legs have shoes clamped thereto, configured to rest on an internal circumferential flange within the reactor vessel. The third support leg is provided with a specially adapted shoe configured to engage a locating element, the exact position of which is known, within the vessel to achieve fine positioning of the inspection device relative to the reactor vessel. The support ring is additionally provided with an annular key which runs longitudinally about its outer periphery. Clamping means utilized to secure the guide stud bushings and the support legs to the support ring are provided with keyways to insure automatic self-alignment when fully tightened. (auth)

Licensing of safety critical software for nuclear reactors. Common position of seven European nuclear regulators and authorised technical support organisations

International Nuclear Information System (INIS)

2010-01-01

It is widely accepted that the assessment of software cannot be limited to verification and testing of the end product, i.e. the computer code. Other factors such as the quality of the processes and methods for specifying, designing and coding have an important impact on the implementation. Existing standards provide limited guidance on the regulatory and safety assessment of these factors. An undesirable consequence of this situation is that the licensing approaches taken by nuclear safety authorities and by technical support organisations are determined independently with only limited informal technical co-ordination and information exchange. It is notable that several software implementations of nuclear safety systems have been marred by costly delays caused by difficulties in co-ordinating the development and qualification process. It was thus felt necessary to compare the respective licensing approaches, to identify where a consensus already exists, and to see how greater consistency and more mutual acceptance could be introduced into current practices. This report is the result of the work of a group of regulator and safety authorities' experts. The 2007 version was completed at the invitation of the Western European Nuclear Regulators' Association (WENRA). The major result of the work is the identification of consensus and common technical positions on a set of important licensing issues raised by the design and operation of computer based systems used in nuclear power plants for the implementation of safety functions. The purpose is to introduce greater consistency and more mutual acceptance into current practices. To achieve these common positions, detailed consideration was paid to the licensing approaches followed in the different countries represented by the experts of the task force. The report is intended to be useful: - to coordinate regulators' and safety experts' technical viewpoints in the design of regulators' national policies and in revisions

Licensing of safety critical software for nuclear reactors. Common position of seven European nuclear regulators and authorised technical support organisations

Energy Technology Data Exchange (ETDEWEB)

2010-07-01

It is widely accepted that the assessment of software cannot be limited to verification and testing of the end product, i.e. the computer code. Other factors such as the quality of the processes and methods for specifying, designing and coding have an important impact on the implementation. Existing standards provide limited guidance on the regulatory and safety assessment of these factors. An undesirable consequence of this situation is that the licensing approaches taken by nuclear safety authorities and by technical support organisations are determined independently with only limited informal technical co-ordination and information exchange. It is notable that several software implementations of nuclear safety systems have been marred by costly delays caused by difficulties in co-ordinating the development and qualification process. It was thus felt necessary to compare the respective licensing approaches, to identify where a consensus already exists, and to see how greater consistency and more mutual acceptance could be introduced into current practices. This report is the result of the work of a group of regulator and safety authorities' experts. The 2007 version was completed at the invitation of the Western European Nuclear Regulators' Association (WENRA). The major result of the work is the identification of consensus and common technical positions on a set of important licensing issues raised by the design and operation of computer based systems used in nuclear power plants for the implementation of safety functions. The purpose is to introduce greater consistency and more mutual acceptance into current practices. To achieve these common positions, detailed consideration was paid to the licensing approaches followed in the different countries represented by the experts of the task force. The report is intended to be useful: - to coordinate regulators' and safety experts' technical viewpoints in the design of regulators' national

Common positioning errors in panoramic radiography: A review

International Nuclear Information System (INIS)

Randon, Rafael Henrique Nunes; Pereira, Yamba Carla Lara; Nascimento, Glauce Crivelaro do

2014-01-01

Professionals performing radiographic examinations are responsible for maintaining optimal image quality for accurate diagnoses. These professionals must competently execute techniques such as film manipulation and processing to minimize patient exposure to radiation. Improper performance by the professional and/or patient may result in a radiographic image of unsatisfactory quality that can also lead to a misdiagnosis and the development of an inadequate treatment plan. Currently, the most commonly performed extraoral examination is panoramic radiography. The invention of panoramic radiography has resulted in improvements in image quality with decreased exposure to radiation and at a low cost. However, this technique requires careful, accurate positioning of the patient's teeth and surrounding maxillofacial bone structure within the focal trough. Therefore, we reviewed the literature for the most common types of positioning errors in panoramic radiography to suggest the correct techniques. We would also discuss how to determine if the most common positioning errors occurred in panoramic radiography, such as in the positioning of the patient's head, tongue, chin, or body.

Tracking the position of the underwater robot for nuclear reactor inspection

International Nuclear Information System (INIS)

Jeo, J. W.; Kim, C. H.; Seo, Y. C.; Choi, Y. S.; Kim, S. H.

2003-01-01

The tracking procedure of the underwater mobile robot moving and submerging ahead to nuclear reactor vessel for visual inspection, which is required to find the foreign objects such as loose parts, is described. The yellowish underwater robot body tends to present a big contrast to boron solute cold water of nuclear reactor vessel, tinged with indigo by the Cerenkov effect. In this paper, we have found and tracked the positions of underwater mobile robot using the two color information, yellow and indigo. From the horizontal and vertical profiles analysis of the color image, the blue, green, and the gray component have the inferior signal-to-noise characteristics compared to the red component. The center coordinates extraction procedures areas follows. The first step is to segment the underwater robot body to cold water with indigo background. From the RGB color components of the entire monitoring image taken with the color CCD camera, we have selected the red color component. In the selected red image, we extracted the positions of the underwater mobile robot using the following process sequences; binarization, labelling, and centroid extraction techniques. In the experiment carried out at the Youngkwang unit 5 nuclear reactor vessel, we have tracked the center positions of the underwater robot submerged near the cold leg and the hot leg way, which is fathomed to 10m deep in depth. When the position of the robot vehicle fluctuates between the previous and the current image frame due to the flickering noise and light source, installed temporally in the bottom of the reactor vessel, we adaptively adjusted the ROI window. Adding the ROI windows of the previous frame to the current frame, and then setting up the ROI window of the next image frame, we can robustly track the positions of the underwater robot and control the target position's divergence. From these facts, we can conclude that using the red component from color camera is more efficient tracking method

Common positioning errors in panoramic radiography: A review

Energy Technology Data Exchange (ETDEWEB)

Randon, Rafael Henrique Nunes [Stomathology and Oral Diagnostic Program, School of Dentistry of Sao Paulo, University of Sao Paulo, Sao Paulo (Brazil); Pereira, Yamba Carla Lara [Biology Dental Buco Graduate Program, School of Dentistry of Piracicaba, University of Campinas, Piracicaba (Brazil); Nascimento, Glauce Crivelaro do [Psychobiology Graduate Program, School of Philosophy, Science and Literature of Ribeirao Preto, University of Sao Paulo, Ribeirao Preto (Brazil)

2014-03-15

Professionals performing radiographic examinations are responsible for maintaining optimal image quality for accurate diagnoses. These professionals must competently execute techniques such as film manipulation and processing to minimize patient exposure to radiation. Improper performance by the professional and/or patient may result in a radiographic image of unsatisfactory quality that can also lead to a misdiagnosis and the development of an inadequate treatment plan. Currently, the most commonly performed extraoral examination is panoramic radiography. The invention of panoramic radiography has resulted in improvements in image quality with decreased exposure to radiation and at a low cost. However, this technique requires careful, accurate positioning of the patient's teeth and surrounding maxillofacial bone structure within the focal trough. Therefore, we reviewed the literature for the most common types of positioning errors in panoramic radiography to suggest the correct techniques. We would also discuss how to determine if the most common positioning errors occurred in panoramic radiography, such as in the positioning of the patient's head, tongue, chin, or body.

78 FR 56174 - In-Core Thermocouples at Different Elevations and Radial Positions in Reactor Core

Science.gov (United States)

2013-09-12

... 52 [Docket No. PRM-50-105; NRC-2012-0056] In-Core Thermocouples at Different Elevations and Radial Positions in Reactor Core AGENCY: Nuclear Regulatory Commission. ACTION: Petition for rulemaking; denial...-core thermocouples at different elevations and radial positions throughout the reactor core to enable...

MDEP Design-Specific Common Position CP-APR1400WG-01. Common position addressing Fukushima Daiichi nuclear power plant accident

International Nuclear Information System (INIS)

2016-05-01

The MDEP APR1400 Working Group (APR1400WG) members consist of members from Republic of Korea, United Arab Emirates, and the United States. A main objectives of MDEP is to encourage convergence of code, standard and safety goals with exploring the opportunities for harmonization of regulatory practice and cooperation on safety review of APR-1400 specific designs. This common position addressing is aimed at sharing knowledge, information and experience on safety improvement related to lessons learned from the Fukushima Daiichi NPP Accident or Fukushima Daiichi NPP Accident-related issues amongst APR-1400 WG member states to achieve the MEDP goal. Because not all of these Regulators have completed the regulatory review of their APR1400 applications yet, this paper identifies common preliminary approaches to address potential safety improvements for APR1400 plants, as well as common general expectations for new nuclear power plants, as related to lessons learned from the Fukushima Daiichi NPP Accident or Fukushima Daiichi NPP Accident-related issues. While some asymmetry exists among those of three Regulators in terms of design, regulatory practice and licensing milestone sharing information and common understanding on post-Fukushima Daiichi NPP Accident enhancement would be promote resilient design for countering beyond design extreme external event like Fukushima Daiichi NPP nuclear disaster. This common position paper aims at identifying characteristics of post-Fukushima Daiichi NPP Accident enhancements putting in place by each country and setting common position to achieve balanced and harmonized APR-1400 design. After the safety reviews of the APR1400 design applications that are currently in review are completed, the regulators will update this paper to reflect their safety conclusions regarding the APR1400 design and how the design could be enhanced to address Fukushima Daiichi NPP Accident-related issues. The common preliminary approaches are organised into

Fuel assembly transfer and storage system for nuclear reactors

International Nuclear Information System (INIS)

Allain, Albert; Thomas, Claude.

1981-01-01

Transfer and storage system on a site comprising several reactors and at least one building housing the installations common to all these reactors. The system includes: transfer and storage modules for the fuel assemblies comprising a containment capable of containing several assemblies carried on a transport vehicle, a set of tracks for the modules between the reactors and the common installations, handling facilities associated with each reactor for moving the irradiated assemblies from the reactor to a transfer module placed in loading position on a track serving the reactor and conversely to move the new assemblies from the transfer module to the reactor, and at least one handling facility located in the common installation building for loading the modules with new assemblies [fr

MDEP AP1000WG Design-Specific Common Position CP-AP1000WG-02. Common position addressing Fukushima Daiichi NPP accident-related issues

International Nuclear Information System (INIS)

2016-09-01

A severe accident involving several units took place in Japan at Fukushima Daiichi nuclear power plant (NPP) in March 2011. The immediate cause of the accident was an earthquake followed by a tsunami coupled with inadequate provisions against the consequences of such events in the design. Opportunities to improve protection against a realistic design basis tsunami had not been taken. As a consequence of the tsunami, safety equipment and the related safety functions were lost at the plant, leading to core damage in three units and subsequently to large radioactive release. Several studies have already been performed to better understand the accident progression and detailed technical studies are still in progress in Japan and elsewhere. In the meantime, on-going studies on the behaviour of nuclear power plants in very severe situations, similar to Fukushima Daiichi, seek to identify potential vulnerabilities in plant design and operation; to suggest reasonably practicable upgrades; or to recommend enhanced regulatory requirements and guidance to address such situations. Likewise, agencies around the world that are responsible for regulating the design, construction and operation of AP1000 R plants are engaged in similar activities. The MDEP AP1000 R Working Group (AP1000 WG) members consist of members from Canada, China, the United Kingdom and the United States. Since the regulatory review of their AP1000 R applications have not been completed by all of these Countries yet, this paper identifies common preliminary approaches to address potential safety improvements for AP1000 R plants as related to lessons learned from the Fukushima Daiichi accident or Fukushima Daiichi-related issues. In seeking common position, regulators will provide input to this paper to reflect their safety conclusions regarding the AP1000 R design and how the design could be enhanced to address Fukushima Daiichi issues. The common preliminary approaches are organized into five sections

Impact of asymmetric lamp positioning on the performance of a closed-conduit UV reactor

Directory of Open Access Journals (Sweden)

Tipu Sultan

2017-06-01

Full Text Available Computational fluid dynamics (CFD analyses for the performance improvement of a closed-conduit ultraviolet (UV reactor were performed by changing the lamp positions from symmetric to asymmetric. The asymmetric lamp positioning can be useful for UV reactor design and optimization. This goal was achieved by incorporating the two performance factors, namely reduction equivalent dose (RED and system dose performance. Four cases were carried out for asymmetric lamp positioning within the UV reactor chamber and each case consisted of four UV lamps that were simulated once symmetrically and four times asymmetrically. The results of the four asymmetric cases were compared with the symmetric one. Moreover, these results were evaluated by using CFD simulations of a closed-conduit UV reactor. The fluence rate model, UVCalc3D was employed to validate the simulations results. The simulation results provide detailed information about the dose distribution, pathogen track modeling and RED. The RED value was increased by approximately 15% by using UVCalc3D fluence rate model. Additionally, the asymmetric lamp positioning of the UV lamps had more than 50% of the pathogens received a better and a higher UV dose than in the symmetric case. Consequently, the system dose performance was improved by asymmetric lamp positioning. It was concluded that the performance parameters (higher RED and system dose performance were improved by using asymmetric lamp positioning.

New approach for control rod position indication system for light water power reactor

International Nuclear Information System (INIS)

Bahuguna, Sushil; Dhage, Sangeeta; Nawaj, S.; Salek, C.; Lahiri, S.K.; Marathe, P.P.; Mukhopadhyay, S.; Taly, Y.K.

2015-01-01

Control rod position indication system is an important system in a nuclear power plant to monitor and display control rod position in all regimes of reactor operation. A new approach to design a control rod position indication system for sensing absolute position of control rod in Light Water Power Reactor has been undertaken. The proposed system employs an inductive type, hybrid measurement strategy providing both analog position as well as digital zone indication with built-in temperature compensation. The new design approach meets single failure criterion through redundancy in design without sacrificing measurement resolution. It also provides diversity in measurement technique by indirect position sensing based on analysis of drive coil current signature. Prototype development and qualification at room temperature of the control rod position indication system (CRPIS) has been demonstrated. The article presents the design philosophy of control rod position indication system, the new measurement strategy for sensing absolute position of control rod, position estimation algorithm for both direct and indirect sensing and a brief account associated processing electronics. (author)

Common safety approach for future pressurized water reactors in France and Germany

International Nuclear Information System (INIS)

Frisch, W.; Jahns, A.; Queniart, D.; Gros, G.

1995-01-01

In France and Germany all major activities related to future pressurized water reactors are now proceeding in a coordinated way among the two countries. The proposed paper covers this common development of a safety approach and particular technical safety objectives. The main topics of this document are presented in this paper, together with a rationale for the approach and the recommended technical principles. (K.A.). 5 refs., 1 fig

Final results of the cadmium and spectral ratios obtained inside of the fuel rod positioned in the central position of the IPEN/MB-01 nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Bitelli, Ulysses d' Utra; Mura, Luiz Ernesto C.; Santos, Diogo Feliciano dos, E-mail: ubitelli@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Lambiasi, Beatriz G.N. [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil)

2015-07-01

The spectral ratios are very important to determine some nuclear reactors parameters such as reaction rates, fuel lifetime, etc and some safety operational conditions. This study aims to determine the spectral ratios in 2 (two) spatial positions located inside the core of the Nuclear Reactor IPEN/MB-01. These places are at the central position of the nuclear reactor core in an asymptotic neutron flux region. The experiment consists in inserting different activation foil detectors inside an experimental fuel rod. The experimental rod is assembled at the central position of the reactor core. Activation neutron foil detectors of different elements such as {sup 197}Au, {sup 238}U, {sup 45}Sc, {sup 58}Ni, {sup 24}Mg, {sup 47}Ti and {sup 115m}In were used to cover a large range of neutron spectrum. Saturation activity per target nucleus was obtained by gamma spectrometry using a HPGe system. The experimental cadmium ratios compared with values computed by MCNP-4C code show good agreement. (author)

Franco-German nuclear cooperation: from the 'common product' to the first European pressurized water reactor

International Nuclear Information System (INIS)

Vignon, D.

1999-01-01

It has now been 10 years since Framatome and Siemens decided to collaborate on the design and sales of an advanced nuclear power plant (NPP) model based on pressurized water reactor (PWR) technology. Originally called the 'common product', this model was renamed the European pressurized water reactor when Electricite de France (EDF) and the German electric utilities joined this cooperative development effort in 1992. Since the beginning, this cooperation has been formalized in the framework of an agreement that led to the founding of a joint and equally owned subsidiary, Nucler Power International (NPI), which is reponsible for leading the development of the new model and later handling its export sales

Conceptual Design of Angular Position Detector for Control Element Drive Mechanism of Small and Medium Reactor

International Nuclear Information System (INIS)

Yu, Je-Yong; Huh, Hyung; Kim, Ji-Ho; Choi, Suhn

2007-01-01

When the small and medium reactor is designed with a soluble boron free operation and nuclear heating for the reactor start-up, the design features require a Control Element Drive Mechanism (CEDM) to have a fine-step movement capability as well as a high reliability for a fine reactivity control. Also the reliability and accuracy of the information for the control rod position is important to the reactor safety as well as to design of the core protection system. The position signal of control rod is classified as a Class 1E because the rod position signal is used in the safety related systems. Therefore it will be separated from the control systems to the extent that a failure of any single control system component of a channel and shall have sufficient independence, redundancy, and testability to perform its safety functions assuming a single failure. The position indicator is composed of a permanent magnet, reed switches and a voltage divider. Four independent position indicators around the upper pressure housing provide an indication of the position of a control rod comprising of a permanent magnet with a magnetic field concentrator which moves with the extension shaft connected to the control rod. The zigzag arranged reed switches are positioned along a line parallel to the path of the movement of the permanent magnet and it is activated selectively when the permanent magnet passes by. A voltage divider electrically connected to the reed switches provides a signal commensurate with the position of the control rod. The signal may then be transmitted to a position indicating device. But position indicator can not recognize the malfunction of the rotary step motor of CEDM instantly because its signal output is changed after the control rod moves more than a distance of reed switch interval

Common Positioning Errors in Digital Panoramic Radiographies Taken In Mashhad Dental School

Directory of Open Access Journals (Sweden)

Ali Bagherpour

2018-06-01

Full Text Available Introduction: The present study was aimed at evaluating common positioning errors on panoramic radiographs taken in the Radiology Department of Mashhad Dental School. Materials and methods: The study sample included 1,990 digital panoramic radiographs taken in the Radiology Department of Mashhad Dental School by a Planmeca Promax (Planmeca Oy, Helsinki, Finland, during a 2-year period (2010–2012. All radiographs, according to dentition and sex, were evaluated for positioning errors. Results: There were 1,927 (96.8% panoramic radiographs with one or more errors. While the number of errors in each image varied between one and five, most images had one error (48.4%. The most common error was that the tongue was not in contact with the hard palate (94.8%. "Open lips" was an error not seen in any patients. Conclusions:positioning errors are common in panoramic radiographies. The most common error observed in this study was a failure to place the tongue on the palate. This error and the other errors reported in this study can be reduced by training the technicians and spending little more time for patient positioning and more effective communication with the patients.

Sensory systems for a control rod position using reed switches for the integral reactor

International Nuclear Information System (INIS)

Yu, J. Y.; Choi, S.; Kim, J. H.; Lee, D. J.

2007-01-01

The system-integrated modular advanced reactor (SMART) currently under development at the Korea Atomic Energy Research Institute is being designed with a soluble boron free operation and the use of nuclear heating for the reactor start-up. These design features require a Control Element Drive Mechanism (CEDM) for the SMART to have a fine-step movement capability as well as a high reliability for a fine reactivity control. Also the reliability and accuracy of the information for the control rod position is very important to the reactor safety as well as the design of the core protection system. The position indicator is classified as a Class 1E component because the rod position signal of the position indicator is used in the safety related systems. Therefore it will be separated from the control systems to the extent that a failure of any single control system component of a channel and shall have sufficient independence, redundancy, and testability to perform its safety functions assuming a single failure. The position indicator is composed of a permanent magnet, reed switches and a voltage divider. Four independent position indicators around the upper pressure housing provide an indication of the position of a control rod comprising of a permanent magnet with a magnetic field concentrator which moves with the extension shaft connected to the control rod. The zigzag arranged reed switches are positioned along a line parallel to the path of the movement of the permanent magnet and it is activated selectively when the permanent magnet passes by. A voltage divider electrically connected to the reed switches provides a signal commensurate with the position of the control rod. The signal may then be transmitted to a position indicating device. In order to monitor the operating condition of the rotary step motor of CEDM, the angular position detector was installed at the top of the rotary step motor by means of connecting between the planetary gear and the rotating

Improved SAFARI-1 research reactor irradiation position modeling in OSCAR-3 code system

International Nuclear Information System (INIS)

Moloko, L. E.; Belal, M. G. A. H.

2009-01-01

The demand on the availability of irradiation positions in the SAFARI-1 reactor is continuously increasing due to the commercial pressure to produce isotopes more efficiently. This calls for calculational techniques and modeling methods to be improved regularly to optimize irradiation services. The irradiation position models are improved using the OSCAR-3 code system, and results are compared to experimental measurements. It is concluded that the irradiation position models are essential if realistic core follow and reload studies are to be performed and most importantly, for the realization of improved agreement between experimental data and calculated results. (authors)

Plasma position control in a tokamak reactor around ignition

International Nuclear Information System (INIS)

Carretta, U.; Minardi, E.; Bacelli, N.

1986-01-01

Plasma position control in a tokamak reactor in the phase approaching ignition is closely related to burn control. If ignited burn corresponds to a thermally unstable situation the plasma becomes sensitive to the thermal instability already in the phase when ignition is approached so that the trajectory in the position-pressure (R,p) space becomes effectively unpredictable. For example, schemes involving closed cycles around ignition can be unstable in the heating-cooling phases, and the deviations may be cumulative in time. Reliable plasma control in pressure-position (p, R) space is achieved by beforehand constraining the p, R trajectory rigidly with suitable feedback vertical field stabilization, which is to be established already below ignition. A scheme in which ignition is approached in a stable and automatic way by feedback stabilization on the vertical field is proposed and studied in detail. The values of the gain coefficient ensuring stabilization and the associated p and R excursions are discussed both analytically, with a 0-D approximation including non-linear effects, and numerically with a 1-D code in cylindrical geometry. Profile effects increase the excursions, in particular above ignition. (author)

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

International Nuclear Information System (INIS)

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

2007-01-01

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

Cross-flow electrochemical reactor cells, cross-flow reactors, and use of cross-flow reactors for oxidation reactions

Science.gov (United States)

Balachandran, Uthamalingam; Poeppel, Roger B.; Kleefisch, Mark S.; Kobylinski, Thaddeus P.; Udovich, Carl A.

1994-01-01

This invention discloses cross-flow electrochemical reactor cells containing oxygen permeable materials which have both electron conductivity and oxygen ion conductivity, cross-flow reactors, and electrochemical processes using cross-flow reactor cells having oxygen permeable monolithic cores to control and facilitate transport of oxygen from an oxygen-containing gas stream to oxidation reactions of organic compounds in another gas stream. These cross-flow electrochemical reactors comprise a hollow ceramic blade positioned across a gas stream flow or a stack of crossed hollow ceramic blades containing a channel or channels for flow of gas streams. Each channel has at least one channel wall disposed between a channel and a portion of an outer surface of the ceramic blade, or a common wall with adjacent blades in a stack comprising a gas-impervious mixed metal oxide material of a perovskite structure having electron conductivity and oxygen ion conductivity. The invention includes reactors comprising first and second zones seprated by gas-impervious mixed metal oxide material material having electron conductivity and oxygen ion conductivity. Prefered gas-impervious materials comprise at least one mixed metal oxide having a perovskite structure or perovskite-like structure. The invention includes, also, oxidation processes controlled by using these electrochemical reactors, and these reactions do not require an external source of electrical potential or any external electric circuit for oxidation to proceed.

Possibilities of achieving non-positive void reactivity effect in fast sodium-cooled reactors with increased self-protection

International Nuclear Information System (INIS)

Alekseev, P.N.; Zverkov, Yu.A.; Morozov, A.G.; Orlov, V.V.; Slesarev, I.S.; Subbotin, S.A.

1989-01-01

The problems of self-protection inhancement for the liquid-metal cooled fast reactors with intra-assembly heterogeneity of the core are studied. Possible approaches to arrangement of such reactors with various powers characterized by high levels of coolant natural circulation, minimum reactivity changes during fuel burn-up and non-positive void effect of reactivity are found. 10 refs.; 11 figs

An analytical approach to the positive reactivity void coefficient of TRIGA Mark-II reactor

International Nuclear Information System (INIS)

Edgue, Erdinc; Yarman, Tolga

1988-01-01

Previous calculations of reactivity void coefficient of I.T.U. TRIGA Mark-II Reactor was done by the second author et al. The theoretical predictions were afterwards, checked in this reactor experimentally. In this work an analytical approach is developed to evaluate rather quickly the reactivity void coefficient of I.T.U. TRIGA Mark-II, versus the size of the void inserted into the reactor. It is thus assumed that the reactor is a cylindrical, bare nuclear system. Next a belt of water of 2πrΔrH is introduced axially at a distance r from the center line of the system. r here, is the thickness of the belt, and H is the height of the reactor. The void is described by decreasing the water density in the belt region. A two group diffusion theory is adopted to determine the criticality of our configuration. The space dependency of the group fluxes are, thereby, assumed to be J 0 (2.405 r / R) cos (π Z / H), the same as that associated with the original bare reactor uniformly loaded prior to the change. A perturbation type of approach, thence, furnishes the effect of introducing a void in the belt region. The reactivity void coefficient can, rather surprisingly, be indeed positive. To our knowledge, this fact had not been established, by the supplier. The agreement of our predictions with the experimental results is good. (author)

Positioning calibration apparatus for transducers employed in nuclear reactor vessel inspection apparatus

International Nuclear Information System (INIS)

Elsner, H.J.

1981-01-01

The invention provides a calibration apparatus suitable for verifying the position and orientation of transducers used in reactor vessel ultrasonic inspection. The apparatus includes moveable mounting means which secures a transducer within the tank in its normal inspection orientation. A drive is also provided for moving the transducer in the tank relative to a target. The target is slidably positioned in the tank at a distance from the transducer which is selected to avoid the distortion effects in the near field of the transducer. The drive mechanism may be provided with graduated indicia of travel, or a scale may be affixed to the side of the tank. (L.L.)

Licensing of safety critical software for nuclear reactors. Common position of seven European nuclear regulators and authorised technical support organisations

International Nuclear Information System (INIS)

2007-01-01

The major result of the work is the identification of consensus and common technical positions on a set of important licensing issues raised by the design and operation of computer-based systems used in Nuclear Power Plants for safety functions. The purpose is to introduce greater consistency and more mutual acceptance into current practices. To achieve these common positions, detailed consideration was paid to the licensing approaches followed in the different countries represented by the experts of the task force. The report is intended to be useful: - to coordinate regulators' and safety experts' technical viewpoints in the design of regulators' national policies and in revisions of guidelines; - as a reference in safety cases and demonstrations of safety of software based systems; - as guidance for system design specifications by manufacturers and major I and C suppliers on the international market. The task force decided at an early stage to focus attention on computer based systems used in Nuclear Power Plants for the implementation of safety functions; namely, those systems classified by the IAEA as 'Safety Systems'. Therefore, recommendations of this report - except those of chapter 1.11 - primarily address 'safety systems' and not 'safety related systems'. It was felt that the most difficult aspects of the licensing of digital programmable systems are rooted in the specific properties of the technology. The objective was therefore to delineate practical and technical licensing guidance, rather than discussing or proposing basic principles or requirements. The design requirements and the basic principles of nuclear safety in force in each member state are assumed to remain applicable. This report represents the consensus view achieved by the experts who contributed to the task force. It is the result of what was at the time of its initiation a first attempt at the international level to achieve consensus among nuclear regulators on practical methods for

Dynamics and Control of Chemical Reactors-Selectively Surveyed

DEFF Research Database (Denmark)

Jørgensen, S. B.; Jensen, N.

1989-01-01

The chemical reactor or bioreactor is physically at a central position in a process, and often with a decisive role on the overall technical and economical performance. Even though application of feedback control on reactors is gaining momentum and on-line optimization has been implemented....... For bioreactors the theory and practice of reactor design, dynamics and control have to be adapted to the peculiarities of the biological catalysts. Enzymes, the protein catalysts, are the simplest ones, which have many common features with chemical catalysts. The living cells are much more complex, these growing...... in industry, many reactor control problems are still left unsolved or only partly solved using open loop strategies where disturbance rejection and model inaccuracies have to be handled through manual reactor control and feedback control of raw material preprocessing and product purification operations...

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...

Positioning calibration apparatus for transducers employed in nuclear reactor vessel inspection apparatus

International Nuclear Information System (INIS)

Elsner, H.J.

1979-01-01

Calibration apparatus for verifying the position and orientation of transducers used in nuclear reactor vessel inspection apparatus is disclosed. A tank, filled with water, the operating inspection medium, is fitted with a movable mounting assembly adapted to securely accommodate a transducer and the mounting assembly in which it is normally secured during an inspection procedure. The tank is also provided with a slidably mounted target positioned therein at a predetermined distance from the target which is selected to avoid the distortion effects in the near field of the transducer response. The calibration apparatus can be used to check the normal transducer mounting for either perpendicularity or angular orientation by moving the tank's mounting assembly via a lead screw with which it is threadingly engaged. 6 claims

Position-dependency of Fuel Pin Homogenization in a Pressurized Water Reactor

Energy Technology Data Exchange (ETDEWEB)

Heo, Woong; Kim, Yonghee [Korea Advanced Institute of Science and Technolgy, Daejeon (Korea, Republic of)

2016-05-15

By considering the multi-physics effects more comprehensively, it is possible to acquire precise local parameters which can result in a more accurate core design and safety assessment. A conventional approach of the multi-physics neutronics calculation for the pressurized water reactor (PWR) is to apply nodal methods. Since the nodal methods are basically based on the use of assembly-wise homogenized parameters, additional pin power reconstruction processes are necessary to obtain local power information. In the past, pin-by-pin core calculation was impractical due to the limited computational hardware capability. With the rapid advancement of computer technology, it is now perhaps quite practical to perform the direct pin-by-pin core calculation. As such, fully heterogeneous transport solvers based on both stochastic and deterministic methods have been developed for the acquisition of exact local parameters. However, the 3-D transport reactor analysis is still challenging because of the very high computational requirement. Position-dependency of the fuel pin homogenized cross sections in a small PWR core has been quantified via comparison of infinite FA and 2-D whole core calculations with the use of high-fidelity MC simulations. It is found that the pin environmental affect is especially obvious in FAs bordering the baffle reflector regions. It is also noted that the downscattering cross section is rather sensitive to the spectrum changes of the pins. It is expected that the pinwise homogenized cross sections need to be corrected somehow for accurate pin-by-pin core calculations in the peripheral region of the reactor core.

Biological samples positioning device for irradiations on a radial channel at the nuclear research reactor

International Nuclear Information System (INIS)

Rodriguez Gual, Maritza; Mas Milian, Felix; Deppman, Airton; Pinto Coelho, Paulo Rogerio

2010-01-01

For the demand of an experimental device for biological samples positioning system for irradiations on a radial channel at the nuclear research reactor in operation was constructed and started up a device for the place and remove of the biological samples from the irradiation channels without interrupting the operation of the reactor. The economical valuations are effected comparing with another type of device with the same functions. This work formed part of an international project between Cuba and Brazil that undertook the study of the induced damages by various types of ionizing radiation in DNA molecules. Was experimentally tested the proposed solution, which demonstrates the practical validity of the device. As a result of the work, the experimental device for biological samples irradiations are installed and operating in the radial beam hole No3(BH3) for more than five years at the IEA-R1 Brazilian research reactor according to the solicited requirements the device. The designed device increases considerably the type of studies can be conducted in this reactor. Its practical application in research taking place in that facility, in the field of radiobiology and dosimetry, and so on is immediate

Gas-cooled reactor programs. Fuel-management positioning and accounting module: FUELMANG Version V1. 11, September 1981

Energy Technology Data Exchange (ETDEWEB)

Medlin, T.W.; Hill, K.L.; Johnson, G.L.; Jones, J.E.; Vondy, D.R.

1982-01-01

This report documents the code module FUELMANG for fuel management of a reactor. This code may be used to position fuel during the calculation of a reactor history, maintain a mass balance history of the fuel movement, and calculate the unit fuel cycle component of the electrical generation cost. In addition to handling fixed feed fuel without recycle, provision has been made for fuel recycle with various options applied to the recycled fuel. A continuous fueling option is also available with the code. A major edit produced by the code is a detailed summary of the mass balance history of the reactor and a fuel cost analysis of that mass balance history. This code is incorporated in the system containing the VENTURE diffusion theory neutronics code for routine use. Fuel movement according to prescribed instructions is performed without the access of additional user input data during the calculation of a reactor operating history. Local application has been primarily for analysis of the performance of gas-cooled thermal reactor core concepts.

Gas-cooled reactor programs. Fuel-management positioning and accounting module: FUELMANG Version V1.11, September 1981

International Nuclear Information System (INIS)

Medlin, T.W.; Hill, K.L.; Johnson, G.L.; Jones, J.E.; Vondy, D.R.

1982-01-01

This report documents the code module FUELMANG for fuel management of a reactor. This code may be used to position fuel during the calculation of a reactor history, maintain a mass balance history of the fuel movement, and calculate the unit fuel cycle component of the electrical generation cost. In addition to handling fixed feed fuel without recycle, provision has been made for fuel recycle with various options applied to the recycled fuel. A continuous fueling option is also available with the code. A major edit produced by the code is a detailed summary of the mass balance history of the reactor and a fuel cost analysis of that mass balance history. This code is incorporated in the system containing the VENTURE diffusion theory neutronics code for routine use. Fuel movement according to prescribed instructions is performed without the access of additional user input data during the calculation of a reactor operating history. Local application has been primarily for analysis of the performance of gas-cooled thermal reactor core concepts

Nuclear reactor, reactor core thereof, and device for constituting the reactor

International Nuclear Information System (INIS)

Takiyama, Masashi.

1994-01-01

A reactor core is constituted by charging coolants (light water) in a reactor pressure vessel and distributing fuel assemblies, reflecting material sealing pipes, moderator (heavy water and helium gas) sealing pipes, and gas sealing pipes therein. A fuel guide tube is surrounded by a cap and the gap therebetween is made hollow and filled with coolant steams. The cap is supported by a baffle plate. The moderator sealing pipe is disposed in a flow channel of coolants in adjacent with the cap. The position of the moderator sealing tube in the reactor core is controlled by water stream from a hydraulic pump with a guide tube extending below the baffle plate being as a guide. Then, the position of the moderator sealing tube is varied to conduct power control, burnup degree compensation, and reactor shut down. With such procedures, moderator cooling facility is no more necessary to simplify the structure. Further, heat generated from the moderator is transferred to the coolants thereby improving heat efficiency of the reactor. (I.N.)

Estimating Common Pedestrian Routes through Indoor Path Networks using Position Traces

DEFF Research Database (Denmark)

Prentow, Thor Siiger; Blunck, Henrik; Grønbæk, Kaj

2014-01-01

routes between locations. These methods are sufficiently efficient to provide common routes based on real-time data from thousands of devices simultaneously. Furthermore, we show that the methods operate robustly even on basis of noisy and coarse-grained position estimates as provided by large......Abstract—Accurate information about how people commonly travel in a given large-scale building environment and which routes they take for given start and destination points is essential for applications such as indoor navigation, route prediction, and mobile work planning and logistics...

Positive Orientation-a Common Base for Hedonistic and Eudemonistic Happiness?

Science.gov (United States)

Oleś, Piotr; Jankowski, Tomasz

2018-01-01

Positive orientation (PO) is proposed as a common base for hedonistic and eudemonistic senses of happiness. PO involves a tendency to formulate positive judgments concerning the self, one's personal life, and the future. Previously, PO had been investigated in the context of the hedonistic approach to well-being. In this article, we tested a broader understanding of PO, which is conceptualized, here, as a latent factor underlying variables that exemplify hedonistic and eudemonistic view on happiness. Using two samples ( N  = 159 and N  = 200), we tested three models of PO extended to include various measures of meaning of life. The extended models fitted the data well. Results suggest that PO can be a general factor that is the basis for integrating two aspects of well-being: searching for positivity and pleasure, as well as striving for meaning.

Development of an asymmetric multiple-position neutron source (AMPNS) method to monitor the criticality of a degraded reactor core

International Nuclear Information System (INIS)

Kim, S.S.; Levine, S.H.

1985-01-01

An analytical/experimental method has been developed to monitor the subcritical reactivity and unfold the k/sub infinity/ distribution of a degraded reactor core. The method uses several fixed neutron detectors and a Cf-252 neutron source placed sequentially in multiple positions in the core. Therefore, it is called the Asymmetric Multiple Position Neutron Source (AMPNS) method. The AMPNS method employs nucleonic codes to analyze the neutron multiplication of a Cf-252 neutron source. An optimization program, GPM, is utilized to unfold the k/sub infinity/ distribution of the degraded core, in which the desired performance measure minimizes the error between the calculated and the measured count rates of the degraded reactor core. The analytical/experimental approach is validated by performing experiments using the Penn State Breazeale TRIGA Reactor (PSBR). A significant result of this study is that it provides a method to monitor the criticality of a damaged core during the recovery period

The CAREM reactor and present currents in reactor design

International Nuclear Information System (INIS)

Ordonez, J.P.

1990-01-01

INVAP has been working on the CAREM project since 1983. It concerns a very low power reactor for electrical energy generation. The design of the reactor and the basic criteria used were described in 1984. Since then, a series of designs have been presented for reactors which are similar to CAREM regarding the solutions presented to reduce the chance of major nuclear accidents. These designs have been grouped under different names: Advanced Reactors, Second Generation Reactors, Inherently Safe Reactors, or even, Revolutionary Reactors. Every reactor fabrication firm has, at least, one project which can be placed in this category. Presently, there are two main currents of Reactor Design; Evolutionary and Revolutionary. The present work discusses characteristics of these two types of reactors, some revolutionary designs and common criteria to both types. After, these criteria are compared with CAREM reactor design. (Author) [es

FBR type reactor

International Nuclear Information System (INIS)

Hayase, Tamotsu.

1991-01-01

The present invention concerns an FBR type reactor in which transuranium elements are eliminated by nuclear conversion. There are loaded reactor core fuels being charged with mixed oxides of plutonium and uranium, and blanket fuels mainly comprising depleted uranium. Further, liquid sodium is used as coolants. As transuranium elements, isotope elements of neptunium, americium and curium contained in wastes taken out from light water reactors or the composition thereof are used. The reactor core comprises a region with a greater mixing ratio and a region with a less mixing ratio of the transuranium elements. The mixing ratio of the transuranium elements is made greater for the fuels in the reactor core region at the boundary with the blanket of great neutron leakage. With such a constitution, since the positive reactivity value at the reactor core central portion is small in the Na void reactivity distribution in the reactor core, the positive reactivity is small upon Na boiling in the reactor core central region upon occurrence of imaginable accident, to attain reactor safety. (I.N.)

Maximization of a reflected slab type reactor, and with double criticality as function of the position of the interface core-reflector

International Nuclear Information System (INIS)

Ierardi, Maria Clara Filippini

1980-01-01

Using optimal control techniques, it is obtained the fuel distribution which maximizes the power of an infinite, reflected slab nuclear reactor. The reactor is permissible power density and the maximum permissible fuel introduces the reflector into the reactor in a non-conventional way. Numerical results for both two and four energy groups are presented. It is observed a double criticality depending on the position of the core-reflector interface, however described by a diffusion theory model. Both the maximum concentration are the only restraints. (author)

Solving the uncommon reactor core neutronics problems

International Nuclear Information System (INIS)

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

1980-01-01

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

Positioning means for circumferentially locating inspection apparatus in a nuclear reactor vessel

International Nuclear Information System (INIS)

Burns, D.C.

1979-01-01

Positioning means for locating inspection apparatus used to volumetrically examine a nuclear reactor vessel is disclosed. The positioning means is provided with a support ring having an annular key positioned longitudinally about its periphery. Three support legs are attached to the support ring by brackets adapted to fit the annular key. The support ring also carries three guide stud bushings which are movably mounted by clamps adapted to engage the support ring key. Prior to lowering the inspection apparatus into the vessel, the guide stud bushings are each moved to a point of alignment with one of three guide studs extending upwardly from the vessel. After alignment has been verified, the guide stud bushings are clamped in position. The inspection apparatus is lowered towards its fully seated position within the vessel and is coarsely circumferentially positioned with by the engagement of the guide studs within the guide stud bushings. A fine degree of circumferential positioning is achieved by providing a specially configured shoe for one of the support legs. With the core barrel internals in, the special shoe is adapted to key onto a core barrel pin the exact location of which is known. With the core barrel internals removed, the special shoe is adapted to place a locating key into a notch in a vessel flange, the location of which is known. As the inspection apparatus is lowered into its fully seated position, exact circumferential positioning with respect to the vessel is achieved. The other support legs rest on an inner circumferential flange so that no portion of the inspection apparatus touches or threatens the vessel's top flange. 19 claims

New Design for the Inductive Position Sensor of the CAREM Reactor Control Bars

International Nuclear Information System (INIS)

Esparza, Daniel; D'Ovidio, Claudio; Taglialavore, Eduardo

2000-01-01

We describe the new design of the sensor used for determining the position of the control bars in the CAREM reactor.It presently operates under real 'cold' conditions, being under progress the final selection of materials for operation under the 'hot' condition of the reactor.The actual design is a modification of the previous one and is based on the same principle.A solenoid is placed on the outer side of the mechanism that moves the control bar, which has some part made of a magnetic material, and the variation of an electrical property of the bobbin with the movement of this magnetic piece is studied.This new design was proposed both to increase the output voltage and simplify the electronics and construction of the sensor.The output voltage is lineal with the bar position, with a correlation coefficient R = 0.9997, a sensibility of 43 % and a resolution better than 1 in 1000.The output sensibility was improved in almost three orders of magnitude, from 1.204 μV/mm to 0.924 mV/mm. Considering that the typical electric noise was ±1 mV RMS, we are able to measure the mm in the total bar excursion of 1400 mm. It is to be noticed that we obtained a resolution 10 times higher than the required: half step of the mechanism, that is ±10 mm. Both the employed electronics and the bobbin construction were markedly simplified

Reactor core and initially loaded reactor core of nuclear reactor

International Nuclear Information System (INIS)

Koyama, Jun-ichi; Aoyama, Motoo.

1989-01-01

In BWR type reactors, improvement for the reactor shutdown margin is an important characteristic condition togehter with power distribution flattening . However, in the reactor core at high burnup degree, the reactor shutdown margin is different depending on the radial position of the reactor core. That is , the reactor shutdown margin is smaller in the outer peripheral region than in the central region of the reactor core. In view of the above, the reactor core is divided radially into a central region and as outer region. The amount of fissionable material of first fuel assemblies newly loaded in the outer region is made less than the amount of the fissionable material of second fuel assemblies newly loaded in the central region, to thereby improve the reactor shutdown margin in the outer region. Further, the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower portion of the first fuel assemblies is made smaller than the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower region of the second fuel assemblies, to thereby obtain a sufficient thermal margin in the central region. (K.M.)

Position adopted by the government about the safety options of the EPR reactor project; Prise de position du gouvernement concernant les options de surete du projet de reacteur EPR

Energy Technology Data Exchange (ETDEWEB)

NONE

2004-10-01

On September 28, 2004, on behalf of the French ministers in charge of nuclear safety, the general director of nuclear safety and radiation protection addressed to the president of Electricite de France (EdF) a letter presenting the government's position about the safety options of the EPR (European Pressurized Reactor) project. On the basis of the examination carried out by the nuclear safety authority (ASN) and by the permanent group of reactor experts, the government has considered these options as satisfactory with respect to the safety improvement objectives. Therefore, the government requested EdF to comply with these technical rules for any future reactor development. This dossier includes: the letter of the government, the technical directives for the design and construction of the next generation of PWR-type reactors, the technical rules relative to the design of the main primary and secondary coolant circuits of PWR-type reactors, and the technical file about the safety of the EPR project reprinted from the 2003 report of nuclear safety and radiation protection authority. (J.S.)

Human rights: common meaning and differences in positioning

Directory of Open Access Journals (Sweden)

Willem Doise

Full Text Available Human rights are defined as normative social representations embedded in institutional juridical definitions. Research findings show that human rights can be studied as normative social representations implying a degree of common understanding across cultures together with organized differences within and between cultures. Important factors in modulating individual positioning in the realm of human rights are experiences of social conflict and injustice, beliefs about the efficiency of various social actors to have rights enforced and attitudes of liberalism or collectivism. On the other hand, an ethnocentric use of human rights is well documented and has been experimentally studied. Generally, concerns about these rights expressed by citizens of Western countries become much stronger when non-Western countries are involved, whereas violations of these rights in their own country are often not severely condemned.

Irradiation positions for fission-track dating in the University of Pavia TRIGA Mark II nuclear reactor

International Nuclear Information System (INIS)

Oddone, Massimo; Meloni, Sandro; Balestrieri, Maria Laura; Bigazzi, Giulio

2002-01-01

An irradiation position arranged is described in the present paper for fission-track dating in the Triga Mark II reactor of the University of Pavia. Fluence values determined using the NIST glass standard SRM 962a for fission-track dating and the traditional metal foils are compared. Relatively good neutron thermalization (φ th /φ f = 0.956) and lack of significant fluence spatial gradients are good factors for fission-track dating. Finally, international age standards (or putative age standards) irradiated in this new position yielded results consistent with independent reference ages. (author)

Common cause failures of reactor pressure components

International Nuclear Information System (INIS)

Mankamo, T.

1978-01-01

The common cause failure is defined as a multiple failure event due to a common cause. The existence of common failure causes may ruin the potential advantages of applying redundancy for reliability improvement. Examples relevant to large mechanical components are presented. Preventive measures against common cause failures, such as physical separation, equipment diversity, quality assurance, and feedback from experience are discussed. Despite the large number of potential interdependencies, the analysis of common cause failures can be done within the framework of conventional reliability analysis, utilizing, for example, the method of deriving minimal cut sets from a system fault tree. Tools for the description and evaluation of dependencies between components are discussed: these include the model of conditional failure causes that are common to many components, and evaluation of the reliability of redundant components subjected to a common load. (author)

Development of an Advanced Digital Reactor Protection System Using Diverse Dual Processors to Prevent Common-Mode Failure

International Nuclear Information System (INIS)

Shin, Hyun Kook; Nam, Sang Ku; Sohn, Se Do; Chang, Hoon Seon

2003-01-01

The advanced digital reactor protection system (ADRPS) with diverse dual processors has been developed to prevent common-mode failure (CMF). The principle of diversity is applied to both hardware design and software design. For hardware diversity, two different types of CPUs are used for the bistable processor and local coincidence logic (LCL) processor. The Versa Module Eurocard-based single board computers are used for the CPU hardware platforms. The QNX operating system and the VxWorks operating system were selected for software diversity. Functional diversity is also applied to the input and output modules, and to the algorithm in the bistable processors and LCL processors. The characteristics of the newly developed digital protection system are described together with the preventive capability against CMF. Also, system reliability analysis is discussed. The evaluation results show that the ADRPS has a good preventive capability against the CMF and is a highly reliable reactor protection system

Nuclear reactor

International Nuclear Information System (INIS)

Garabedian, G.

1988-01-01

A liquid reactor is described comprising: (a) a reactor vessel having a core; (b) one or more satellite tanks; (c) pump means in the satellite tank; (d) heat exchanger means in the satellite tank; (e) an upper liquid metal conduit extending between the reactor vessel and the satellite tank; (f) a lower liquid metal duct extending between the reactor vessel and satellite tanks the upper liquid metal conduit and the lower liquid metal duct being arranged to permit free circulation of liquid metal between the reactor vessel core and the satellite tank by convective flow of liquid metal; (g) a separate sealed common containment vessel around the reactor vessel, conduits and satellite tanks; (h) the satellite tank having space for a volume of liquid metal that is sufficient to dampen temperature transients resulting from abnormal operating conditions

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

International Nuclear Information System (INIS)

Hirayama, Shoichi; Kawabe, Takaya

1995-01-01

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

Fast-acting nuclear reactor control device

International Nuclear Information System (INIS)

Kotlyar, O.M.; West, P.B.

1993-01-01

A fast-acting nuclear reactor control device is described for controlling a safety control rod within the core of a nuclear reactor, the reactor controlled by a reactor control system, the device comprising: a safety control rod drive shaft and an electromagnetic clutch co-axial with the drive shaft operatively connected to the safety control rod for driving and positioning the safety control rod within or without the reactor core during reactor operation, the safety rod being oriented in a substantially vertical position to allow the rod to fall into the reactor core under the influence of gravity during shutdown of the reactor; the safety control rod drive shaft further operatively connected to a hydraulic pump such that operation of the drive shaft simultaneously drives and positions the safety control rod and operates the hydraulic pump such that a hydraulic fluid is forced into an accumulator, filling the accumulator with oil for the storage and supply of primary potential energy for safety control rod insertion such that the release of potential energy in the accumulator causes hydraulic fluid to flow through the hydraulic pump, converting the hydraulic pump to a hydraulic motor having speed and power capable of full length insertion and high speed driving of the safety control rod into the reactor core; a solenoid valve interposed between the hydraulic pump and the accumulator, said solenoid valve being a normally open valve, actuated to close when the safety control rod is out of the reactor during reactor operation; and further wherein said solenoid opens in response to a signal from the reactor control system calling for shutdown of the reactor and rapid insertion of the safety control rod into the reactor core, such that the opening of the solenoid releases the potential energy in the accumulator to place the safety control rod in a safe shutdown position

Reactor container

International Nuclear Information System (INIS)

Naruse, Yoshihiro.

1990-01-01

The thickness of steel shell plates in a reactor container embedded in sand cussions is monitored to recognize the corrosion of the steel shell plates. That is, the reactor pressure vessel is contained in a reactor container shell and the sand cussions are disposed on the lower outside of the reactor container shell to elastically support the shell. A pit is disposed at a position opposing to the sand cussions for measuring the thickness of the reactor container shell plates. The pit is usually closed by a closing member. In the reactor container thus constituted, the closing member can be removed upon periodical inspection to measure the thickness of the shell plates. Accordingly, the corrosion of the steel shell plates can be recognized by the change of the plate thickness. (I.S.)

Pool-type reactor

International Nuclear Information System (INIS)

Hopkins, S.R.

1977-01-01

This invention relates to a pool nuclear reactor fitted with a perfected system to raise the buckets into a vertical position at the bottom of a channel. This reactor has an inclined channel to guide a bucket containing a fuel assembly to introduce it into the reactor jacket or extract it therefrom and a damper at the bottom of the channel to stop the drop of the bucket. An upright vertically movable rod has a horizontally articulated arm with a hook. This can pivot to touch a radial lug on the bucket and pivot the bucket around its base in a vertical position, when the rod moves up [fr

Anticipation of high arousal aversive and positive movie clips engages common and distinct neural substrates.

Science.gov (United States)

Greenberg, Tsafrir; Carlson, Joshua M; Rubin, Denis; Cha, Jiook; Mujica-Parodi, Lilianne

2015-04-01

The neural correlates of anxious anticipation have been primarily studied with aversive and neutral stimuli. In this study, we examined the effect of valence on anticipation by using high arousal aversive and positive stimuli and a condition of uncertainty (i.e. either positive or aversive). The task consisted of predetermined cues warning participants of upcoming aversive, positive, 'uncertain' (either aversive or positive) and neutral movie clips. Anticipation of all affective clips engaged common regions including the anterior insula, dorsal anterior cingulate cortex, thalamus, caudate, inferior parietal and prefrontal cortex that are associated with emotional experience, sustained attention and appraisal. In contrast, the nucleus accumbens and medial prefrontal cortex, regions implicated in reward processing, were selectively engaged during anticipation of positive clips (depicting sexually explicit content) and the mid-insula, which has been linked to processing aversive stimuli, was selectively engaged during anticipation of aversive clips (depicting graphic medical procedures); these three areas were also activated during anticipation of 'uncertain' clips reflecting a broad preparatory response for both aversive and positive stimuli. These results suggest that a common circuitry is recruited in anticipation of affective clips regardless of valence, with additional areas preferentially engaged depending on whether expected stimuli are negative or positive. © The Author (2014). Published by Oxford University Press. For Permissions, please email: journals.permissions@oup.com.

Chernobyl reactor transient simulation study

International Nuclear Information System (INIS)

Gaber, F.A.; El Messiry, A.M.

1988-01-01

This paper deals with the Chernobyl nuclear power station transient simulation study. The Chernobyl (RBMK) reactor is a graphite moderated pressure tube type reactor. It is cooled by circulating light water that boils in the upper parts of vertical pressure tubes to produce steam. At equilibrium fuel irradiation, the RBMK reactor has a positive void reactivity coefficient. However, the fuel temperature coefficient is negative and the net effect of a power change depends upon the power level. Under normal operating conditions the net effect (power coefficient) is negative at full power and becomes positive under certain transient conditions. A series of dynamic performance transient analysis for RBMK reactor, pressurized water reactor (PWR) and fast breeder reactor (FBR) have been performed using digital simulator codes, the purpose of this transient study is to show that an accident of Chernobyl's severity does not occur in PWR or FBR nuclear power reactors. This appears from the study of the inherent, stability of RBMK, PWR and FBR under certain transient conditions. This inherent stability is related to the effect of the feed back reactivity. The power distribution stability in the graphite RBMK reactor is difficult to maintain throughout its entire life, so the reactor has an inherent instability. PWR has larger negative temperature coefficient of reactivity, therefore, the PWR by itself has a large amount of natural stability, so PWR is inherently safe. FBR has positive sodium expansion coefficient, therefore it has insufficient stability it has been concluded that PWR has safe operation than FBR and RBMK reactors

Differential autoshaping to common and distinctive elements of positive and negative discriminative stimuli.

Science.gov (United States)

Wasserman, E A; Anderson, P A

1974-11-01

The learning by hungry pigeons of a discrimination between two successively presented compound visual stimuli was investigated using a two-key autoshaping procedure. Common and distinctive stimulus elements were simultaneously presented on separate keys and either followed by food delivery, S+, or not, S-. The subjects acquired both between-trial and within-trial discriminations. On S+ trials, pigeons pecked the distinctive stimulus more than the common stimulus; before responding ceased on S- trials, they pecked the common stimulus more than the distinctive one. Mastery of the within-display discrimination during S+ trials preceded mastery of the between-trials discrimination. These findings extend the Jenkins-Sainsbury analysis of discriminations based upon a single distinguishing feature to discriminations in which common and distinctive elements are associated with both the positive and negative discriminative stimuli. The similarity of these findings to other effects found in autoshaping-approach to signals that forecast reinforcement and withdrawal from signals that forecast nonreinforcement-is also discussed.

Sensitivity analysis on the effect of software-induced common cause failure probability in the computer-based reactor trip system unavailability

International Nuclear Information System (INIS)

Kamyab, Shahabeddin; Nematollahi, Mohammadreza; Shafiee, Golnoush

2013-01-01

Highlights: ► Importance and sensitivity analysis has been performed for a digitized reactor trip system. ► The results show acceptable trip unavailability, for software failure probabilities below 1E −4 . ► However, the value of Fussell–Vesley indicates that software common cause failure is still risk significant. ► Diversity and effective test is founded beneficial to reduce software contribution. - Abstract: The reactor trip system has been digitized in advanced nuclear power plants, since the programmable nature of computer based systems has a number of advantages over non-programmable systems. However, software is still vulnerable to common cause failure (CCF). Residual software faults represent a CCF concern, which threat the implemented achievements. This study attempts to assess the effectiveness of so-called defensive strategies against software CCF with respect to reliability. Sensitivity analysis has been performed by re-quantifying the models upon changing the software failure probability. Importance measures then have been estimated in order to reveal the specific contribution of software CCF in the trip failure probability. The results reveal the importance and effectiveness of signal and software diversity as applicable strategies to ameliorate inefficiencies due to software CCF in the reactor trip system (RTS). No significant change has been observed in the rate of RTS failure probability for the basic software CCF greater than 1 × 10 −4 . However, the related Fussell–Vesley has been greater than 0.005, for the lower values. The study concludes that consideration of risk associated with the software based systems is a multi-variant function which requires compromising among them in more precise and comprehensive studies

An engineering approach to common mode failure analysis

International Nuclear Information System (INIS)

Gangloff, W.C.; Franke, T.H.

1975-01-01

Safety systems for nuclear reactors can be designed using standard reliability engineering techniques such that system failure due to random component faults is extremely unlikely. However, the common-mode failure where several components fail together from a common cause is not susceptible to prevention by the usual tactics. In systems where a high degree of redundancy has been employed, the actual reliability of the system in service may be limited by common-mode failures. A methodical and thorough procedure for evaluation of system vulnerability to common-mode failures is presented. This procedure was developed for use in nuclear reactor safety systems and has been applied specifically to reactor protection. The method offers a qualitative assessment of a system whereby weak points can be identified and the resistance to common-mode failure can be judged. It takes into account all factors influencing system performance including design, manufacturing, installation, operation, testing, and maintenance. It is not a guarantee or sure solution, but rather a practical tool which can provide good assurance that the probability of common-mode protection failure has been made acceptably low. (author)

Nuclear reactor monitoring device

International Nuclear Information System (INIS)

Mihashi, Ishi; Honma, Hitoshi.

1993-01-01

The monitoring device of the present invention comprises a reactor core/reactor system data measuring and controlling device, a radioactivity concentration calculation device for activated coolants for calculating a radioactivity concentration of activated coolants in a main steam and reactor water by using an appropriate physical model, a radioactivity concentration correlation and comparison device for activated coolants for comparing correlationship with a radiation dose and an abnormality alarm device. Since radioactivity of activated primary coolants is monitored at each of positions in the reactor system and occurrence of leakage and the amount thereof from a primary circuit to a secondary circuit is monitored if the reactor has secondary circuit, integrity of the reactor system can be ensured and an abnormality can be detected rapidly. Further, radioactivity concentration of activated primary circuit coolants, represented by 16 N or 15 C, is always monitored at each of positions of PWR primary circuits. When a heat transfer pipe is ruptured in a steam generator, leakage of primary circuit coolants is detected rapidly, as well as the amount of the leakage can be informed. (N.H.)

Safeguards and Security by Design (SSBD) for Small Modular Reactors (SMRs) through a Common Global Approach

Energy Technology Data Exchange (ETDEWEB)

Badwan, Faris M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Demuth, Scott Francis [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Miller, Michael Conrad [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Pshakin, Gennady [Obninsk Institute of Physics and Power Engineering (Russian Federation)

2015-02-23

Small Modular Reactors (SMR) with power levels significantly less than the currently standard 1000 to 1600-MWe reactors have been proposed as a potential game changer for future nuclear power. SMRs may offer a simpler, more standardized, and safer modular design by using factory built and easily transportable components. Additionally, SMRs may be more easily built and operated in isolated locations, and may require smaller initial capital investment and shorter construction times. Because many SMRs designs are still conceptual and consequently not yet fixed, designers have a unique opportunity to incorporate updated design basis threats, emergency preparedness requirements, and then fully integrate safety, physical security, and safeguards/material control and accounting (MC&A) designs. Integrating safety, physical security, and safeguards is often referred to as integrating the 3Ss, and early consideration of safeguards and security in the design is often referred to as safeguards and security by design (SSBD). This paper describes U.S./Russian collaborative efforts toward developing an internationally accepted common approach for implementing SSBD/3Ss for SMRs based upon domestic requirements, and international guidance and requirements. These collaborative efforts originated with the Nuclear Energy and Nuclear Security working group established under the U.S.-Russia Bilateral Presidential Commission during the 2009 Presidential Summit. Initial efforts have focused on review of U.S. and Russian domestic requirements for Security and MC&A, IAEA guidance for security and MC&A, and IAEA requirements for international safeguards. Additionally, example SMR design features that can enhance proliferation resistance and physical security have been collected from past work and reported here. The development of a U.S./Russian common approach for SSBD/3Ss should aid the designer of SMRs located anywhere in the world. More specifically, the application of this approach may

Safeguards and Security by Design (SSBD) for Small Modular Reactors (SMRs) through a Common Global Approach

International Nuclear Information System (INIS)

Badwan, Faris M.; Demuth, Scott Francis; Miller, Michael Conrad; Pshakin, Gennady

2015-01-01

Small Modular Reactors (SMR) with power levels significantly less than the currently standard 1000 to 1600-MWe reactors have been proposed as a potential game changer for future nuclear power. SMRs may offer a simpler, more standardized, and safer modular design by using factory built and easily transportable components. Additionally, SMRs may be more easily built and operated in isolated locations, and may require smaller initial capital investment and shorter construction times. Because many SMRs designs are still conceptual and consequently not yet fixed, designers have a unique opportunity to incorporate updated design basis threats, emergency preparedness requirements, and then fully integrate safety, physical security, and safeguards/material control and accounting (MC&A) designs. Integrating safety, physical security, and safeguards is often referred to as integrating the 3Ss, and early consideration of safeguards and security in the design is often referred to as safeguards and security by design (SSBD). This paper describes U.S./Russian collaborative efforts toward developing an internationally accepted common approach for implementing SSBD/3Ss for SMRs based upon domestic requirements, and international guidance and requirements. These collaborative efforts originated with the Nuclear Energy and Nuclear Security working group established under the U.S.-Russia Bilateral Presidential Commission during the 2009 Presidential Summit. Initial efforts have focused on review of U.S. and Russian domestic requirements for Security and MC&A, IAEA guidance for security and MC&A, and IAEA requirements for international safeguards. Additionally, example SMR design features that can enhance proliferation resistance and physical security have been collected from past work and reported here. The development of a U.S./Russian common approach for SSBD/3Ss should aid the designer of SMRs located anywhere in the world. More specifically, the application of this approach may

Research reactor standards and their impact on the TRIGA reactor community

International Nuclear Information System (INIS)

Richards, W.J.

1980-01-01

The American Nuclear Society has established a standards committee devoted to writing standards for research reactors. This committee was formed in 1971 and has since that time written over 15 standards that cover all aspects of research reactor operation. The committee has representation from virtually every group concerned with research reactors and their operation. This organization includes University reactors, National laboratory reactors, Nuclear Regulatory commission, Department of Energy and private nuclear companies and insurers. Since its beginning the committee has developed standards in the following areas: Standard for the development of technical specifications for research reactors; Quality control for plate-type uranium-aluminium fuel elements; Records and reports for research reactors; Selection and training of personnel for research reactors; Review of experiments for research reactors; Research reactor site evaluation; Quality assurance program requirements for research reactors; Decommissioning of research reactors; Radiological control at research reactor facilities; Design objectives for and monitoring of systems controlling research reactor effluents; Physical security for research reactor facilities; Criteria for the reactor safety systems of research reactors; Emergency planning for research reactors; Fire protection program requirements for research reactors; Standard for administrative controls for research reactors. Besides writing the above standards, the committee is very active in using communications with the nuclear regulatory commission on proposed rules or positions which will affect the research reactor community

Device for positioning ultrasonic probes and/or television cameras on the outer surface of reactor pressure vessels

International Nuclear Information System (INIS)

Zipser, R.; Dose, G.F.

1977-01-01

The device makes possible periodical in-service inspections of welding seams and material of a reactor pressure vessel without local human presence. A 'support ring' encloses the pressure vessel in a horizontal plane with free space. It is vertically moved up and down in the space between pressure vessel and thermal shield by means of tackles. At a control desk placed in a protected area its movement is controlled and its vertical position is indicated. A 'rotating track' with its own drive is rotating remote-controlled on the 'support ring'. By a combination of the vertical with the rotating movement, an ultrasonic probe placed removably on the 'rotating hack', or a television camera will be brought to any position on the cylindrical circumference of the pressure vessel. Special devices extend the radius of action, in upward direction for inspecting the welding seams of the coolant nozzles, and in downward direction for the inspection of welds on the hemispherical bottom of the pressure vessel or on the outlet pipe nozzle placed there. The device remains installed during reactor operation, but is moved down to the lower horizontal surface of the thermal shield. Parts which are sensible to radiation like probes or television cameras and special devices will then be removed respectively mounted before beginning an inspection compaign. This position may be reached by the lower access in the biological shield and through an opening in the horizontal surface of the thermal shield. (HP) [de

Reactor power peaking information display

International Nuclear Information System (INIS)

Book, T.L.; Kochendarfer, R.A.

1986-01-01

This patent describes a system for monitoring operating conditions within a nuclear reactor. The system consists of a method for measuring the operating parameters within the nuclear reactor, including the position of axial power shaping rods and regulating control rod. It also includes a method for determining from the operating parameters the operating limits before a power peaking condition exists within the nuclear reactor, and a method for displaying the operating limits which consists of a visual display permitting the continuous monitoring of the operating conditions within the nuclear reactor as a graph of the shaping rod position vs the regulating rod position having a permissible area and a restricted area. The permissible area is further divided into a recommended operating area for steady state operation and a cursor located on the graph to indicate the present operating condition of the nuclear reactor to allow an operator to view any need for corrective action based on the movement of the cursor out of the recommended operating area and to take any corrective transient action within the permissible area

Licensed reactor nuclear safety criteria applicable to DOE reactors

International Nuclear Information System (INIS)

1993-11-01

This document is a compilation and source list of nuclear safety criteria that the Nuclear Regulatory Commission (NRC) applies to licensed reactors; it can be used by DOE and DOE contractors to identify NRC criteria to be evaluated for application to the DOE reactors under their cognizance. The criteria listed are those that are applied to the areas of nuclear safety addressed in the safety analysis report of a licensed reactor. They are derived from federal regulations, USNRC regulatory guides, Standard Review Plan (SRP) branch technical positions and appendices, and industry codes and standards

Nuclear reactor theory

International Nuclear Information System (INIS)

Sekimoto, Hiroshi

2007-09-01

This textbook is composed of two parts. Part 1 'Elements of Nuclear Reactor Theory' is composed of only elements but the main resource for the lecture of nuclear reactor theory, and should be studied as common knowledge. Much space is therefore devoted to the history of nuclear energy production and to nuclear physics, and the material focuses on the principles of energy production in nuclear reactors. However, considering the heavy workload of students, these subjects are presented concisely, allowing students to read quickly through this textbook. (J.P.N.)

Generation III+ Reactor Portfolio

International Nuclear Information System (INIS)

2010-03-01

While the power generation needs of utilities are unique and diverse, they are all faced with the double challenge of meeting growing electricity needs while curbing CO 2 emissions. To answer these diverse needs and help tackle this challenge, AREVA has developed several reactor models which are briefly described in this document: The EPR TM Reactor: designed on the basis of the Konvoi (Germany) and N4 (France) reactors, the EPRTM reactor is an evolutionary model designed to achieve best-in-class safety and operational performance levels. The ATMEA1 TM reactor: jointly designed by Mitsubishi Heavy Industries and AREVA through ATMEA, their common company. This reactor design benefits from the competencies and expertise of the two mother companies, which have commissioned close to 130 reactor units. The KERENA TM reactor: Designed on the basis of the most recent German BWR reactors (Gundremmingen) the KERENA TM reactor relies on proven technology while also including innovative, yet thoroughly tested, features. The optimal combination of active and passive safety systems for a boiling water reactor achieves a very low probability of severe accident

BWR type reactor

International Nuclear Information System (INIS)

Watanabe, Shoichi

1983-01-01

Purpose : To flatten the radial power distribution in the reactor core thereby improve the thermal performance of the reactor core by making the moderator-fuel ratio of fuel assemblies different depending on their position in the reactor core. Constitution : The volume of fuels disposed in the peripheral area of the reactor core is decreased by the increase of the volume of moderators in fuel assemblies disposed in the peripheral area of the reactor core to thereby make the moderator-fuel volume greater in the peripheral area than that in the central area. The moderator-fuel ratio adjustment is attained by making the number of water rods greater, decreasing the diameter of fuel pellets or decreasing the number of fuel pins in fuel assemblies disposed at the peripheral area of the reactor core as compared with fuel assemblies disposed at the central area of the reactor core. In this way, the infinite multiplication factors of fuels can be increased to thereby improve the reactor core performance. (Aizawa, K.)

Fast reactors worldwide

International Nuclear Information System (INIS)

Hall, R.S.; Vignon, D.

1985-01-01

The paper concerns the evolution of fast reactors over the past 30 years, and their present status. Fast reactor development in different countries is described, and the present position, with emphasis on cost reduction and collaboration, is examined. The French development of the fast breeder type reactor is reviewed, and includes: the acquisition of technical skills, the search for competitive costs and the spx2 project, and more advanced designs. Future prospects are also discussed. (U.K.)

Licensed reactor nuclear safety criteria applicable to DOE reactors

International Nuclear Information System (INIS)

1991-04-01

The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC [Nuclear Regulatory Commission] licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor

TSO Study Project on Development of a Common Safety Approach in the EU for Large Evolutionary Pressurised Water Reactors

International Nuclear Information System (INIS)

2001-10-01

In pursuance of the objectives of the Council Resolutions of 1975 and 1992 on the technological issues of nuclear safety, the European Commission (EC) is seeking to promote a sustained joint in-depth study on possible significant future nuclear power reactor safety cases. To that end the EC decided to support financially a study by the grouping of the European Union Technical Safety Organisations (TSOG). The general objective of the study programme was to promote, through a collaboration of European Union Technical Safety Organisations (TSOs), common views on technical safety issues related to large evolutionary PWRs in Europe, which could be ready for operation during the next decade. AVN (Belgium) (Technical project leader), AEA Technology (United Kingdom), ANPA (Italy) CIEMAT (Spain), GRS (Germany), IPSN (France), were the TSOs participating in the study which was co-ordinated by RISKAUDIT. The study focused notably on the EPR project initiated by the French and German utilities and vendors. It also considered relevant projects, even of plants of different size, developed outside the European Union in order to provide elements important for the safety characterisation and which could contribute to the credibility and confidence of EPR. It is expected that this study will constitute a significant step towards the development of a common safety approach in EU countries. The study constitutes an important step forward in the development of a common approach of the TSOs to the safety of advanced evolutionary pressurised water reactors. This goal was mainly achieved by an in-depth analysis of the key safety issues, taking into account new developments in the national technical safety objectives and in the EPR design. For this reason the Commission has decided to publish at least the present summary report containing the main outcomes of the TSO study. Confidentiality considerations unfortunately prevent the open publication of the full series of reports. (author)

Reactor pressure tank

International Nuclear Information System (INIS)

Dorner, H.; Scholz, M.; Jungmann, A.

1975-01-01

In a reactor pressure tank for a nuclear reactor, self-locking hooks engage a steel ring disposed over the removable cover of the steel vessel. The hooks exert force upon the cover to maintain the cover in a closed position during operation of the reactor pressure tank. The force upon the removal cover is partly the result of the increasing temperature and thermal expansion of the steel vessel during operation. The steel vessel is surrounded by a reinforced-concrete tank. (U.S.)

Report of researches by common utilization of facilities in Kyoto University Research Reactor Institute, first half of fiscal year 1981

International Nuclear Information System (INIS)

1983-01-01

The technical report of the Kyoto University Research Reactor Institute is published any time to immediately report on the results of the functional tests of various experimental facilities, the test results for the products made for trial, radiation control, the situation of waste treatment, the data required for research and experiment such as the reports of study meetings, the conspicuous results obtained amid researches, new processes, and the discussion on other papers and reports. In this report, the title, the names of reporters and the summary of 57 researches carried out by the common utilization of the facilities in the Kyoto University Research Reactor Institute are collected. The themes of the researches are such as neutron radiography using a research reactor, measurement of Zr/Hf ratio in zirconium, interstitial germanium atoms in thermal neutron irradiation study, measurement of induced radioactivity due to neutrons in Nagasaki and Hiroshima atomic bombings, properties of semiconductor electrons in radiation study, induction of mutation in crops by neutron irradiation and utilization for breeding, thermal fluorescence mechanism of alkali halide and MgO single crystals, atomic configuration in PZT rhombohedron phase, modulated structure of Cu-Co alloys, excitation of nuclei by positron annihilation and others. (Kako, I.)

Report of researches by common utilization of facilities in Kyoto University Research Reactor Institute, latter half of fiscal year 1981

International Nuclear Information System (INIS)

1983-01-01

The technical report of the Kyoto University Research Reactor Institute is published any time to immediately report on the results of the functional tests of various experimental facilities, the test results for the products made for trial, radiation control, the situation of waste treatment, the data required for research and experiment such as the reports of study meetings, the conspicuous results obtained amid researches, new processes, and the discussion on other papers and reports. In this report, the title, the names of reporters and the summary of 61 researches carried out by the common utilization of the facilities in the Kyoto University Research Reactor Institute are collected. The themes of the researches are such as radioactivation analysis of trace elements in rocks and minerals, anodic oxidation films of GaAs and structure, measurement of yield of uranium isotopes produced by reactor neutron irradiation of thorium, geochemical study of trace elements in hydrosphere by radio-activation analysis, various diseases and variation of elements in rat furs, Moessbauer spectroscopic study of gold compounds with singular coupling by Au-197, measurement of grass-eating quantity and rate of digestion of cows using Au and Eu, sickness biochemical study of trace elements in hair samples of patients and others. (Kako, I.)

Inherently safe reactors

International Nuclear Information System (INIS)

Maartensson, Anders

1992-01-01

A rethinking of nuclear reactor safety has created proposals for new designs based on inherent and passive safety principles. Diverging interpretations of these concepts can be found. This article reviews the key features of proposed advanced power reactors. An evaluation is made of the degree of inherent safety for four different designs: the AP-600, the PIUS, the MHTGR and the PRISM. The inherent hazards of today's most common reactor principles are used as reference for the evaluation. It is concluded that claims for the new designs being inherently, naturally or passively safe are not substantiated by experience. (author)

Lateral restraint assembly in a nuclear reactor

International Nuclear Information System (INIS)

Brown, S.J.; Gorholt, W.

1977-01-01

A lateral restraint assembly is described for a reactor of, for example, the high temperature gas-cooled type which commonly includes a reactor core of relatively complex construction supported within a shell or vessel providing a shielded cavity for containing the reactor core. (U.K.)

Reactor core for FBR type reactor

International Nuclear Information System (INIS)

Fujita, Tomoko; Watanabe, Hisao; Kasai, Shigeo; Yokoyama, Tsugio; Matsumoto, Hiroshi.

1996-01-01

In a gas-sealed assembly for a FBR type reactor, two or more kinds of assemblies having different eigen frequency and a structure for suppressing oscillation of liquid surface are disposed in a reactor core. Coolant introduction channels for introducing coolants from inside and outside are disposed in the inside of structural members of an upper shielding member to form a shielding member-cooling structure in the reactor core. A structure for promoting heat conduction between a sealed gas in the assembly and coolants at the inner side or the outside of the assembly is disposed in the reactor core. A material which generates heat by neutron irradiation is disposed in the assembly to heat the sealed gases positively by radiation heat from the heat generation member also upon occurrence of power elevation-type event to cause temperature expansion. Namely, the coolants flown out from or into the gas sealed-assemblies cause differential fluctuation on the liquid surface, and the change of the capacity of a gas region is also different on every gas-sealed assemblies thereby enabling to suppress fluctuation of the reactor power. Pressure loss is increased by a baffle plate or the like to lower the liquid surface of the sodium coolants or decrease the elevating speed thereof thereby suppressing fluctuation of the reactor power. (N.H.)

Design Modification of Double Acting Positioning Mechanism Inside a Research Reactor Pool

International Nuclear Information System (INIS)

El-Bahrawy, A.H.A.

2002-01-01

One of the major problems that can face the designers of long actuators is the blocking due to high friction and severe wear of sliding parts, when the case become metal to metal in contact. The present work is an optimum solution for a typical blocking problem that had been raised during the commissioning test of the fission counter three positioning mechanisms inside MPR research reactor at Inshas. A detailed technical investigation had been made in order to define precisely the main reasons behind the problem. High misalignment values of piston and piston rod assembly inside the hydraulic cylinders due to poor design has been discovered to be the main source of the problem. The author has introduced a new design idea, and other mechanisms have been manufactured based on the new design to replace the old ones. The system was tested and found to function with good performance

Apparatus for removing and/or positioning fuel assemblies of a nuclear reactor

International Nuclear Information System (INIS)

Vuckovich, M.; Burkett, J.P.; Sallustio, J.

1983-01-01

Apparatus for positioning fuel assemblies of a nuclear reactor includes a control for a crane comprising a strain gauge connected to the crane line which raises and lowers the load. The signal from the strain gauge is compared with setpoints; which if the strain gauge signal exceeds a high-level setpoint, indicating that the movement of a fuel assembly is obstructed, the line drive is disabled. The line drive is also disabled if the strain gauge signal is less than a low-level setpoint, indicating that a fuel being deposited contacts the bottom of its slot or an obstruction. To preclude lateral movement of the fuel assembly suspended from the crane line, the traverse drive of the crane is disabled once the strain-gauge signal exceeds the low-level setpoint. The traverse drive can only be enabled after the strain-gauge signal is less than a slack-line setpoint. (author)

Apparatus for removing and/or positioning fuel assemblies of a nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Vuckovich, M; Burkett, J P; Sallustio, J

1981-11-30

Apparatus for positioning fuel assemblies of a nuclear reactor includes a control for a crane comprising a strain gauge connected to the crane line which raises and lowers the load. The signal from the strain gauge is compared with setpoints; which if the strain gauge signal exceeds a high-level setpoint, indicating that the movement of a fuel assembly is obstructed, the line drive is disabled. The line drive is also disabled if the strain gauge signal is less than a low-level setpoint, indicating that a fuel being deposited contacts the bottom of its slot or an obstruction. To preclude lateral movement of the fuel assembly suspended from the crane line, the traverse drive of the crane is disabled once the strain-gauge signal exceeds the low-level setpoint. The traverse drive can only be enabled after the strain-gauge signal is less than a slack-line setpoint.

Differential autoshaping to common and distinctive elements of positive and negative discriminative stimuli1

Science.gov (United States)

Wasserman, Edward A.; Anderson, Patricia A.

1974-01-01

The learning by hungry pigeons of a discrimination between two successively presented compound visual stimuli was investigated using a two-key autoshaping procedure. Common and distinctive stimulus elements were simultaneously presented on separate keys and either followed by food delivery, S+, or not, S−. The subjects acquired both between-trial and within-trial discriminations. On S+ trials, pigeons pecked the distinctive stimulus more than the common stimulus; before responding ceased on S− trials, they pecked the common stimulus more than the distinctive one. Mastery of the within-display discrimination during S+ trials preceded mastery of the between-trials discrimination. These findings extend the Jenkins-Sainsbury analysis of discriminations based upon a single distinguishing feature to discriminations in which common and distinctive elements are associated with both the positive and negative discriminative stimuli. The similarity of these findings to other effects found in autoshaping—approach to signals that forecast reinforcement and withdrawal from signals that forecast nonreinforcement—is also discussed. PMID:16811812

Tank type LMFBR type reactors

International Nuclear Information System (INIS)

Shimizu, Hiroshi

1985-01-01

Purpose: To detect the abnormality in the suspended body or reactor core supporting structures thereby improve the safety and reliability of tank type LMFBR reactors. Constitution: Upon inspection during reactor operation period, the top end of the gripper sensing rod of a fuel exchanger is abutted against a supporting bed and the position of the reactor core supporting structures from the roof slab is measured by a stroke measuring device. Then, the sensing rod is pulled upwardly to abut against the arm portion and the position is measured by the stroke measuring device. The measuring procedures are carried out for all of the sensing rods and the measured values are compared with a previously determined value at the initial stage of the reactor operation. As a result, it is possible to detect excess distortions and abnormal deformation in the suspended body or reactor core supporting structures. Furthermore, integrity of the suspended body against thermal stresses can be secured by always measuring the coolant liquid level by the level measuring sensor. (Kamimura, M.)

Investigation of the basic reactor physics characteristics of the Dalat Nuclear Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Khang, Ngo Phu [Nuclear Research Inst., Da Lat (Viet Nam)

1994-10-01

The Dalat nuclear research reactor was reconstructed from TRIGA MARK II reactor, built in 1963 with nominal power of 250 KW, and reached its planned nominal power of 500 kW for the first time in Feb. 1984. The Dalat reactor has some characteristics distinct from the former TRIGA reactor. Investigation of its characteristics is carried out by the determination of the reactor physics parameters. This paper represents the experimental results obtained for the effective fraction of the delayed photoneutrons, the extraneous neutron source left after the reactor is shut down, the lowest power levels of reactor critical states, the relative axial and radial distributions of thermal neutrons, the safe positive reactivity inserted into the reactor at deep subcritical state, the reactivity temperature coefficient of water, the temperature on the surface of the fuel elements, etc. (author). 10 refs., 10 figs., 2 tabs.

Reactor core

International Nuclear Information System (INIS)

Matsuura, Tetsuaki; Nomura, Teiji; Tokunaga, Kensuke; Okuda, Shin-ichi

1990-01-01

Fuel assemblies in the portions where the gradient of fast neutron fluxes between two opposing faces of a channel box is great are kept loaded at the outermost peripheral position of the reactor core also in the second operation cycle in the order to prevent interference between a control rod and the channel box due to bending deformation of the channel box. Further, the fuel assemblies in the second row from the outer most periphery in the first operation cycle are also kept loaded at the second row in the second operation cycle. Since the gradient of the fast neutrons in the reactor core is especially great at the outer circumference of the reactor core, the channel box at the outer circumference is bent such that the surface facing to the center of the reactor core is convexed and the channel box in the second row is also bent to the identical direction, the insertion of the control rod is not interfered. Further, if the positions for the fuels at the outermost periphery and the fuels in the second row are not altered in the second operation cycle, the gaps are not reduced to prevent the interference between the control rod and the channel box. (N.H.)

Fast reactor cover gas purification - The UK position

Energy Technology Data Exchange (ETDEWEB)

Thorley, A W

1987-07-01

The cover gas in the Prototype Fast Reactor (PFR) provides an inert gas blanket for both primary and secondary sodium circuits, ensures inert gas padding exists between the upper seals associated with penetrations through the reactor roof and provides argon to items of plant such as the control rods and the rotating shield and also to on line instruments such as the secondary circuit Katharometers. In order to meet these and other requirements purification of the argon cover gas is important to ensure: gas fed to purge gaps in the area of the magnetic hold device in the control rod mechanisms is not laden with sodium aerosols and reactive impurities (O{sub 2}, H{sub 2}) which could cause blocking both within the gaps and pipelines; gas phase detection systems which provide early warning of steam generator failures or oil ingress into the sodium are not affected by the presence of gaseous impurities such as H{sub 2}, CO/CO{sub 2} and CH{sub 4}; mass transfer processes involving both corrosion products and interstitial atoms cannot be sustained in the cover gas environment due to the presence of high levels of O{sub 2}, N{sub 2} and carburising gases; background levels of radioactivity (eg Xe 133) are sufficiently low to enable gas phase detection of failed fuel pins, and the primary circuit gas blanket activity is sufficiently reduced so that discharges to the atmosphere are minimised. This paper describes how the PFR cover gas purification system is coping with these various items and how current thinking regarding the design of cover gas purification systems for a Civil Demonstration Fast Reactor (CDFR), where larger gas volumes and higher levels of radioactivity may be involved, is being guided by current experience on PFR. The paper also briefly review the experimental work planned to study aerosol and caesium behaviour in cove gas environments and discusses the behaviour of those impurities such as Zn, oil and N{sub 2} which are potentially damaging if certain

Fast reactor cover gas purification - The UK position

International Nuclear Information System (INIS)

Thorley, A.W.

1987-01-01

The cover gas in the Prototype Fast Reactor (PFR) provides an inert gas blanket for both primary and secondary sodium circuits, ensures inert gas padding exists between the upper seals associated with penetrations through the reactor roof and provides argon to items of plant such as the control rods and the rotating shield and also to on line instruments such as the secondary circuit Katharometers. In order to meet these and other requirements purification of the argon cover gas is important to ensure: gas fed to purge gaps in the area of the magnetic hold device in the control rod mechanisms is not laden with sodium aerosols and reactive impurities (O 2 , H 2 ) which could cause blocking both within the gaps and pipelines; gas phase detection systems which provide early warning of steam generator failures or oil ingress into the sodium are not affected by the presence of gaseous impurities such as H 2 , CO/CO 2 and CH 4 ; mass transfer processes involving both corrosion products and interstitial atoms cannot be sustained in the cover gas environment due to the presence of high levels of O 2 , N 2 and carburising gases; background levels of radioactivity (eg Xe 133) are sufficiently low to enable gas phase detection of failed fuel pins, and the primary circuit gas blanket activity is sufficiently reduced so that discharges to the atmosphere are minimised. This paper describes how the PFR cover gas purification system is coping with these various items and how current thinking regarding the design of cover gas purification systems for a Civil Demonstration Fast Reactor (CDFR), where larger gas volumes and higher levels of radioactivity may be involved, is being guided by current experience on PFR. The paper also briefly review the experimental work planned to study aerosol and caesium behaviour in cove gas environments and discusses the behaviour of those impurities such as Zn, oil and N 2 which are potentially damaging if certain levels are exceeded in operating

MDEP Design-Specific Common Position CP-EPRWG-02. Common position addressing Fukushima Daiichi-related issues

International Nuclear Information System (INIS)

2015-10-01

A severe accident involving several units took place in Japan at Fukushima Daiichi nuclear power plant (NPP) in March 2011. The immediate cause of the accident was an earthquake followed by a tsunami coupled with inadequate provisions against the consequences of such events in the design. Opportunities to improve protection against a realistic design basis tsunami were not taken. As a consequence of the tsunami, safety equipment and the related safety functions were lost at the plant, leading to core damage in three units and subsequently to large radioactive releases (INES 7). Several studies have already been performed to better understand the accident progression and detailed technical studies are still in progress in Japan and elsewhere. In the meantime, on-going studies on the behaviour of NPPs in very severe situations, similar to Fukushima Daiichi, seek to identify potential vulnerabilities in plant design and operation; to suggest reasonably practicable upgrades; or to recommend enhanced regulatory requirements and guidance to address such situations. Likewise, agencies around the world that are responsible for regulating the design, construction and operation of EPR plants are engaged in similar activities. The MDEP EPR Working Group (EPRWG) members consist of members from the United States, the United Kingdom, France, Finland, China, India and Sweden. Because not all of these countries have completed the regulatory review of their EPR applications yet, this paper identifies common preliminary approaches to address potential safety improvements for EPR plants, as well as common general expectations for new nuclear power plants, as related to lessons learned from the Fukushima Daiichi accident or Fukushima Daiichi-related issues. After the safety reviews of the EPR design applications that are currently in review are completed, the regulators will update this paper to reflect their safety conclusions regarding the EPR design and how the design could be

Common mode and coupled failure

International Nuclear Information System (INIS)

Taylor, J.R.

1975-10-01

Based on examples and data from Abnormal Occurence Reports for nuclear reactors, a classification of common mode or coupled failures is given, and some simple statistical models are investigated. (author)

United Kingdom and USSR reactor types

International Nuclear Information System (INIS)

Lewins, Jeffery

1988-01-01

The features of the RBMK reactor operated at Chernobyl are compared with reactor types pertinent to the UK. The UK reactors covered are in three classes: the commercial reactors now built and operated or in commission (Magnox and Advanced Gas-cooled Reactor (AGR)); the prototype Steam Generating Heavy Water Reactor (SGHWR) and Prototype Fast Reactor (PFR) that have comparable performance to commercial reactors; and the proposed Pressurised Water Reactor (PWR) or Sizewell 'B' design which, it will be recollected, is different in detail from PWRs built elsewhere. We do not include research and test reactors nor the Royal Navy PWRs. The appendices explain resonances, Doppler and Xenon effects, the reactor physics of Chernobyl and positive void coefficients all of which are relevant to the comparisons. (author)

Elmo Bumpy Torus Reactor

International Nuclear Information System (INIS)

McAlees, D.G.; Uckan, N.A.; Lidsky, L.M.

1976-01-01

In the Elmo Bumpy Torus Reactor (EBTR) study the feasibility of achieving a fusion power plant based on the EBT confinement concept was evaluated. If the present understanding of the physics can be extrapolated to reactor scale devices the reactor could operate at high beta, high power density, and at steady state. The high aspect ratio of the device eases the accessibility, structural design and remote maintenance problems which are common to low aspect ratio machines. A version of the EBTR reference design described here could be constructed with only minor extrapolations in available technology

Method for temporary shielding of reactor vessel internals

International Nuclear Information System (INIS)

Grimm, N.P.; Sejvar, J.

1991-01-01

This patent describes a method for shielding stored internals for reactor vessel annealing. It comprises removing nuclear fuel from the reactor vessel containment building; removing and storing upper and lower core internals under water in a refueling canal storage area; assembling a support structure in the refueling canal between the reactor vessel and the stored internals; introducing vertical shielding tanks individually through a hatch in the containment building and positioning each into the support structure; introducing horizontal shielding tanks individually through a hatch in the containment building and positioning each above the stored internals and vertical tanks; draining water from the refueling canal to the level of a flange of the reactor vessel; placing an annealing apparatus in the reactor vessel; pumping the remaining water from the reactor vessel; and annealing the reactor vessel

Study of power reactor dynamics by stochastic reactor oscillator method; Proucavanje dinamike reaktora snage metodom stohastickog reaktorskog oscilatora

Energy Technology Data Exchange (ETDEWEB)

Velickovic, Lj; Petrovic, M [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

1968-12-15

Stochastic reactor oscillator and cross correlation method were used for determining reactor dynamics characteristics. Experimental equipment, fast reactor oscillator (BOR-1) was activated by random pulses from the GBS-16 generator. Tape recorder AMPEX-SF-300 and data acquisition tool registered reactor response to perturbations having different frequencies. Reactor response and activation signals were cross correlated by digital computer for different positions of stochastic oscillator and ionization chamber.

A television/still camera with common optical system for reactor inspection

International Nuclear Information System (INIS)

Hughes, G.; McBane, P.

1976-01-01

One of the problems of reactor inspection is to obtain permanent high quality records. Video recordings provide a record of poor quality but known content. Still cameras can be used but the frame content is not predictable. Efforts have been made to combine T.V. viewing to align a still camera but a simple combination does not provide the same frame size. The necessity to preset the still camera controls severely restricts the flexibility of operation. A camera has, therefore, been designed which allows a search operation using the T.V. system. When an anomaly is found the still camera controls can be remotely set, an exact record obtained and the search operation continued without removal from the reactor. An application of this camera in the environment of the blanket gas region above the sodium region in PFR at 150 0 C is described

Fission track dating method: I. Study of neutron flux uniformity in some irradiation positions of IEA-R1 reactor

International Nuclear Information System (INIS)

Osorio, A.M.; Hadler, J.C.; Iunes, P.J.; Paulo, S.R. de

1993-06-01

In order to use the fission track dating method the flux gradient was verified within the sample holder, in some irradiation positions of the IEA-R1 reactor at IPEN/CNEN, Sao Paulo. The fission track dating method considers only the thermal neutron fission tracks, to subtract the other contributions sample irradiations with a cadmium cover was performed. The neutron flux cadmium influence was studied. (author)

In-Pile Loop Safety in Integrated with the Multipurpose Reactor in the case of in-Pile Loop Leakage at the Core Position

International Nuclear Information System (INIS)

Suharno; Sugianto; Giarno; Aliq; Widodo, Surip; Aji, Bintoro; Purba, Julwan Hendry; Karyanta, Edy

1999-01-01

In-Pile Loop Safety analysis in integrated with the multipurpose reactor in the case of In-Pile Loop leakage at the core position has been conducted which intended to evaluate the failure of fuel element. By considering design of In-Pile Loop and the highest possibility position of of leakage, the failure of fuel element is emphasized on mechanical aspect. The thermal hydraulic aspect is not taken into account due to the condition that when the leakage occurred the reactor has been in shut down condition. It is determined that the spray attacks the top position of fuel element, and to be calculated the force, of spray that produces 1 cm deflection on the single fuel element. Using that four (4) fuel elements is calculated because in the real condition 4 fuel elements will undergo deflection of 43.8 kg is obtained that producing 1 cm deflection and the force of 1228 kg that causes failure on the bottom of fuel element as shear force is also obtained. Whatever the force, high or low, the damage of fuel element occurred at the bottom part or at the position of grid plate. Therefore there is no damage on the fuel part (uranium meat) and the releasing of radioactive material from fuel plate is not happened

Device for transferring fast nuclear reactor fuel assemblies

International Nuclear Information System (INIS)

Plagnard, Andre.

1982-01-01

The description is given of a device for transferring fuel assemblies between a storage position near the reactor vessel and a position where the irradiated assemblies are evacuated and the provision of new assemblies for the reactor. This device can be dismantled and is movable as a whole for its successive use on several reactors and includes: - a platform mounted so as to rotate on a support made to rest on the structure of the reactor, the platform having at least one opening then being horizontal and mobile about a vertical axis to bring the opening successively in position with vertical wells giving access to the storage and evacuation positions of the assemblies provided in the reactor structure, - at least one hopper that can contain one assembly in a vertical position, located on the upper surface of the platform around the opening provided in it and fitted with a winch for the vertical moving of the assemblies inside the wells and the hopper, when these follow each other by rotation of the platform, - at least one connecting device carried on the platform for connecting the hopper and wells when these are in line [fr

Bioconversion reactor

Science.gov (United States)

McCarty, Perry L.; Bachmann, Andre

1992-01-01

A bioconversion reactor for the anaerobic fermentation of organic material. The bioconversion reactor comprises a shell enclosing a predetermined volume, an inlet port through which a liquid stream containing organic materials enters the shell, and an outlet port through which the stream exits the shell. A series of vertical and spaced-apart baffles are positioned within the shell to force the stream to flow under and over them as it passes from the inlet to the outlet port. The baffles present a barrier to the microorganisms within the shell causing them to rise and fall within the reactor but to move horizontally at a very slow rate. Treatment detention times of one day or less are possible.

FBR Italian position. The PEC reactor and its control rods system

International Nuclear Information System (INIS)

Artioli, C.

1984-01-01

The greatest effort in Italy concerning Fast Breeder Reactors is concentrated on the PEC project. This project represents in fact the sole Italian work on the FBR option. Up to now the state of progress of the PEC project is as follows: - Design work on reactor components and related systems: 80% - Civil engineering work: 45% - Components construction: 18%. It has recently been stated that construction of the plant must be finished by 1987. The progressive design of the PEC control system and the supporting research and development programmes are described in this paper

Activities of research-reactor-technology project in FNCA from FY2005 to FY2007. Sharing neutronics calculation technique for core management and utilization of research reactors

International Nuclear Information System (INIS)

2010-07-01

RRT project (Research-Reactor-Technology Project) was carried out with the theme of 'sharing neutronics calculation technique for core management and utilization of research reactors' in the framework of FNCA (Forum for Nuclear Cooperation in Asia) from FY2005 to FY2007. The objective of the project was to improve and equalize the level of neutronics calculation technique for the reactor core management among participating countries to assure the safe and stable operation of research reactors and the promotion of the effective utilization. Neutronics calculation codes, namely SRAC code system and MVP code, were adopted as common codes. Participating countries succeeded in applying the common codes to analyzing the core of each domestic research reactor. Some participating countries succeeded in applying the common codes to analyzing for utilization of own research reactors. Activities of RRT project have improved and equalized the level of neutronics calculation technique among participating countries. (author)

Water level monitoring device in nuclear reactor

International Nuclear Information System (INIS)

Miura, Kiyohide; Otake, Tomohiro.

1988-01-01

Purpose: To monitor the water level in a pressure vessel of BWR type nuclear reactors at high accuracy by improving the compensation functions. Constitution: In the conventional water level monitor in a nuclear reactor, if the pressure vessel is displaced by the change of the pressure in the reactor or the temperature of the reactor water, the relative level of the reference water head in a condensation vessel is changed to cause deviation between the actual water level and the indicated water level to reduce the monitoring accuracy. According to the invention, means for detecting the position of the reference water head and means for detection the position in the condensation vessel are disposed to the pressure vessel. Then, relative positional change between the condensation vessel and the reference water head is calculated based on detection sinals from both of the means. The water level is compensated and calculated by water level calculation means based on the relative positional change, water level signals from the level gage and the pressure signals from the pressure gage. As a result, if the pressure vessel is displaced due to the change of the temperature or pressure, it is possible to measure the reactor water level accurately thereby remakably improve the reliability for the water level control in the nuclear reactor. (Horiuchi, T.)

A right to choose how to live: the Australian common law position on refusals of care.

Science.gov (United States)

Curnow, Katherine

2014-12-01

There has been limited examination of the Australian common law position regarding contemporaneous refusals of care or medical treatment by competent adults since the first two Australian cases to adjudicate on refusals of this type: H Ltd v J and Brightwater Care Group (Inc) v Rossiter. This article maps the legal position in Australia in light of the two cases with particular emphasis on the finding in H Ltd v J that self-starvation is not suicide at common law. Finally, this article highlights the broader relevance of this area of the law and its capacity to inform debates as disparate as whether to legalise voluntary euthanasia and the possible implications for the autonomy of pregnant women of proposed laws giving legal status to fetuses (particularly Zoe's Law).

Fourth Generation Reactor Concepts

International Nuclear Information System (INIS)

Furtek, A.

2008-01-01

were selected to generation IV by the GIF to further studies: Gas-Cooled Fast Reactor (GFR), Lead-Cooled Fast Reactor (LFR), Molten Salt Reactor (MSR), Sodium-Cooled Fast Reactor (SFR), Supercritical Water-Cooled Reactor (SCWR), Very High Temperature Reactor (VHTR). These six systems would each need a dedicated effort in research and development. Some consideration for the fuel and recycling technology are common and can be shared. These common areas encompass: fuel cycles, fuels and materials choice, energy products, risk and safety, economics and proliferation and physical protection concerns.(author)

Reactor Neutron Sources

International Nuclear Information System (INIS)

Aksenov, V.L.

1994-01-01

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

Pulsed fusion reactors

International Nuclear Information System (INIS)

1975-01-01

This summer school specialized in examining specific fusion center systems. Papers on scientific feasibility are first presented: confinement of high-beta plasma, liners, plasma focus, compression and heating and the use of high power electron beams for thermonuclear reactors. As for technological feasibility, lectures were on the theta-pinch toroidal reactors, toroidal diffuse pinch, electrical engineering problems in pulsed magnetically confined reactors, neutral gas layer for heat removal, the conceptual design of a series of laser fusion power plants with ''Saturn'', implosion experiments and the problem of the targets, the high brightness lasers for plasma generation, and topping and bottoming cycles. Some problems common to pulsed reactors were examined: energy storage and transfer, thermomechanical and erosion effects in the first wall and blanket, the problems of tritium production, radiation damage and neutron activation in blankets, and the magnetic and inertial confinement

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.)

Materials for passively safe reactors

International Nuclear Information System (INIS)

Simnad, T.

1993-01-01

Future nuclear power capacity will be based on reactor designs that include passive safety features if recent progress in advanced nuclear power developments is realized. There is a high potential for nuclear systems that are smaller and easier to operate than the current generation of reactors, especially when passive or intrinsic characteristics are applied to provide inherent stability of the chain reaction and to minimize the burden on equipment and operating personnel. Taylor, has listed the following common generic technical features as the most important goals for the principal reactor development systems: passive stability, simplification, ruggedness, case of operation, and modularity. Economic competitiveness also depends on standardization and assurance of licensing. The performance of passively safe reactors will be greatly influenced by the successful development of advanced fuels and materials that will provide lower fuel-cycle costs. A dozen new designs of advanced power reactors have been described recently, covering a wide spectrum of reactor types, including pressurized water reactors, boiling water reactors, heavy-water reactors, modular high-temperature gas-cooled reactors (MHTGRs), and fast breeder reactors. These new designs address the need for passive safety features as well as the requirement of economic competitiveness

Test reactor risk assessment methodology

International Nuclear Information System (INIS)

Jennings, R.H.; Rawlins, J.K.; Stewart, M.E.

1976-04-01

A methodology has been developed for the identification of accident initiating events and the fault modeling of systems, including common mode identification, as these methods are applied in overall test reactor risk assessment. The methods are exemplified by a determination of risks to a loss of primary coolant flow in the Engineering Test Reactor

The Dragon project and high temperature reactor (HTR position)

International Nuclear Information System (INIS)

Shepherd, L.

1981-01-01

After introduction describing the initiation of HTR work at AERE and in West Germany and the USA, the subject is discussed in detail under the headings: the Dragon Reactor Experiment (design and objectives); fuel elements and graphite (description of cooperative research programmes; development of coated fuel particles); helium technology; other Dragon activities. (U.K.)

Italian position paper on heat and mass transfer in the reactor cover gas

International Nuclear Information System (INIS)

Caponetti, R.; Olivieri, P.; Petrazzuolo, F.

1986-01-01

The major effort being made in Italy with regard to the development of fast nuclear reactors is concentrated, as is known, in the construction of the PEC reactor, whose mechanical completion is expected early in 1988. The 116MWt PEC (Prova Elementi di Combustibile; i.e. Fuel Element Testing) reactor is sodium cooled. It is being built to study the behavior of fuel elements under thermal and neutronic conditions similar to those of fast nuclear power stations. Particular attention is being dedicated to safety aspects. This document furnishes a number of construction solutions with regard to that reactor and preparatory approaches to its operation, namely: a brief description of the construction solutions as far as concerns the Closure Head Assembly and the cover gas circuit together with its main components; the description of some test facilities arranged for abatement and measurement of sodium aerosol concentration; a number of preliminary evaluation results obtained thus far with regard to the formation, transport and depositing of sodium aerosols

Report on progress of researches by common utilization of JAERI nuclear facilities, for fiscal 1988

International Nuclear Information System (INIS)

1989-01-01

In 1988, this system called 'Common utilization of JAERI facilities' so far was changed to 'Joint research utilizing JAERI facilities', and by evaluating more positively the function of the General Research Center for Nuclear Energy, it has been emphasized to promote and coordinate the joint research among universities centering around the utilization of JAERI facilities. The total number of the research subjects in fiscal year 1988 reached 138, but the results of 120 of them are collected in this book. General joint research is the standard form of the utilization of various facilities that JAERI has opened to common utilization. Cooperation research is to be carried out by concluding research cooperation contracts between university researchers and JAERI researchers, and the facilities which are not opened to common utilization can be used. In the general joint research, the utilization of irradiation such as activation analysis, radiochemistry, irradiation effect, neutron diffraction and so on and the research using beams are mostly carried out, but in the cooperation research, reactor engineering, reactor materials,, nuclear physics measurement and so on are the main subjects. The total number of visitors in one year was 3829 man-day. (K.I.)

Effect of relative position of the control rods in the power calibration of the running chains of the RP-10 nuclear reactor

International Nuclear Information System (INIS)

Zuniga, Agustin; Arrieta, Rolando

2014-01-01

The monitoring of the thermal power levels of an experimental nuclear reactor, such as the RP-10 reactor is performed by current levels reported by the driving chains and their correlation with the power. The calibration curves for each channel will depend on the neutron distribution observed for the corresponding configuration. One way to modify this configuration is positioning the control rods that can be chosen according to the convenience of providing greater neutron flux in a given irradiation position. This article shows this variation, and consequently the current power curves corresponding thereto. We have chosen two relative positions of the control rods (BC1 and BC2), the first is the one that favors the production of radioisotopes (PPR) (Case 1) and the second favors the use of neutron activation (NAA) (Case 2). The used method was based on the record of the current (up camera, M4) by the acquisition system (SAD) of the reactor and the power from the digital reading of nitrogen 16, which was previously calibrated by thermal balance. The results for the nuclear configuration 30, M4 chain up were: Case 1: I M4 (A) = 2.95 x 10 -6 x P (MW) + 1.45 x 10 -6 R 2 = 9.99 x 10 -1 and Case 2: I M4 (A) = 2.39 x 10 -6 x P (MW) + 5.73 x 10 -7 , R 2 = 9.97 x 10 -1 . The dispersions of the data are in the order of 10 % for low power (under 2 MW), decreasing as the power rises to under 3 % (for values greater than 3 MW). (authors).

Sensitivity of reactor multiplication factor to positions of cross-section ...

Indian Academy of Sciences (India)

V GOPALAKRISHNAN

2017-08-16

Aug 16, 2017 ... an energetically excited system formed as an interme- diate state with the ... The evaluated nuclear data files/libraries are major sources of nuclear ...... reactor case, the net relative deviation is about 330 pcm, which is close to ...

Proposal of a synchro panel meter instrument to replace the obsolete Synchro/Resolver reading device used as position indicator of safety rods assembly of the Brazilian IEA-R1 Nuclear Research Reactor

International Nuclear Information System (INIS)

Toledo, Fabio de; Brancaccio, Franco; Cardenas, Jose Patricio N.

2015-01-01

IPEN (Instituto de Pesquisas Energeticas e Nucleares) was founded in 1956 (as Atomic Energy Institute - IEA) as a facility complex, for the research, development and application, in the nuclear technology field. The institute is recognized as a national leader in nuclear research and development (R and D), including the areas of reactor operation, radiopharmaceuticals, industrial and laboratory applications, materials science and laser technologies and applications. IPEN's main facility is the IEA-R1, nuclear research reactor (NRR), today, the only one in Brazil with a power level suitable for applications in physics, chemistry, biology and engineering. Some radioisotopes are also produced in IEA-R1, for medical and other applications. A common problem faced in the IEA-R1 maintenance is instrumentation obsolescence; spare parts are no more available, because of discontinued production, and an updating program is mandatory, aiming at modernization of old-aged I and C systems. In the presented context, an electronic system is here proposed, as a replacement for the reactor safety (shim) rods assembly position indicator, based on an open-source physical computing platform called Arduino, which includes a simple microcontroller board and a software-code development environment. A mathematical algorithm for the synchro-motor signal processing was developed, and the obtained resolution was better than 1.5%. (author)

Proposal of a synchro panel meter instrument to replace the obsolete Synchro/Resolver reading device used as position indicator of safety rods assembly of the Brazilian IEA-R1 Nuclear Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Toledo, Fabio de; Brancaccio, Franco; Cardenas, Jose Patricio N., E-mail: fatoledo@ipen.br, E-mail: fbrancac@ipen.br, E-mail: ahiru@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2015-07-01

IPEN (Instituto de Pesquisas Energeticas e Nucleares) was founded in 1956 (as Atomic Energy Institute - IEA) as a facility complex, for the research, development and application, in the nuclear technology field. The institute is recognized as a national leader in nuclear research and development (R and D), including the areas of reactor operation, radiopharmaceuticals, industrial and laboratory applications, materials science and laser technologies and applications. IPEN's main facility is the IEA-R1, nuclear research reactor (NRR), today, the only one in Brazil with a power level suitable for applications in physics, chemistry, biology and engineering. Some radioisotopes are also produced in IEA-R1, for medical and other applications. A common problem faced in the IEA-R1 maintenance is instrumentation obsolescence; spare parts are no more available, because of discontinued production, and an updating program is mandatory, aiming at modernization of old-aged I and C systems. In the presented context, an electronic system is here proposed, as a replacement for the reactor safety (shim) rods assembly position indicator, based on an open-source physical computing platform called Arduino, which includes a simple microcontroller board and a software-code development environment. A mathematical algorithm for the synchro-motor signal processing was developed, and the obtained resolution was better than 1.5%. (author)

Sodium cooled fast reactor

Energy Technology Data Exchange (ETDEWEB)

Hokkyo, N; Inoue, K; Maeda, H

1968-11-21

In a sodium cooled fast neutron reactor, an ultrasonic generator is installed at a fuel assembly hold-down mechanism positioned above a blanket or fission gas reservoir located above the core. During operation of the reactor an ultrsonic wave of frequency 10/sup 3/ - 10/sup 4/ Hz is constantly transmitted to the core to resonantly inject the primary bubble with ultrasonic energy to thereby facilitate its growth. Hence, small bubbles grow gradually to prevent the sudden boiling of sodium if an accident occurs in the cooling system during operation of the reactor.

Selection method and device for reactor core performance calculation input indication

International Nuclear Information System (INIS)

Yuto, Yoshihiro.

1994-01-01

The position of a reactor core component on a reactor core map, which is previously designated and optionally changeable, is displayed by different colors on a CRT screen by using data of a data file incorporating results of a calculation for reactor core performance, such as incore thermal limit values. That is, an operator specifies the kind of the incore component to be sampled on a menu screen, to display the position of the incore component which satisfies a predetermined condition on the CRT screen by different colors in the form of a reactor core map. The position for the reactor core component displayed on the CRT screen by different colors is selected and designated on the screen by a touch panel, a mouse or a light pen, thereby automatically outputting detailed data of evaluation for the reactor core performance of the reactor core component at the indicated position. Retrieval of coordinates of fuel assemblies to be data sampled and input of the coordinates and demand for data sampling can be conducted at once by one menu screen. (N.H.)

The fast breeder reactor

International Nuclear Information System (INIS)

Patterson, W.

1990-01-01

The author criticises the United Kingdom Atomic Energy Authority's fast breeder reactor programme in his evidence to the House of Commons Select Committee on Energy in January 1990. He argues for power generation by renewable means and greater efficiency in the use rather than in the generation of electricity. He refutes the arguments for nuclear power on the basis of reduced global warming as he claims support technology produces significant amounts of carbon dioxide in any case. Serious doubts are raised about the costs of a fast breeder reactor programme compared to, say, generation by pressurised water reactors. The idea of a uranium scarcity in several decades is also refuted. The reliability of fast breeder reactor technology is called into question. He argues against reprocessing plutonium for economic, health and safety reasons. (UK)

Workshop on transport for a common ion driver

International Nuclear Information System (INIS)

Olson, C.C.; Lee, E.; Langdon, B.

1994-01-01

This report contains research in the following areas related to beam transport for a common ion driver: multi-gap acceleration; neutralization with electrons; gas neutralization; self-pinched transport; HIF and LIF transport, and relevance to common ion driver; LIF and HIF reactor concepts and relevance to common ion driver; atomic physics for common ion driver; code capabilities and needed improvement

Control rod position detector for nuclear reactor

International Nuclear Information System (INIS)

Kudo, Mitsuru; Fujiwara, Hiroshi.

1981-01-01

Purpose: To improve the reliability of a control rod position detector by detecting a reactive code with a combination of control rod position change signals produced from vertical and horizontal axis decoders, generation an error signal and thus simultaneously detecting the operation of more than two lead switches. Constitution: Horizontal and vertical axis position signals responsive to changes in the control rod position are applied from lead switches connected in a predetermined matrix connection corresponding to the notches of the positions of respective position detecting probes, the reactive output from the decoder is detected by a reactive code detecting circuit, which in turn generates a fault signal, and the control rod position code converted in a notch number generating circuit is converted to a predetermined value indicating invalidity. Accordingly, a fault caused by the simultaneous operation of a plurality of failed lead switches can be effectively detected. (Yoshino, Y.)

Prometheus Project Reactor Module Final Report, For Naval Reactors Information

International Nuclear Information System (INIS)

MJ Wollman; MJ Zika

2006-01-01

The Naval Reactors Prime Contractor Team (NRPCT) led the development of a power plant for a civilian nuclear electric propulsion (NEP) system concept as part of the Prometheus Project. This report provides a summary of the facts, technical insights, and programmatic perspectives gained from this two-year program. The Prometheus Project experience has been extensively documented to better position the US for future space reactor development. Major Technological and engineering challenges exist to develop a system that provides useful electric power from a nuclear fission heat source operating in deep space. General issues include meeting mission requirements in a system that has a mass low enough to launch from earth while assuring public safety and remaining safely shutdown during credible launch accidents. These challenges may be overcome in the future if there is a space mission with a compelling need for nuclear power to drive development. Past experience and notional mission requirements indicate that any useful space reactor system will be unlike past space reactors and existing terrestrial reactors

Fuel assembly for FBR type reactor

International Nuclear Information System (INIS)

Hayashi, Hideyuki.

1995-01-01

Ordinary sodium bond-type fuel pins using nitride fuels, carbide fuels or metal fuels and pins incorporated with hydride moderators are loaded in a wrapper tube at a ratio of from 2 to 10% based on the total number of fuel pins. The hydride moderators are sealed in the hydride moderator incorporated pins at the position only for a range from the upper end to a reactor core upper position of substantially 1/4 of the height of the reactor core from the upper end of the reactor core as a center. Then, even upon occurrence of ULOF (loss of flow rate scram failure phenomenon), it gives characteristic of reducing the power only by a doppler coefficient and not causing boiling of coolant sodium but providing stable cooling to the reactor core. Therefore, a way of thinking on the assurance of passive safety is simplified to make a verification including on the reactor structure unnecessary. In an LMFBR type reactor using the fuel assembly, a critical experiment for confirming accuracy of nuclear design is sufficient for the item required for study and development, which provides a great economical effect. (N.H.)

Accident analysis of heavy water cooled thorium breeder reactor

International Nuclear Information System (INIS)

Yulianti, Yanti; Su’ud, Zaki; Takaki, Naoyuki

2015-01-01

Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The

Accident analysis of heavy water cooled thorium breeder reactor

Energy Technology Data Exchange (ETDEWEB)

Yulianti, Yanti [Department of Physics, University of Lampung Jl. Sumantri Brojonegoro No.1 Bandar Lampung, Indonesia Email: y-yanti@unila.ac.id (Indonesia); Su’ud, Zaki [Department of Physics, Bandung Institute of Technology Jl. Ganesha 10 Bandung, Indonesia Email: szaki@fi.itb.ac.id (Indonesia); Takaki, Naoyuki [Department of Nuclear Safety Engineering Cooperative Major in Nuclear Energy (Graduate School) 1-28-1 Tamazutsumi,Setagayaku, Tokyo158-8557, Japan Email: ntakaki@tcu.ac.jp (Japan)

2015-04-16

Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The

Effect of calcium on microbial aggregation during UASB reactor start-up

Energy Technology Data Exchange (ETDEWEB)

Mahoney, E.M.; Varangu, L.K.; Cairns, W.L.; Kosaric, N.; Murray, R.G.E.

1987-01-01

The dynamics of granule formation were studied using cells from two bench-scale UASB reactors. The objective was to elucidate factors which influence formation and maintenance of highly active self-agglomerated microbial biomass. Simultaneous examination of biological and physical parameters was performed during the start-up of a calcium-positive (100 mg/l) reactor and a reactor without added calcium. The influence of carbon nutrients and Ca++ on the cell surface and microbial aggregation was studied. The granules formed in both reactors but were larger in the calcium-positive reactor in which they settled 3-4 times faster. A higher rate of biomass accumulation also was evident in the calcium-positive reactor and this allowed a more frequent increase in the substrate loading rate and earlier development of the granular sludge. (Refs. 17).

A grid for the accurate positioning of fuel batteries in a reactor core

International Nuclear Information System (INIS)

Berens, T.; Maansson, R.; Gunnarsson, C.

1976-01-01

A grid for the accurate positioning of the fuel batteries in a reactor core, said grid being constituted by a large member of so called first and second metal rails of rectangular cross-section, resting on their upper edge, said first rails being in parallel relationship and at right angles to said second rails, welded coupling and slots being provided at the intersections of said rails, characterized by relatively great height of said first rails and by the relatively small height of said second rails, and also by the construction of said slots in the high rails, said slots being in the form of elongated recesses, the height of which is smaller than the maximum height of the smaller rails, and one long said of which is provided with a few pins pointing towards the other long side and welded to the surface a small height rail located in said recess. (author)

Method of estimating the reactor power distribution

International Nuclear Information System (INIS)

Mitsuta, Toru; Fukuzaki, Takaharu; Doi, Kazuyori; Kiguchi, Takashi.

1984-01-01

Purpose: To improve the calculation accuracy for the power distribution thereby improve the reliability of power distribution monitor. Constitution: In detector containing strings disposed within a reactor core, movable type neutron flux monitors are provided in addition to position fixed type neutron monitors conventionally disposed so far. Upon periodical monitoring, a power distribution X1 is calculated from a physical reactor core model. Then, a higher power position X2 is detected by position detectors and value X2 is sent to a neutron flux monitor driving device to displace the movable type monitors to a higher power position in each of the strings. After displacement, the value X1 is amended by an amending device using measured values from the movable type and fixed type monitors and the amended value is sent to a reactor core monitor device. Upon failure of the fixed type monitors, the position is sent to the monitor driving device and the movable monitors are displaced to that position for measurement. (Sekiya, K.)

Simulating Neutronic Core Parameters in a Research and Test Reactor

International Nuclear Information System (INIS)

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

2011-01-01

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

Reactor containment

International Nuclear Information System (INIS)

Kawabe, Ryuhei; Yamaki, Rika.

1990-01-01

A water vessel is disposed and the gas phase portion of the water vessel is connected to a reactor container by a pipeline having a valve disposed at the midway thereof. A pipe in communication with external air is extended upwardly from the liquid phase portion to a considerable height so as to resist against the back pressure by a waterhead in the pipeline. Accordingly, when the pressure in the container is reduced to a negative level, air passes through the pipeline and uprises through the liquid phase portion in the water vessel in the form of bubbles and then flows into the reactor container. When the pressure inside of the reactor goes higher, since the liquid surface in the water vessel is forced down, water is pushed up into the pipeline. Since the waterhead pressure of a column of water in the pipeline and the pressure of the reactor container are well-balanced, gases in the reactor container are not leaked to the outside. Further, in a case if a great positive pressure is formed in the reactor container, the inner pressure overcomes the waterhead of the column of water, so that the gases containing radioactive aerosol uprise in the pipeline. Since water and the gases flow being in contact with each other, this can provide the effect of removing aerosol. (T.M.)

Computer monitoring of the RB reactor operation

International Nuclear Information System (INIS)

Milovanovic, S.; Pesic, M.; Milovanovic, T.

1998-01-01

Personal computer based acquisition system designed for monitoring of operation of the RB experimental reactor in the Institute of Nuclear Sciences 'Vinca' (former 'Boris Kidric') and experiences acquired during its use are shown in this paper. The monitoring covers generally all nuclear aspects of the reactor operation (start-up, nominal power operation, power changing, shut down and maintenance), but the emphasis is put on: real time (especially fast changing) reactivity measurement; supervising time dependence of the safety rods positions during shut down, and detection of position inaccuracy or failure operation of safety/control rods during the reactor operation or maintenance. (author)

Dispositivo de posicionamiento de muestras biológicas para su irradiación en un canal radial de un reactor nuclear // Biological samples positioning device for irradiations on a radial channel at the nuclear research reactor

Directory of Open Access Journals (Sweden)

Maritza Rodríguez - Gual

2010-05-01

Full Text Available ResumenPor la demanda de un dispositivo experimental para el posicionamiento de las muestras biológicaspara su irradiación en un canal radial de un reactor nuclear de investigaciones en funcionamiento, seconstruyó y se puso en marcha un dispositivo para la colocación y retirada de las muestras en laposición de irradiación de dicho canal. Se efectuaron las valoraciones económicas comparando conotro tipo de dispositivo con las mismas funciones. Este trabajo formó parte de un proyectointernacional entre Cuba y Brasil que abarcó el estudio de los daños inducidos por diferentes tipos deradiación ionizante en moléculas de ADN. La solución propuesta es comprobada experimentalmente,lo que demuestra la validez práctica del dispositivo. Como resultado del trabajo, el dispositivoexperimental para la irradiación de las muestras biológicas se encuentra instalado y funcionando yapor 5 años en el canal radial # 3(BH#3 Palabras claves: reactor nuclear de investigaciones, dispositivo para posicionamiento de muestras,___________________________________________________________________________AbstractFor the demand of an experimental device for biological samples positioning system for irradiationson a radial channel at the nuclear research reactor in operation was constructed and started up adevice for the place and remove of the biological samples from the irradiation channels withoutinterrupting the operation of the reactor. The economical valuations are effected comparing withanother type of device with the same functions. This work formed part of an international projectbetween Cuba and Brazil that undertook the study of the induced damages by various types ofionizing radiation in DNA molecules. Was experimentally tested the proposed solution, whichdemonstrates the practical validity of the device. As a result of the work, the experimental device forbiological samples irradiations are installed and operating in the radial beam hole #3(BH#3

Device for preventing coolant outflow in a reactor

International Nuclear Information System (INIS)

Nemoto, Kiyomitsu; Mochizuki, Keiichi.

1975-01-01

Object: To prevent outflow of coolant from a reactor vessel even in an occurrence of leaking trouble at a low position in a primary cooling system or the like in the reactor vessel. Structure: An inlet at the foremost end of a coolant inlet pipe inserted into a reactor vessel is arranged at a level lower than a core, and a check valve is positioned at a level higher than the core in a rising portion of the inlet. In normal condition, the check valve is pushed up by discharge pressure of a main circulating pump and remains closed, and hence, producing no flow loss of coolant, sodium. However, when a trouble such as rupture occurs at the lower position in the primary cooling system, the attractive force for allowing the coolant to back-flow outside the reactor vessel and the load force of the coolant within the reactor vessel cause the check valve to actuate, as a consequence of which a liquid level of the coolant downwardly moves to the position of the check valve to intake the cover gases into a gas intake, thereby cutting off a flow passage of the coolant to stop outflow thereof. (Kamimura, M.)

Nuclear reactors

International Nuclear Information System (INIS)

Yoshioka, Michiko.

1985-01-01

Purpose: To obtain an optimum structural arrangement of IRM having a satisfactory responsibility to the inoperable state of a nuclear reactor and capable of detecting the reactor power in an averaged manner. Constitution: As the structural arrangement of IRM, from 6 to 16 even number of IRM are bisected into equial number so as to belong two trip systems respectively, in which all of the detectors are arranged at an equal pitch along a circumference of a circle with a radius rl having the center at the position of the central control rod in one trip system, while one detector is disposed near the central control rod and other detectors are arranged substantially at an equal pitch along the circumference of a circle with a radius r2 having the center at the position for the central control rod in another trip system. Furthermore, the radius r1 and r2 are set such that r1 = 0.3 R, r2 = 0.5 R in the case where there are 6 IRM and r1 = 0.4 R and R2 = 0.8 R where there are eight IRM where R represents the radius of the reactor core. (Kawakami, Y.)

Report of researches by common utilization of facilities in Kyoto University Research Reactor Institute, latter half of fiscal year 1982

International Nuclear Information System (INIS)

1983-01-01

The technical report of the Kyoto University Research Reactor Institute is published any time to immediately report on the results of the functional tests of various experimental facilities, the test results for the products made for trial, radiation control, the situation of waste treatment, the data required for research and experiment such as the reports of study meetings, the conspicuous results obtained amid researches, new processes, and the discussion on other papers and reports. In this report, the title, the names of reporters and the summary of 65 researches carried out by the common utilization of the facilities in the Kyoto University Research Reactor Institute are collected. The themes of the researches are such as Moessbauer spectroscopic study of ferrocene and its derivative iodides by I-129, decomposition of cadmium telluride during heat treatment, element distribution in resource living things and environmental substances produced in northern ocean, radioactivation analysis of trace elements in blood of tumor-bearing animals, radioactivation analysis of noble metal elements in geochemical samples, relaxation phenomena by gamma-gamma perturbation angle correlation, separation of components in Allende meteorite and their radioactivation analysis, measurement of cross section of Pa-231 (n, gamma) reaction and others. (Kako, I.)

Report of researches by common utilization of facilities in Kyoto University Research Reactor Institute, first half of fiscal year 1982

International Nuclear Information System (INIS)

1983-01-01

The technical report of the Kyoto University Research Reactor Institute is published any time to immediately report on the results of the functional tests of various experimental facilities, the test results for the products made for trial, radiation control, the situation of waste treatment, the data required for research and experiment such as the reports of study meetings, the conspicuous results obtained amid researches, new processes, and the discussion on other papers and reports. In this report, the title, the names of reporters and the summary of 47 researches carried out by the common utilization of the facilities in the Kyoto University Research Reactor Institute are collected. The themes of the researches are such as diffusion of impurities ion-implanted in silicon into natural oxide films, origin of igneous rocks by trace element distribution study, element distribution in black ore and its accompanying rocks and origin of black ore, reprocessing of molten salt fuel of thorium group, forerunning martensite transformation of Fe-Pt invar alloy, change of nucleic acid component to recoil tritium at cryogenic temperature, gamma irradiation effect of KC1 containing Pb 2+ , radiation effect on cadmium halide crystals and impurity metallic ions and others. (Kako, I.)

The timing of reactor dismantling

International Nuclear Information System (INIS)

Roberts, P.

2000-01-01

Work has been progressing across the world for the decommissioning of nuclear reactors. The initial work focused on the early, complete dismantling but this was associated with small size reactors and was done for experimental or demonstration purposes. The situation now is that an increasing number of full size power reactors are being shutdown and decision are being made as to the decommissioning strategy to be applied, e.g. with respect to the appropriate timing of reactor dismantling. There are two basic approaches to the timing of reactor dismantling, which are to either proceed with dismantling on an early time scale or to delay it for a period of years. There are a number of examples worldwide of both approaches being taken but one common feature of the approach taken by most countries is that decisions are made on a case by case basis, taking account of relevant factors, and as a result the strategy can vary from reactor to reactor and from country to country. Decisions on timing take account of the following main factors: safety, radioactive decay, financial factors, radioactive waste, reactor type, technology, repository availability, site re-use, regulatory standards, plant knowledge/records, other issues

Effects of Common Cause Failure on Electrical Systems

International Nuclear Information System (INIS)

Pepper, Kevin

2015-01-01

The essential electrical systems of reactor designs have developed progressively with an increased focus on the use of redundant, segregated and independent safety system equipment 'trains'. In this arrangement, essential safety functions associated with safe shutdown and cooling of the reactor are replicated on near identical electrical systems with each of the trains of safety system equipment supported by a fully rated standby generator. Development in designs has seen the number of trains increased to enable maintenance to be undertaken with reactors at power, improving the economics of the units whilst maintaining nuclear safety. This paper provides a background to common cause failure and provides examples where supporting guidance and international experience is available. It also highlights the regulatory guidance available to UK licensees. Recent examples of common cause failures on plant in the UK are presented together with an issue identified during the recent Generic Design Assessment review of new reactor designs within the UK. It was identified that one design was claiming a very low probability of failure associated with the loss of a single break and no-break voltage level, orders of magnitude below the target figure within ONR's Safety Assessment Principles. On closer scrutiny it was established that a significant safety function provided from identical low voltage switchboards would be lost in the event of a common cause failure affecting these boards. The paper will explain the action that has been taken by the requesting party to improve the resilience of the design and how this impacts on the ONR reliability targets for reactor designs within the UK. (authors)

Reactor container

International Nuclear Information System (INIS)

Oyamada, Osamu; Furukawa, Hideyasu; Uozumi, Hiroto.

1979-01-01

Purpose: To lower the position of an intermediate slab within a reactor container and fitting a heat insulating material to the inner wall of said intermediate slab, whereby a space for a control rod exchanging device and thermal stresses of the inner peripheral wall are lowered. Constitution: In the pedestal at the lower part of a reactor pressure vessel there is formed an intermediate slab at a position lower than diaphragm floor slab of the outer periphery of the pedestal thereby to secure a space for providing automatic exchanging device of a control rod driving device. Futhermore, a heat insulating material is fitted to the inner peripheral wall at the upper side of the intermediate slab part, and the temperature gradient in the wall thickness direction at the time of a piping rupture trouble is made gentle, and thermal stresses at the inner peripheral wall are lowered. (Sekiya, K.)

Reactor-vessel-sectioning demonstration

International Nuclear Information System (INIS)

Lundgren, R.A.

1981-09-01

A technical demonstration was successfully completed of simulated reactor vessel sectioning using the combined techniques of air arc gouging and flame cutting. A 4-ft x 3-ft x 9-in. thick sample was fabricated of A36 carbon steel to simulate a reactor vessel wall. A 1/4-in. layer of stainless steel (SS) was tungsten inert gas (TIG)-welded to the carbon steel. Several techniques were considered to section the simulated reactor vessel; air arc gouging was selected to penetrate the stainless steel, and flame cutting was selected to sever the carbon steel. Three sectioning operations were demonstrated. For all three, the operating parameters were the same; but the position of the sample was varied. For the first cut, the sample was placed in a horizontal position, and it was successfully severed from the SS side. For the second cut, the sample was turned over and cut from the carbon steel side. Cutting from the carbon steel side has the advantages of cost reduction

A model for nuclear research reactor dynamics

Energy Technology Data Exchange (ETDEWEB)

Barati, Ramin, E-mail: Barati.ramin@aut.ac.ir; Setayeshi, Saeed, E-mail: setayesh@aut.ac.ir

2013-09-15

Highlights: • A thirty-fourth order model is used to simulate the dynamics of a research reactor. • We consider delayed neutrons fraction as a function of time. • Variable fuel and temperature reactivity coefficients are used. • WIMS, BORGES and CITVAP codes are used for initial condition calculations. • Results are in agreement with experimental data rather than common codes. -- Abstract: In this paper, a useful thirty-fourth order model is presented to simulate the kinetics and dynamics of a research reactor core. The model considers relevant physical phenomena that govern the core such as reactor kinetics, reactivity feedbacks due to coolant and fuel temperatures (Doppler effects) with variable reactivity coefficients, xenon, samarium, boron concentration, fuel burn up and thermal hydraulics. WIMS and CITVAP codes are used to extract neutron cross sections and calculate the initial neuron flux respectively. The purpose is to present a model with results similar to reality as much as possible with reducing common simplifications in reactor modeling to be used in different analyses such as reactor control, functional reliability and safety. The model predictions are qualified by comparing with experimental data, detailed simulations of reactivity insertion transients, and steady state for Tehran research reactor reported in the literature and satisfactory results have been obtained.

Structure of thermonuclear reactor wall

International Nuclear Information System (INIS)

Yamazaki, Seiichiro.

1991-01-01

In a thermonuclear reactor wall, there has been a worry that the brazing material is melted by high temperature heat and particle load, to peel off the joined portion and the protecting material is destroyed by temperature elevation, to expose the heat sink material. Then, in the reactor core structures of a thermonuclear reactor, such as a divertor plate comprising a protecting material made of carbon material and the heat sink material joined by brazing, a plate material made of a so-called refractory metal having a high atomic number such as tungsten, molybdenum or the alloy thereof is embedded or attached to an accurate position of the protecting material. This can prevent the brazing portion from destruction by escaping electrons generated upon occurrence of abnormality in the thermonuclear reactor, and peeling or destroy of the protecting material and the heat sink material. Sufficient characteristics of plasmas can always be maintained by disposing a material having a small atomic number, for example, carbon material, to the position facing to the plasmas. (N.H.)

Culham conceptual Tokamak reactor MkII. Conceptual layout of buildings for a twin reactor power station

International Nuclear Information System (INIS)

Guthrie, J.A.S.; Harding, N.H.

1981-01-01

This paper discusses the conceptual design of the nuclear complex of a 2400 MWe twin fusion reactor power station utilising common services and a single containment building. The design is based upon environmental and maintenance logistical requirements, the provision of adequate storage, workshop and construction facilities and the constraints imposed by the geometry of the main and auxiliary reactor coolant systems. (author)

A review of the United Kingdom fast reactor programme

International Nuclear Information System (INIS)

Bramman, J.I.; Hickey, H.B.; Whitlow, W.H.; Frew, J.D.; Gregory, C.V.

1990-01-01

Total energy consumption in the UK in 1989 was 340 million tonnes of coal or coal equivalent, made up as follows: coal 31%, petroleum 35%, natural gas 24%, nuclear electricity 8%, hydroelectricity 1% and imported electricity 1%. About half of the nuclear electricity generated came from 14 Advanced Gas-Cooled Reactors (AGRs) and about half from the 24 older gas-cooled Magnox reactors, one Steam-Generating Heavy-Water Reactor (SGHWR) and one fast reactor (the Prototype Fast Reactor, PFR, at Dounreay). The privatization of the Electricity Supply Industry (ESI) in the UK is proceeding. On 9 November 1989, however, it was announced by the Secretary of State for Energy that the privatization plan would be changed and that the CEGB's nuclear stations were to remain in state ownership, through the formation of an additional company, Nuclear Electric. At the same time, the Secretary of State for Scotland announced the formation of a similar state-owned company, Scottish Nuclear. Nuclear Electric was asked, in the interim, to examine priorities in the whole nuclear field with particular reference to the improvement of the economics and performance of existing reactors, to the development of the Sizewell and alternative reactors and to the development of longer-term options such as the fast reactor and fusion. Nuclear Electric has been asked to formulate its new policy by June 1990. The PFR programme will continue to be funded by the UK government until March 1994. AEA Technology is endeavouring to find alternative funding to maintain the operation of the PFR until at least the year 2000. The House of Commons Select Committee on Energy stated in its report that the fast reactor ''is a matter for the British Government to foster as a long-term option for the generation of electricity in this country'', and recommended that in the interim the Government reassesses its position on this new technology in the light of increasing concern about CO 2 emissions and the long

Proceedings of the Third MDEP Conference on New Reactor Design Activities

International Nuclear Information System (INIS)

Holahan, Gary; Burkhart, Lawrence; Santos, Daniel J.; Mourlon, Sophie; Virolainen, Tapani; Lagarde, Dominique; Rasmussen, Richard; Johnson, Gary; Jun Wei, Xue; Prinja, Nawal; Ennis, Kevin; Buisine, Denis

2014-05-01

The MDEP is a unique multinational initiative leveraging the resources and knowledge of national regulators to review new reactor designs. The programme celebrated its 10. anniversary in 2006 and is recognised as an effective framework for regulatory co-operation and harmonisation. Over the first 10 years of its existence, MDEP has published 24 common positions and 13 technical reports. In order to gather feedback on its current activities and discuss its future, MDEP organised its third conference at Bethesda, Maryland, USA on 14-15 May 2014. The event provided a forum for MDEP stakeholders (including industry representatives, standard development organisations and other international organisations) to share the results of their engagement with the program and to deliver presentations on ongoing activities related to new reactor licensing. The two-day conference included sessions on Design-Specific Working Groups, Commissioning Activities, Vendor Inspection, Digital Instrumentation and Control, New reactor activities related to the Fukushima Daiichi accident, codes and standards harmonisation, MDEP Way Forward and Related Activities. This document gathers the available presentations (slides) given at the conference

Intrinsically secure fast reactors with dense cores

International Nuclear Information System (INIS)

Slessarev, Igor

2007-01-01

Secure safety, resistance to weapons material proliferation and problems of long-lived wastes remain the most important 'painful points' of nuclear power. Many innovative reactor concepts have been developed aimed at a radical enhancement of safety. The promising potential of innovative nuclear reactors allows for shifting accents in current reactor safety 'strategy' to reveal this worth. Such strategy is elaborated focusing on the priority for intrinsically secure safety features as well as on sure protection being provided by the first barrier of defence. Concerning the potential of fast reactors (i.e. sodium cooled, lead-cooled, etc.), there are no doubts that they are able to possess many favourable intrinsically secure safety features and to lay the proper foundation for a new reactor generation. However, some of their neutronic characteristics have to be radically improved. Among intrinsically secure safety properties, the following core parameters are significantly important: reactivity margin values, reactivity feed-back and coolant void effects. Ways of designing intrinsically secure safety features in fast reactors (titled hereafter as Intrinsically Secure Fast Reactors - ISFR) can be found in the frame of current reactor technologies by radical enhancement of core neutron economy and by optimization of core compositions. Simultaneously, respecting resistance to proliferation, by using non-enriched fuel feed as well as a core breeding gain close to zero, are considered as the important features (long-lived waste problems will be considered in a separate paper). This implies using the following reactor design options as well as closed fuel cycles with natural U as the reactor feed: ·Ultra-plate 'dense cores' of the ordinary (monolithic) type with negative total coolant void effects. ·Modular type cores. Multiple dense modules can be embedded in the common reflector for achieving the desired NPP total power. The modules can be used also independently (as

Cooling device for reactor container

International Nuclear Information System (INIS)

Arai, Kenji.

1996-01-01

Upon assembling a static container cooling system to an emergency reactor core cooling system using dynamic pumps in a power plant, the present invention provides a cooling device of lowered center of gravity and having a good cooling effect by lowering the position of a cooling water pool of the static container cooling system. Namely, the emergency reactor core cooling system injects water to the inside of a pressure vessel using emergency cooling water stored in a suppression pool as at least one water source upon loss of reactor coolant accident. In addition, a cooling water pool incorporating a heat exchanger is disposed at the circumference of the suppression pool at the outside of the container. A dry well and the heat exchanger are connected by way of steam supply pipes, and the heat exchanger is connected with the suppression pool by way of a gas exhaustion pipe and a condensate returning pipeline. With such a constitution, the position of the heat exchanger is made higher than an ordinary water level of the suppression pool. As a result, the emergency cooling water of the suppression pool water is injected to the pressure vessel by the operation of the reactor cooling pumps upon loss of coolant accident to cool the reactor core. (I.S.)

Study on the technical feasibility of Fission-Track dating at two irradiation positions of the RA-6 research reactor

International Nuclear Information System (INIS)

Dorval, Eric

2005-01-01

The method of Fission-Track dating is based upon the detection of the damage caused by fission fragments from the Uranium contained in geological samples.In order to determine the age of a sample, both the amount of spontaneous fissions occurred and the Uranium concentration must be known.The latter requires the irradiation of the samples inside a reactor with a well-thermalized flux, so that fissions are induced over 235 U targets only. Therefore, the Uranium concentration may be determined.The main inconvenient presented by the irradiation sites at the RA-6 MTR-type reactor is that neutron flux is not completely thermal there, which means that fissions due to epithermal and fast neutrons will not be negligible.In the same way, tracks due to fissions of 238 U and 232 Th will be detected. In order to know the corrections that must be applied to those measurements performed in this reactor, it is necessary to characterize fast flux.Because of it, this laboratory's gamma spectrometry equipment had to be calibrated. After that, several activation detectors were irradiated and results were analyzed. Finally, it was determined that it is feasible to Fission-Track date at the I6 position. However, limitations associated to this method were analyzed for the values of flux measured in the different sites

Advanced liquid metal fast breeder reactor designs

International Nuclear Information System (INIS)

Sayles, C.W.

1978-01-01

Fast Breeder reactor power plants in the 1000-1200 MW(e) range are being built overseas and are being designed in this country. While these reactors have many characteristics in common, a variety of different approaches have been adopted for some of the major features. Some of those alternatives are discussed

Needs for development in nondestructive testing for advanced reactor systems

International Nuclear Information System (INIS)

McClung, R.W.

1978-01-01

The needs for development of nondestructive testing (NDT) techniques and equipment were surveyed and analyzed relative to problem areas for the Liquid-Metal Fast Breeder Reactor, the Molten-Salt Breeder Reactor, and the Advanced Gas-Cooled Reactor. The paper first discusses the developmental needs that are broad-based requirements in nondestrutive testing, and the respective methods applicable, in general, to all components and reactor systems. Next, the requirements of generic materials and components that are common to all advanced reactor systems are examined. Generally, nondestructive techniques should be improved to provide better reliability and quantitativeness, improved flaw characterization, and more efficient data processing. Specific recommendations relative to such methods as ultrasonics, eddy currents, acoustic emission, radiography, etc., are made. NDT needs common to all reactors include those related to materials properties and degradation, welds, fuels, piping, steam generators, etc. The scope of applicability ranges from initial design and material development stages through process control and manufacturing inspection to in-service examination

LWR type reactor

International Nuclear Information System (INIS)

Kato, Kiyoshi.

1993-01-01

A water injection tank in an emergency reactor core cooling system is disposed at a position above a reactor pressure vessel. A liquid phase portion of the water injection tank and an inlet plenum portion in the reactor pressure vessel are connected by a water injection pipe. A gas phase portion of the water injection tank and an upper portion in the reactor pressure vessel are connected by a gas ventilation pipe. Hydraulic operation valves are disposed in the midway of the water injection pipe and the gas ventilation pipe respectively. A pressure conduit is disposed for connecting a discharge port of a main recycling pump and the hydraulic operation valve. In a case where primary coolants are not sent to the main recycling pump by lowering of a liquid level due to loss of coolants or in a case where the main recycling pump is stopped by electric power stoppage or occurrence of troubles, the discharge pressure of the main recycling pump is lowered. Then, the hydraulic operation valve is opened to release the flow channel, then, boric acid water in the water injection tank is sent into the reactor by a falling head, to lead the reactor to a scram state. (I.N.)

Method of detecting fuel failure in FBR type reactor and method of estimating fuel failure position

International Nuclear Information System (INIS)

Sonoda, Yukio; Tamaoki, Tetsuo

1989-01-01

Noise components in a normal state contained in detection signals from delayed neutron monitors disposed to a coolant inlet, etc. of an intermediate heat exchanger are forecast by self-recurring model and eliminated, and resultant detection signals are monitored thereby detecting fuel failure high sensitivity. Subsequently, the reactor is controlled to a low power operation state and a new self-recurring model to the detection signals from the delayed neutron monitors are prepared. Then, noise components in this state are removed and control rods near the delayed neutron monitors are extracted in a short stroke successively to examine the change of response of the delayed neutron monitors. Accordingly, the failed position for each of the fuels can be estimated at a level of one fuel assembly or a level of several assemblies containing the above-mentioned fuel assembly. Since the fuel failure can be detected at a high sensitivity and the position can be estimated, diffusion of abnormality can be prevented and plant shutdown for fuel exchange can be minimized. (I.S.)

Thermosyphoning in the CANDU reactor

International Nuclear Information System (INIS)

Spinks, N.J.; Wright, A.C.D.; Caplan, M.Z.; Prawirosoehardjo, S.; Gulshani, P.

1984-01-01

Thermosyphoning is defined as the natural convective flow of primary coolant over the boilers. It is the predicted mode of heat transport from core to boilers in many postulated scenarios for CANDU reactor safety analysis. The scenarios encompass a wide range of boundary conditions in reactor power, secondary temperature and primary coolant inventory. Loss of pumping of the primary coolant is a common feature. Thermosyphoning is single or two-phase depending on the boundary conditions. The paper describes the important thermohydraulic characteristics of thermosyphoning in CANDU reactors with emphasis on two-phase thermosyphoning. It utilizes predictions of a transient thermohydraulics computer code and describes experiments done for the purpose of verifying these predictions. Predictions are compared with single-phase thermosyphoning tests done during commissioning of the Gentilly-2 and Point Lepreau CANDU 600 reactors. (orig.)

A review of the UK fast reactor programme

International Nuclear Information System (INIS)

Picker, C.; Ainsworth, K.F.

1996-01-01

The general position with regard to nuclear power and fast reactors in UK during 1995 is described. The status of fast reactor studies made in UK is outlined and a description and statement regarding the conclusions of the programme of studies associated with the closure of the Prototype Fast Reactor is included. (author)

A review of the UK fast reactor programme

Energy Technology Data Exchange (ETDEWEB)

Picker, C [AEA Technolgy plc, Risley, Warrington, Cheshire (United Kingdom); Ainsworth, K F [British Nuclear Fuels plc, Sellafield, Cumbria (United Kingdom)

1996-07-01

The general position with regard to nuclear power and fast reactors in UK during 1995 is described. The status of fast reactor studies made in UK is outlined and a description and statement regarding the conclusions of the programme of studies associated with the closure of the Prototype Fast Reactor is included. (author)

MDEP Technical Report TR-DICWG-01. Technical Report on the Relational Structure of the Digital Instrumentation and Controls Working Group (DICWG) Common Positions

International Nuclear Information System (INIS)

2016-10-01

concern arising from regulatory activities in member countries, safety implications, or the general need to develop a common understanding of the topic. DICWG provides a forum for nuclear regulatory organisations from around the world that is vital in demonstrating to the industry that regulatory organisations are actively engaged with these topics and understand their significance. This document describes the relational structure between the MDEP DICWG common positions (CPs), as well as addressing their scope, content and how they relate to the mission of DICWG

Today's attitudes and future prospects of fast reactors in Italy

International Nuclear Information System (INIS)

Barabaschi, S.; Cicognani, G.; Pierantoni, F.

1982-01-01

The Italian fast reactor programme is reviewed. The 15 year collaboration with France has resulted in the construction of the PEC reactor, development of the Superphenix-1 and a common R and D programme for future large fast reactors. The CNEN 4th five year (1980-84) plan is outlined. The budget breakdown for different areas shows the importance attached to the fast reactor. (U.K.)

FFTF and CRBRP reactor vessels

International Nuclear Information System (INIS)

Morgan, R.E.

1977-01-01

The Fast Flux Test Facility (FFTF) reactor vessel and the Clinch River Breeder Reactor Plant (CRBRP) reactor vessel each serve to enclose a fast spectrum reactor core, contain the sodium coolant, and provide support and positioning for the closure head and internal structure. Each vessel is located in its reactor cavity and is protected by a guard vessel which would ensure continued decay heat removal capability should a major system leak develop. Although the two plants have significantly different thermal power ratings, 400 megawatts for FFTF and 975 megawatts for CRBRP, the two reactor vessels are comparable in size, the CRBRP vessel being approximately 28% longer than the FFTF vessel. The FFTF vessel diameter was controlled by the space required for the three individual In-Vessel Handling Machines and Instrument Trees. Utilization of the triple rotating plug scheme for CRBRP refueling enables packaging of the larger CRBRP core in a vessel the same diameter as the FFTF vessel

Simulation development for TRIGA reactor

International Nuclear Information System (INIS)

Handoyo, D.

1997-01-01

A simulator of the dynamic of TRIGA reactor has been made. this simulator is meant to study the reactor kinetic behavior and for operator training to more assure the safety and the reliability of the real operation of TRIGA reactor. the simulator consists of PC (Personal Computer) for processing the calculation of reactivity, neutron flux, period, ect and control panel for regulating the input data such as the change of power range, control rod position as well as cooling flow rate. the result will be displayed on screen monitor of personal computer as given in the real control room of TRIGA reactor. the output of simulator will be verified by comparing with measurement result in the real TRIGA MARK II reactor of Musashi institute of technology. for the change of reactivity of 0.3, 0.5 and 0.7 the reactor power and fuel temperature between the simulator and measurements are comparable

Graphite materials for nuclear reactors

International Nuclear Information System (INIS)

Oku, Tatsuo

1991-01-01

Graphite materials have been used in the nuclear fission reactors from the beginning of the reactor development for the speed reduction and reflection of neutron. Graphite materials are used both as a moderator and as a reflector in the core of high temperature gas-cooled reactors, and both as a radiation shielding material and as a reflector in the surrounding of the core for the fast breeder reactor. On the other hand, graphite materials are being positively used as a first wall of plasma as it is known that low Z materials are useful for holding high temperature plasma in the nuclear fusion devices. In this paper the present status of the application of graphite materials to the nuclear fission reactors and fusion devices (reactors) is presented. In addition, a part of results on the related properties to the structural design and safety evaluation and results examined on the subjects that should be done in the future are also described. (author)

Review of design technology of control rod position indicators

International Nuclear Information System (INIS)

Yu, Je Yong; Huh, Hyung; Kim, Ji Ho; Kim, Jong In; Chang, Moon Hee

1999-10-01

An integral reactor SMART is under development at KAERI. The design characteristics of SMART are radically different from those employer in currently operating loop type water reactors in Korea. The objective of this report is to review the design technology of position indicator, and to study the various sensors which can be used in rod position indicator. Design criteria that rod position indicator should satisfy are also examined. Following position indicators are reviewed in this report. 1. Digital positioning indicator (DRPI), 2. Reed switch type position indicator (RSPT), 3. Choke sensor type position indicator, 4. Ultrasonic sensor type position indicator, 5. Comparison of each position indicator. (author)

Role of Halden Reactor Project for world-wide nuclear energy development

Energy Technology Data Exchange (ETDEWEB)

McGrath, M.A.; Volkov, B.

2011-07-01

The great interest for utilization of nuclear materials to produce energy in the middle of last century needed special investigations using first class research facilities. Common problems in the area of nuclear fuel development motivated the establishment of joint research efforts. The OECD Halden Reactor Project (HRP) is a good example of such a cooperative research effort, which has been performing for more than 50 years. During that time, the Halden Reactor evolved from a prototype heavy water reactor envisaged as a power source for different applications to a research reactor that is able to simulate in-core conditions of modern commercial power reactors. The adaptability of the Halden Reactor enables the HRP to be an important international test facility for nuclear fuels and materials development. The long-term international cooperation is based on the flexible HRP organizational structure which also provides the continued success. [1,2] This paper gives a brief history of the Halden Reactor Project and its contribution to world-wide nuclear energy development. Recent expansion of the Project to the East and Asian countries may also assist and stimulate the development of a nuclear industry within these countries. The achievements of the HRP rely on the versatility of the research carried out in the reactor with reliable testing techniques and in-pile instrumentation. Diversification of scientific activity in the areas of development of alternative energy resources and man-machine technology also provide the HRP with a stable position as one of the leaders in the world scientific community. All of these aspects are described in this paper together with current experimental works, including the investigation of ULBA (Kazakhstan) production fuel in comparison with other world fuel suppliers, as well as other future and prospective plans of the Project.(Author)

Role of Halden Reactor Project for world-wide nuclear energy development

International Nuclear Information System (INIS)

McGrath, M.A.; Volkov, B.

2011-01-01

The great interest for utilization of nuclear materials to produce energy in the middle of last century needed special investigations using first class research facilities. Common problems in the area of nuclear fuel development motivated the establishment of joint research efforts. The OECD Halden Reactor Project (HRP) is a good example of such a cooperative research effort, which has been performing for more than 50 years. During that time, the Halden Reactor evolved from a prototype heavy water reactor envisaged as a power source for different applications to a research reactor that is able to simulate in-core conditions of modern commercial power reactors. The adaptability of the Halden Reactor enables the HRP to be an important international test facility for nuclear fuels and materials development. The long-term international cooperation is based on the flexible HRP organizational structure which also provides the continued success. [1,2] This paper gives a brief history of the Halden Reactor Project and its contribution to world-wide nuclear energy development. Recent expansion of the Project to the East and Asian countries may also assist and stimulate the development of a nuclear industry within these countries. The achievements of the HRP rely on the versatility of the research carried out in the reactor with reliable testing techniques and in-pile instrumentation. Diversification of scientific activity in the areas of development of alternative energy resources and man-machine technology also provide the HRP with a stable position as one of the leaders in the world scientific community. All of these aspects are described in this paper together with current experimental works, including the investigation of ULBA (Kazakhstan) production fuel in comparison with other world fuel suppliers, as well as other future and prospective plans of the Project.(Author)

Proposal of world network on material testing reactors

International Nuclear Information System (INIS)

Takemoto, Noriyuki; Izumo, Hironobu; Hori, Naohiko; Ishitsuka, Etsuo; Ishihara, Masahiro

2011-01-01

Establishment of an international cooperation system of worldwide testing reactor network (world network) is proposed in order to achieve efficient facility utilization and provide high quality irradiation data by role sharing of irradiation tests with materials testing reactors in the world. As for the first step, mutual understanding among materials testing reactors is thought to be necessary. From this point, an international symposium on materials testing reactors (ISMTR) was held to construct the world network from 2008, and a common understanding of world network has begun to be shared. (author)

Research reactor utilization. Summary reports of three study group meetings: Irradiation techniques at research reactors, held in Istanbul 15-19 November 1965; Research reactor operation and maintenance problems, held in Caracas 6-10 December 1965; and Research reactor utilization in the Far East, held in Lucas Heights 28 February - 4 March 1966

International Nuclear Information System (INIS)

1967-01-01

The three sections of this book, which are summary reports of three Study Group meetings of the IAEA: Irradiation techniques at research reactors, Istanbul, 15-19 November 1965; Research reactor operation and maintenance problems, Caracas, 6-10 December 1965; and Research reactor utilization in the Far East, Lucas Heights, Australia, 28 February - 4 March 1966. These meetings were the latest in a series designed to promote efficient utilization of research reactors, to disseminate information on advances in techniques, to discuss common problems in reactor operations, and to outline some advanced areas of reactor-based research. (author)

Multiple Regions of a Cortical Network Commonly Encode the Meaning of Words in Multiple Grammatical Positions of Read Sentences.

Science.gov (United States)

Anderson, Andrew James; Lalor, Edmund C; Lin, Feng; Binder, Jeffrey R; Fernandino, Leonardo; Humphries, Colin J; Conant, Lisa L; Raizada, Rajeev D S; Grimm, Scott; Wang, Xixi

2018-05-16

Deciphering how sentence meaning is represented in the brain remains a major challenge to science. Semantically related neural activity has recently been shown to arise concurrently in distributed brain regions as successive words in a sentence are read. However, what semantic content is represented by different regions, what is common across them, and how this relates to words in different grammatical positions of sentences is weakly understood. To address these questions, we apply a semantic model of word meaning to interpret brain activation patterns elicited in sentence reading. The model is based on human ratings of 65 sensory/motor/emotional and cognitive features of experience with words (and their referents). Through a process of mapping functional Magnetic Resonance Imaging activation back into model space we test: which brain regions semantically encode content words in different grammatical positions (e.g., subject/verb/object); and what semantic features are encoded by different regions. In left temporal, inferior parietal, and inferior/superior frontal regions we detect the semantic encoding of words in all grammatical positions tested and reveal multiple common components of semantic representation. This suggests that sentence comprehension involves a common core representation of multiple words' meaning being encoded in a network of regions distributed across the brain.

The Integral Fast Reactor (IFR) concept

International Nuclear Information System (INIS)

Till, C.E.; Chang, Y.I.

1989-01-01

In addition to maintaining the viability of its present commercial nuclear technology, a principal challenge in the US in the 1990s and beyond will be to regain and maintain a position among the world leadership in advanced reactor research and development. In this paper we'll discuss factors which we believe should today provide the rationale and focus for advanced reactor R and D, and we will then review the status of the major US effort, the Integral Fast Reactor (IFR) program

Economic performance of liquid-metal fast breeder reactor and gas-cooled fast reactor radial blankets

International Nuclear Information System (INIS)

Tsoulfanidis, N.; Jankhah, M.H.

1979-01-01

The economic performance of the radial blanket of a liquid-metal fast breeder reactor (LMFBR) and a gas-cooled fast reactor (GCFR) has been studied based on the calculation of the net financial gain as well as the value of the levelized fuel cost. The necessary reactor physics calculations have been performed using the code CITATION, and the economic analysis has been carried out with the code ECOBLAN, which has been written for that purpose. The residence time of fuel in the blanket is the main variable of the economic analysis. Other parameters that affect the results and that have been considered are the value of plutonium, the price of heat, the effective cost of money, and the holdup time of the spent fuel before reprocessing. The results show that the radial blanket of both reactors is a producer of net positive income for a broad range of values of the parameters mentioned above. The position of the fuel in the blanket and the fuel management scheme applied affect the monetary gain. There is no significant difference between the economic performance of the blanket of an LMFBR and a GCFR

Detailed neutronic study of the power evolution for the European Sodium Fast Reactor during a positive insertion of reactivity

Energy Technology Data Exchange (ETDEWEB)

Facchini, A.; Giusti, V.; Ciolini, R. [Department of Civil and Industrial Engineering (DICI), University of Pisa, Largo Lucio Lazzarino 2, I-56126 Pisa (Italy); Tuček, K.; Thomas, D. [Joint Research Centre, Institute for Energy and Transport (JRC - IET), European Commission, P.O. Box 2, NL-1755 ZG Petten (Netherlands); D' Agata, E., E-mail: elio.dagata@ec.europa.eu [Joint Research Centre, Institute for Energy and Transport (JRC - IET), European Commission, P.O. Box 2, NL-1755 ZG Petten (Netherlands)

2017-03-15

Highlights: • This paper studies the effect of an unexpected runway of a control rod in the ESFR. • The power peaked fuel pin within the core was identified. • The increase of the fission power density of the fuel pin has been evaluated. • Radial/axial fission power density of the power peaked fuel pin has been evaluated. - Abstract: The new reactor concepts proposed in the Generation IV International Forum require the development and validation of new components and new materials. Inside the Collaborative Project on the European Sodium Fast Reactor, several accidental scenario have been studied. Nevertheless, none of them coped with mechanical safety assessment of the fuel cladding under accidental conditions. Among the accidental conditions considered, there is the unprotected transient of overpower (UTOP), due to the insertion, at the end of the first fuel cycle, of a positive reactivity into the reactor core as a consequence of the unexpected runaway of one control rod. The goal of the study was the search for a detailed distribution of the fission power, in the radial and axial directions, within the power peaked fuel pin under the above accidental conditions. Results show that after the control rod ejection an increase from 658 W/cm{sup 3} to 894 W/cm{sup 3}, i.e. of some 36%, is expected for the power peaked fuel pin. This information will represent the base to investigate, in a future work, the fuel cladding safety margin.

Application of non-destructive testing and in-service inspections to research reactors and preparation of ISI programme and manual for WWR-C research reactors

International Nuclear Information System (INIS)

Khattab, M.

1996-01-01

The present report gives a review on the results of application of non-destructive testing and in-service inspections to WWR-C reactors in different countries. The major problems related to reactor safety and the procedure of inspection techniques are investigated to collect the experience gained from this type of reactors. Exchangeable experience in solving common problems in similar reactors play an important role in the effectiveness of their rehabilitation programmes. 9 figs., 4 tabs

Utilizing horizontal reactors channels for neutron therapy

International Nuclear Information System (INIS)

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

2000-01-01

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

Usage of burnable poison on research reactors

International Nuclear Information System (INIS)

Villarino, Eduardo Anibal

2002-01-01

The fuel assemblies with burnable poison are widely used on power reactors, but there are not commonly used on research reactors. This paper shows a neutronic analysis of the advantages and disadvantages of the burnable poison usage on research reactors. This paper analyses both burnable poison design used on research reactors: Boron on the lateral wall and Cadmium wires. Both designs include a parametric study on the design parameters like the amount and geometry of the burnable poison. This paper presents the design flexibility using burnable poisons, it does not find an optimal or final design, which it will strongly depend on the core characteristics and fuel management strategy. (author)

Meltdown reactor core cooling facility

International Nuclear Information System (INIS)

Matsuoka, Tsuyoshi.

1992-01-01

The meltdown reactor core cooling facility comprises a meltdown reactor core cooling tank, a cooling water storage tank situates at a position higher than the meltdown reactor core cooling tank, an upper pipeline connecting the upper portions of the both of the tanks and a lower pipeline connecting the lower portions of them. Upon occurrence of reactor core meltdown, a high temperature meltdown reactor core is dropped on the cooling tank to partially melt the tank and form a hole, from which cooling water is flown out. Since the water source of the cooling water is the cooling water storage tank, a great amount of cooling water is further dropped and supplied and the reactor core is submerged and cooled by natural convection for a long period of time. Further, when the lump of the meltdown reactor core is small and the perforated hole of the meltdown reactor cooling tank is small, cooling water is boiled by the high temperature lump intruding into the meltdown reactor core cooling tank and blown out from the upper pipeline to the cooling water storage tank to supply cooling water from the lower pipeline to the meltdown reactor core cooling tank. Since it is constituted only with simple static facilities, the facility can be simplified to attain improvement of reliability. (N.H.)

Fuel designs for VVER reactors

International Nuclear Information System (INIS)

Simonov, K.V.; Carbon, P.; Silberstein, A.

1995-01-01

That progresses in efficiency and safety through progresses in technology and better prediction with fully benchmarked upgraded computer codes is a common goal for on the one hand the original designer of the VVER reactors and their respective fuels and on the other hand for EVF a western company resulting from a combined force with highly diversified and complementary talents in reactor and fuel design and manufacturing. It can be expected that this new challenge and dialogue between the two Russian and European industrial ventures will be mutually beneficial and yield innovative and high quality products and as a consequence strong return will be produced for the best interest of utilities operating VVER reactors. (orig./HP)

Engineering safety features for high power experimental reactors

International Nuclear Information System (INIS)

Doval, A.; Villarino, E.; Vertullo, A.

2000-01-01

In the present analysis we will focus our attention in the way engineering safety features are designed in order to prevent fuel damage in case of abnormal or accidental situations. To prevent fuel damage two main facts must be considered, the shutdown of the reactor and the adequate core cooling capacity, it means that both, neutronic and thermohydraulic aspects must be analysed. Some neutronic safety features are common to all power ranges like negative feedback reactivity coefficients and the required number of control rods containing the proper absorber material to shutdown the reactor. From the thermohydraulic point of view common features are siphon-breaker devices and flap valves for those powers requiring cooling in the forced convection regime. For the high power reactor group, the engineering safety features specially designed for a generic reactor of 20 MW, will be presented here. From the neutronic point of view besides the common features, and to comply with our National Regulatory Authority, a Second Shutdown System was designed as a redundant shutdown system in case the control plates fail. Concerning thermohydraulic aspects besides the pump flywheels and the flap valves providing the natural convection loop, a metallic Chimney and a Chimney Water Injection System were supplied. (author)

Steam up over reactor policy

International Nuclear Information System (INIS)

Kovan, D.

1976-01-01

Britain is once more assessing its nuclear power programme in the light of recent forecasts that there is unlikely to be any growth in the demand for electricity for many years to come. This means that the extra costs of launching a commercially unproven reactor, the Steam Generating Heavy Water Reactor (SGHWR), will be an even greater burden than previously expected, because they would be spread over fewer reactors. Sir John Hill's reported assessment concludes that the present strategy would be the most expensive way of developing Britain's nuclear power programme; and under the circumstances, may not be the best option. The SGHWR programme will certainly be more expensive than either relaunching a programme of advanced gas-cooled reactors (AGRs), or building American designed pressurised water reactors (PWRs). Recent developments of the AGR and PWR's and their advantages in the present position are outlined. (U.K.)

Monitoring and Control Research Using a University Reactor and SBWR Test-Loop

International Nuclear Information System (INIS)

Edwards, Robert M.

2003-01-01

The existing hybrid simulation capability of the Penn State Breazeale nuclear reactor was expanded to conduct research for monitoring, operations and control. Hybrid simulation in this context refers to the use of the physical time response of the research reactor as an input signal to a real-time simulation of power-reactor thermal-hydraulics which in-turn provides a feedback signal to the reactor through positioning of an experimental changeable reactivity device. An ECRD is an aluminum tube containing an absorber material that is positioned in the central themble of the reactor kinetics were used to expand the hybrid reactor simulation (HRS) capability to include out-of-phase stability characteristics observed in operating BWRs

Reactor power control method and device

International Nuclear Information System (INIS)

Fushimi, Atsushi; Ishii, Yoshihiko; Miyamoto, Yoshiyuki; Ishii, Kazuhiko; Kiyoharu, Norihiko; Aizawa, Yuko.

1997-01-01

The present invention provides a method and a device suitable to rise the temperature and increase the pressure of the reactor to an aimed pressure in accordance with an aimed value for a reactor water temperature changing rate in the course of rising temperature and increasing pressure of the reactor upon start up of a BWR type power plant. Namely, neutron fluxes in the reactor and the temperature of reactor water are detected respectively. The maximum value among the detected values for the neutron fluxes is detected. The reactor water temperature changing rate is calculated based on the detected values of the reactor water temperature, from which the maximum value of the reactor water temperature changing rate is detected. An aimed value for the neutron flux is calculated in accordance with both detected maximum values and the aimed value of the reactor water temperature changing rate. The position of control rods is adjusted in accordance with the aimed value for the calculated neutron flux. Then, an aimed value for the neutron flux for realizing the aimed value for the reactor water temperature changing rate can be obtained accurately with no influence of the sensitivity of the detected values of the neutron fluxes and the time delay of the reactor water temperature changing rate. (I.S.)

Nuclear reactor monitoring system

International Nuclear Information System (INIS)

Drummond, C.N.; Bybee, R.T.; Mason, F.L.; Worsham, H.J.

1976-01-01

The invention pertains to an improved monitoring system for the neutron flux in a nuclear reactor. It is proposed to combine neutron flux detectors, a thermoelement, and a background radiation detector in one measuring unit. The spatial arrangement of these elements is fixed with great exactness; they are enclosed by an elastic cover and are brought into position in the reactor with the aid of a bent tube. The arrangement has a low failure rate and is easy to maintain. (HP) [de

Chemical surveillance of commercial fast breeder reactors

International Nuclear Information System (INIS)

Stamm, H.H.; Stade, K.Ch.

1988-01-01

After BN-600 (USSR) and SUPERPHENIX (France) were started succesfully, the international development of LMFBRs is standing at the doorstep of commercial use. For commercial use of LMFBRs cost reductions for construction and operation are highly desirable and necessary. Several nations developing breeder reactors have joined in a common effort in order to reach this aim by standardization and harmonization. On the base of more than 20 years of operation experience of experimental reactors (EBR-II, FFTF, RAPSODIE, DFR, BR-5/BR-10, BOR-60, JOYO, KNK-II) and demonstration plants (PHENIX, PFR, BN-350), possibilities for standardization in chemical surveillance of commercial breeder reactors without any loss of availability, reliability and reactor safety will be discussed in the following chapters. Loop-type reactors will be considered as well as pool-type reactors, although all commercial plants under consideration so far (SUPERPHENIX II, BN-800, BN-1600, CFBR, SNR-2, EFR) include pool-type reactors only. Table 1 gives a comparison of the Na inventories of test reactors, prototype plants and commercial LMFBRs

Nuclear reactor machine refuelling system

International Nuclear Information System (INIS)

Cashen, W.S.; Erwin, D.

1977-01-01

Part of an on-line fuelling machine for a CANDU pressure-tube reactor is described. The present invention provides a refuelling machine wherein the fuelling components, including the fuel carrier and the closure adapter, are positively positioned and retained within the machine magazine or positively secured to the machine charge tube head, and cannot be accidentally disengaged as in former practice. The positive positioning devices include an arcuate keeper plate. Simplified hooked fingers are used. (NDH)

Reactor container facility

International Nuclear Information System (INIS)

Saito, Takashi; Nagasaka, Hideo.

1990-01-01

A dry-well pool for spontaneously circulating stored pool water and a suppression pool for flooding a pressure vessel by feeding water, when required, to a flooding gap by means of spontaneous falling upto the flooding position, thereby flooding the pressure vessel are contained at the inside of a reactor container. Thus, when loss of coolant accidents such as caused by main pipe rupture accidents should happen, pool water in the suppression pool is supplied to the flooding gap by spontaneously falling. Further, if the flooding water uprises exceeding a predetermined level, the flooding gap is in communication with the dry-well pool at the upper and the lower portions respectively. Accordingly, flooding water at high temperature heated by the after-heat of the reactor core is returned again into the flooding gap to cool the reactor core repeatedly. (T.M.)

Experience with reactor assembly of FBTR

International Nuclear Information System (INIS)

Srinivasan, G.; Ravishankar, K.; Babu, A.; Varadarajan, S.; Arumugam, P.; Sekhar, P.

2006-01-01

Reactor Assembly, also called Block Pile, is the heart of FBTR and houses the core, top and lateral shields, control rod drive mechanisms (CRDM), sodium inlet pipe and outlet pipes etc. Two major problems which arose during commissioning were reactor vessel tilt due to convection in cover gas space and failure of inflatable seals. The reactor vessel tilt was solved by Helium injection. Reactor was operated without pressurising the inflatable seals till 2005, when the seals were replaced. Other major problems in the course of twenty years of reactor operation were failure of three CRDM lower parts, Core Cover plate which houses the core thermocouples getting stuck in the fuel handling position, water leaks from the Biological Shield Cooling (BSC) coils around the reactor, failure of core wires in the trailing cables during fuel handling etc. This paper addresses the major problems faced and modifications carried out. (author)

Energy-positive sewage sludge pre-treatment with a novel ultrasonic flatbed reactor at low energy input.

Science.gov (United States)

Lippert, Thomas; Bandelin, Jochen; Musch, Alexandra; Drewes, Jörg E; Koch, Konrad

2018-05-20

The performance of a novel ultrasonic flatbed reactor for sewage sludge pre-treatment was assessed for three different waste activated sludges. The study systematically investigated the impact of specific energy input (200 - 3,000 kJ/kg TS ) on the degree of disintegration (DD COD , i.e. ratio between ultrasonically and maximum chemically solubilized COD) and methane production enhancement. Relationship between DD COD and energy input was linear, for all sludges tested. Methane yields were significantly increased for both low (200 kJ/kg TS ) and high (2,000 - 3,000 kJ/kg TS ) energy inputs, while intermediate inputs (400 - 1,000 kJ/kg TS ) showed no significant improvement. High inputs additionally accelerated reaction kinetics, but were limited to similar gains as low inputs (max. 12%), despite the considerably higher DD COD values. Energy balance was only positive for 200 kJ/kg TS -treatments, with a maximum energy recovery of 122%. Results suggest that floc deagglomeration rather than cell lysis (DD COD =1% - 5% at 200 kJ/kg TS ) is the key principle of energy-positive sludge sonication. Copyright © 2018 Elsevier Ltd. All rights reserved.

Baffle-former arrangement for nuclear reactor vessel internals

International Nuclear Information System (INIS)

Machado, O.J.; Berringer, R.T.

1978-01-01

A baffle-former arrangement for the reactor vessel internals of a nuclear reactor is described. The arrangement includes positioning of formers at the same elevations as the fuel assembly grids, and positioning flow holes in the baffle plates directly beneath selected former grid elevations. The arrangement reduces detrimental cross flows, maintains proper core barrel and baffle temperatures, and alleviates the potential of overpressurization within the baffle-former assembly under assumed major accident conditions

Nuclear reactor

International Nuclear Information System (INIS)

Schabert, H.P.; Weber, R.; Bauer, A.

1975-01-01

The refuelling of a PWR power reactor of about 1,200 MWe is performed by a transport pipe in the containment leading from an external to an internal fuel pit. A wagon to transport the fuel elements can go from a vertical loading position to an also vertical deloading position in the inner fuel pit via guide rollers. The necessary horizontal movement is effected by means of a cable line through the transport pipe which is inclined at least 10 0 . Gravity thus helps in the movement to the deloading position. The cable line with winch is fastened outside the containment. Swivelling devices tip the wagon from the horizontal to the vertical position or vice versa. Loading and deloading are done laterally. (TK/LH) [de

Method of operating FBR type reactors

International Nuclear Information System (INIS)

Arie, Kazuo.

1984-01-01

Purpose: To secure the controlling performance and the safety of FBR type reactors by decreasing the amount of deformation due to the difference in the heat expansion of a control rod guide tube. Method: The reactor is operated while disposing reactor core fuel assemblies of a same power at point-to-point symmetrical positions relative to the axial center for the control rod assembly. This can eliminate the temperature difference between opposing surfaces of the control rod guide tube and eliminate the difference in the thermal expansion. (Yoshino, Y.)

Strategies of development of reactor types

International Nuclear Information System (INIS)

Bacher, P.

2004-01-01

The development of nuclear energy in the coming decades will depend on the goals followed, on the available technologies and on the strategies implemented in the world in agreement with public acceptation. This article is limited to the technical aspects of the strategies of development of reactor types: 1 - objectives; 2 - common constraints to all reactor types: safety and terrorism risks, wastes, non-proliferation, economics; 3 - different reactor types: general considerations, proven technologies (PWR, BWR, Candu), non-proven technologies but having an important experience, technologies at the design stage; 4 - energy systems and 'Generation IV forum': systems based on thermal neutron reactors and low enrichment, systems for the valorization of 238 U, systems for Pu burning, systems allowing the destruction of minor actinides, thorium-based systems, the Gen IV international forum; 5 - conclusion. (J.S.)

A study of common-mode failures

International Nuclear Information System (INIS)

Edwards, G.T.; Watson, I.A.

1979-07-01

The purpose of the report is to investigate problems of the identification of the common failure mode (CFM) the reliability models used and the data required for their solution, particularly with regard to automatic protection systems for nuclear reactors. The available literature which was surveyed during the study is quoted and used as a basis for the main work of the study. The type of redundancy system under consideration is initially described and the types of CFM to which these systems are prone are identified before a general definition of the term 'common mode failure' is proposed. The definition and proposed classification system for CMF are based on the common cause of failure, so identifying the primary events. Defences against CFM are included and proposals for an overall strategy and detailed recommendations for design and operation are made. Common mode failures in US nuclear reactor systems, aircraft systems, and other sources including chemical plant systems are surveyed. The data indicates the importance of the human error problem in the causes of CMF in design, maintenance and operation. From a study of the collected data a redundancy sub-system model for CMF is developed which identifies three main categories of failure, non-recurrent engineering design errors, maintenance and test errors, and random interest events. The model proposed allows for the improvement in sub-system reliability where appropriate defences are applied. (author)

Guidelines for nuclear reactor equipments safety-analysis

International Nuclear Information System (INIS)

1978-01-01

The safety analysis in approving the applications for nuclear reactor constructions (or alterations) is performed by the Committee on Examination of Reactor Safety in accordance with various guidelines prescribed by the Atomic Energy Commission. In addition, the above Committee set forth its own regulations for the safety analysis on common problems among various types of nuclear reactors. This book has collected and edited those guidelines and regulations. It has two parts: Part I includes the guidelines issued to date by the Atomic Energy Commission: and Part II - regulations of the Committee. Part I has collected 8 categories of guidelines which relate to following matters: nuclear reactor sites analysis guidelines and standards for their applications; standard exposure dose of plutonium; nuclear ship operation guidelines; safety design analysis guidelines for light-water type, electricity generating nuclear reactor equipments; safety evaluation guidelines for emergency reactor core cooling system of light-water type power reactors; guidelines for exposure dose target values around light-water type electricity generating nuclear reactor equipments, and guidelines for evaluation of above target values; and meteorological guidelines for the safety analysis of electricity generating nuclear reactor equipments. Part II includes regulations of the Committee concerning - the fuel assembly used in boiling-water type and in pressurized-water type reactors; techniques of reactor core heat designs, etc. in boiling-water reactors; and others

Defense against common-mode failures in protection system design

International Nuclear Information System (INIS)

Wyman, R.H.; Johnson, G.L.

1998-01-01

The introduction of digital instrumentation and control into reactor safety systems creates a heightened concern about common-mode failure. This paper discusses the concern and methods of cope with the concern. Common-mode failures have been a 'fact-of-life' in existing systems. The informal introduction of defense-in-depth and diversity (D-in-D and D) - coupled with the fact that hardware common-mode failures are often distributed in time - has allowed systems to deal with past common-mode failures. However, identical software operating in identical redundant systems presents the potential for simultaneous failure. Consequently, the use of digital systems raises the concern about common-mode failure to a new level. A more methodical approach to mitigating common-mode failure is needed to address these concerns. Purposeful introduction of D-in-D and D has been used as a defense against common-mode failure in reactor protection systems. At least two diverse systems are provided to mitigate any potential initiating event. Additionally, diverse displays and controls are provided to allow the operator to monitor plant status and manually initiate engineered safety features. A special form of common-mode failure analysis called 'defense-in-depth and diversity analysis' has been developed to identify possible common-mode failure vulnerabilities in digital systems. An overview of this analysis technique is provided. (author)

WWER-440 type reactor core

International Nuclear Information System (INIS)

Mizov, J.; Svec, P.; Rajci, T.

1987-01-01

Assemblies with patly spent fuel of enrichment within 5 and 36 MWd/kg U or lower than the maximum enrichment of freshly charged fuel are placed in at least one of the peripheral positions of each hexagonal sector of the WWER-440 reactor type core. This increases fuel availability and reduces the integral neutron dose to the reactor vessel. The duration is extended of the reactor campaign and/or the mean fuel enrichment necessary for the required duration of the period between refuellings is reduced. Thus, fuel costs are reduced by 1 up to 3%. The results obtained in the experiment are tabulated. (J.B.). 1 fig., 3 tabs

In-service inspection of pool type research reactors

International Nuclear Information System (INIS)

Rajamani, K.

2002-01-01

In the case of Apsara Reactor, it has been proposed to carry out major modifications in the near future. It is planned to modify the core suitably with a heavy water reflector tank to demonstrate the Multiple Purpose Research Reactor concept. The core structure will be a stationary one and will be located at the 'B' position of the pool. The modified reactor will be operated at 1 MW power level. Suitable methodologies are evolved for carrying out a planned ISI for this modified reactor

Report on the Survey of the Design Review of New Reactor Applications. Volume 3: Reactor

International Nuclear Information System (INIS)

Downey, Steven; Monninger, John; Nevalainen, Janne; Lorin, Aurelie; ); Webster, Philip; Joyer, Philippe; Kawamura, Tomonori; Lankin, Mikhail; Kubanyi, Jozef; Haluska, Ladislav; Persic, Andreja; Reierson, Craig; Kang, Kyungmin; Kim, Walter

2016-01-01

At the tenth meeting of the CNRA Working Group on the Regulation of New Reactors (WGRNR) in March 2013, the Working Group agreed to present the responses to the Second Phase, or Design Phase, of the Licensing Process Survey as a multi-volume text. As such, each report will focus on one of the eleven general technical categories covered in the survey. The general technical categories were selected to conform to the topics covered in the International Atomic Energy Agency (IAEA) Safety Guide GS-G-4.1. This document, which is the third report on the results of the Design Phase Survey, focuses on the Reactor. The Reactor category includes the following technical topics: fuel system design, reactor internals and core support, nuclear design and core nuclear performance, thermal and hydraulic design, reactor materials, and functional design of reactivity control system. For each technical topic, the member countries described the information provided by the applicant, the scope and level of detail of the technical review, the technical basis for granting regulatory authorisation, the skill sets required and the level of effort needed to perform the review. Based on a comparison of the information provided by the member countries in response to the survey, the following observations were made: - Although the description of the information provided by the applicant differs in scope and level of detail among the member countries that provided responses, there are similarities in the information that is required. - All of the technical topics covered in the survey are reviewed in some manner by all of the regulatory authorities that provided responses. - Design review strategies most commonly used to confirm that the regulatory requirements have been met include document review and independent verification of calculations, computer codes, or models used to describe the design and performance of the core and the fuel. - It is common to consider operating experience and

The assessment of voce coefficient for WWR-c reactor

International Nuclear Information System (INIS)

Kochnov, O.Yu.; Rybkin, N.I.

2006-01-01

The air cavity effect in WWR-ts reactor core on the total reactivity is analyzed. The experimental data of void coefficient depending on the air cavity position inside the reactor core are obtained [ru

Material test reactor fuel research at the BR2 reactor

Energy Technology Data Exchange (ETDEWEB)

Dyck, Steven Van; Koonen, Edgar; Berghe, Sven van den [Institute for Nuclear Materials Science, SCK-CEN, Boeretang, Mol (Belgium)

2012-03-15

The construction of new, high performance material test reactor or the conversion of such reactors' core from high enriched uranium (HEU) to low enriched uranium (LEU) based fuel requires several fuel qualification steps. For the conversion of high performance reactors, high density dispersion or monolithic fuel types are being developed. The Uranium-Molybdenum fuel system has been selected as reference system for the qualification of LEU fuels. For reactors with lower performance characteristics, or as medium enriched fuel for high performance reactors, uranium silicide dispersion fuel is applied. However, on the longer term, the U-Mo based fuel types may offer a more efficient fuel alternative and-or an easier back-end solution with respect to the silicide based fuels. At the BR2 reactor of the Belgian nuclear research center, SCK-CEN in Mol, several types of fuel testing opportunities are present to contribute to such qualification process. A generic validation test for a selected fuel system is the irradiation of flat plates with representative dimensions for a fuel element. By flexible positioning and core loading, bounding irradiation conditions for fuel elements can be performed in a standard device in the BR2. For fuel element designs with curved plates, the element fabrication method compatibility of the fuel type can be addressed by incorporating a set of prototype fuel plates in a mixed driver fuel element of the BR2 reactor. These generic types of tests are performed directly in the primary coolant flow conditions of the BR2 reactor. The experiment control and interpretation is supported by detailed neutronic and thermal-hydraulic modeling of the experiments. Finally, the BR2 reactor offers the flexibility for irradiation of full size prototype fuel elements, as 200mm diameter irradiation channels are available. These channels allow the accommodation of various types of prototype fuel elements, eventually using a dedicated cooling loop to provide the

Evaluation of the positional difference between two common geocoding methods

Directory of Open Access Journals (Sweden)

Dustin T. Duncan

2011-05-01

Full Text Available Geocoding, the process of matching addresses to geographic coordinates, is a necessary first step when using geographical information systems (GIS technology. However, different geocoding methodologies can result in different geographic coordinates. The objective of this study was to compare the positional (i.e. longitude/latitude difference between two common geocoding methods, i.e. ArcGIS (Environmental System Research Institute, Redlands, CA, USA and Batchgeo (freely available online at http://www.batchgeo.com. Address data came from the YMCA-Harvard After School Food and Fitness Project, an obesity prevention intervention involving children aged 5-11 years and their families participating in YMCAadministered, after-school programmes located in four geographically diverse metropolitan areas in the USA. Our analyses include baseline addresses (n = 748 collected from the parents of the children in the after school sites. Addresses were first geocoded to the street level and assigned longitude and latitude coordinates with ArcGIS, version 9.3, then the same addresses were geocoded with Batchgeo. For this analysis, the ArcGIS minimum match score was 80. The resulting geocodes were projected into state plane coordinates, and the difference in longitude and latitude coordinates were calculated in meters between the two methods for all data points in each of the four metropolitan areas. We also quantified the descriptions of the geocoding accuracy provided by Batchgeo with the match scores from ArcGIS. We found a 94% match rate (n = 705, 2% (n = 18 were tied and 3% (n = 25 were unmatched using ArcGIS. Forty-eight addresses (6.4% were not matched in ArcGIS with a match score ≥80 (therefore only 700 addresses were included in our positional difference analysis. Six hundred thirteen (87.6% of these addresses had a match score of 100. Batchgeo yielded a 100% match rate for the addresses that ArcGIS geocoded. The median for longitude and latitude

Strengthening IAEA Safeguards for Research Reactors

Energy Technology Data Exchange (ETDEWEB)

Reid, Bruce D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Anzelon, George A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Budlong-Sylvester, Kory [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-09-01

During their December 10-11, 2013, workshop in Grenoble France, which focused on the history and future of safeguarding research reactors, the United States, France and the United Kingdom (UK) agreed to conduct a joint study exploring ways to strengthen the IAEA’s safeguards approach for declared research reactors. This decision was prompted by concerns about: 1) historical cases of non-compliance involving misuse (including the use of non-nuclear materials for production of neutron generators for weapons) and diversion that were discovered, in many cases, long after the violations took place and as part of broader pattern of undeclared activities in half a dozen countries; 2) the fact that, under the Safeguards Criteria, the IAEA inspects some reactors (e.g., those with power levels under 25 MWt) less than once per year; 3) the long-standing precedent of States using heavy water research reactors (HWRR) to produce plutonium for weapons programs; 4) the use of HEU fuel in some research reactors; and 5) various technical characteristics common to some types of research reactors that could provide an opportunity for potential proliferators to misuse the facility or divert material with low probability of detection by the IAEA. In some research reactors it is difficult to detect diversion or undeclared irradiation. In addition, infrastructure associated with research reactors could pose a safeguards challenge. To strengthen the effectiveness of safeguards at the State level, this paper advocates that the IAEA consider ways to focus additional attention and broaden its safeguards toolbox for research reactors. This increase in focus on the research reactors could begin with the recognition that the research reactor (of any size) could be a common path element on a large number of technically plausible pathways that must be considered when performing acquisition pathway analysis (APA) for developing a State Level Approach (SLA) and Annual Implementation Plan (AIP). To

Nuclear reactor capable of electric power generation during in-service inspection

International Nuclear Information System (INIS)

Nakamura, Shinsuke; Nogami, Hitoshi.

1992-01-01

The nuclear power plant according to the present invention can generate electric power even in a period when one of a pair of reactors is put to in-service inspection. That is, the nuclear power plant of the present invention comprises a system constitution of two nuclear reactors each of 50% thermal power and one turbine power generator of 100% electric power. Further, facilities of various systems relevant to the two reactors each of 50% thermal power, as a pair, are used in common as much as possible in order to reduce the cost for construction and maintenance/ inspection. Further, a reactor building and a turbine building disposed in adjacent with each for paired two reactors each of 50% thermal power are arranged vertically. This arrangement can facilitate the common use of the facilities for various systems and equipments to attain branching and joining of fluids in reactor feed water systems and main steam system pipelines easily with low pressure loss and low impact shocks. The facility utilization factor of such reactors is remarkably improved by doubling the period of continuous power generation. As a result, economic property is remarkably improved. (I.S.)

Method of dismantling a nuclear reactor

International Nuclear Information System (INIS)

Shirai, Masato; Hashimoto, Osamu.

1984-01-01

Purpose: To enable rapid and simple positioning for a plasma arc torch disposed to the inside of a nuclear reactor main body. Method: After removing the upper semi-spherical portion, fuel portion and control rod portion of a nuclear reactor, a rotary type girder is placed on the upper edge of a cylindrical portion remained after the removal of the upper semi-spherical portion. Then, the upper portion of a supporting rod provided with a swing arm having a plasma arc torch at the top end is situated at the center of the reactor main body. Then, the top end of the support rod is inserted to fix in the housing of control rod drives. Then, the swing arm is actuated to situate the plasma arc torch to a desired position to be cut, whereafter cutting is initiated while rotating the rotary type girder. Thus, plasma arc torch is moved horizontally along an arcuate trace, whereby pipeways, accessories or the likes disposed to the inside of the main body are at first cut and then the cylindrical portion constituting the main body is cut to dismantle the reactor. (Moriyama, K.)

Temperature measuring element in nuclear reactors

International Nuclear Information System (INIS)

Wada, Takashi.

1987-01-01

Purpose: To easily measure the partial maximum temperature at a portion within the nuclear reactor where the connection with the external portion is difficult. Constitution: Sodium, potassium or the alloy thereof with high heat expansion coefficient is packed into an elastic vessel having elastic walls contained in a capsule. A piercing member formed into an acute triangle is attached to one end in the direction of expansion and contraction of the elastic container. The two sides of the triangle form an acute knife edge. A diaphragm is disposed within a capsule at a position opposed to the sharpened direction of the piercing member. Such a capsule is placed in a predetermined position of the nuclear reactor. The elastic vessel causes thermal expansion displacement depending on the temperature at a certain position, by which the top end of the pierce member penetrates through the diaphragm. A pierced scar of a penetration length depending on the temperature is resulted to the diaphragm. The length of the piercing damage is electroscopically observed and compared with the calibration curve to recognize the maximum temperature in the predetermined portion of the nuclear reactor. (Kamimura, M.)

Seismic attenuation system for a nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Liszkai, Tamas; Cadell, Seth

2018-01-30

A system for attenuating seismic forces includes a reactor pressure vessel containing nuclear fuel and a containment vessel that houses the reactor pressure vessel. Both the reactor pressure vessel and the containment vessel include a bottom head. Additionally, the system includes a base support to contact a support surface on which the containment vessel is positioned in a substantially vertical orientation. An attenuation device is located between the bottom head of the reactor pressure vessel and the bottom head of the containment vessel. Seismic forces that travel from the base support to the reactor pressure vessel via the containment vessel are attenuated by the attenuation device in a direction that is substantially lateral to the vertical orientation of the containment vessel.

Temperature and Doppler coefficients of various space nuclear reactors

International Nuclear Information System (INIS)

Mughabghab, S.F.; Ludewig, H. Schmidt, E.

1993-01-01

Temperature and Doppler feedback effects for a Particle Bed Reactor (PBR) designed to operate as a propulsion reactor are investigated. Several moderator types and compositions fuel enrichments and reactor sizes are considered in this study. From this study it could be concluded that a PBR can be configured which has a negative prompt feedback, zero coolant worth, and a small positive to zero moderator worth. This reactor would put the lowest demands on the control system

Temperature and Doppler Coefficients of Various Space Nuclear Reactors

Science.gov (United States)

Mughabghab, Said F.; Ludewig, Hans; Schmidt, Eldon

1994-07-01

Temperature and Doppler feedback effects for a Particle Bed Reactor (PBR) designed to operate as a propulsion reactor are investigated. Several moderator types and compositions fuel enrichments and reactor sizes are considered in this study. From this study it could be concluded that a PBR can be configured which has a negative prompt feedback, zero coolant worth, and a small positive to zero moderator worth. This reactor would put the lowest demands on the control system.

PWR type reactor plant

International Nuclear Information System (INIS)

Matsuoka, Tsuyoshi.

1993-01-01

A water chamber of a horizontal U-shaped pipe type steam generator is partitioned to an upper high temperature water chamber portion and a lower low temperature water chamber portion. An exit nozzle of a reactor container containing a reactor core therein is connected to a suction port of a coolant pump by way of first high temperature pipelines. The exit port of the coolant pump is connected to the high temperature water chamber portion of the steam generator by way of second high temperature pipelines. The low temperature water chamber portion of the steam generator is connected to an inlet nozzle of the reactor container by way of the low temperature pipelines. The low temperature water chamber portion of the steam generator is positioned lower than the high temperature water chamber portion, but upper than the reactor core. Accordingly, all of the steam generator for a primary coolant system, coolant pumps as well as high temperature pipelines and low temperature pipelines connecting them are disposed above the reactor core. With such a constitution, there is no worry of interrupting core cooling even upon occurrence of an accident, to improve plant safety. (I.N.)

Irradiated topaz in the reactor

International Nuclear Information System (INIS)

Helal, A.I.; Zahran, N.F.; Gomaac, M.A.M.; Salama, S.

2007-01-01

Gem stones are those stones which have beauty that can be based on its color, transparency, brilliance and the crystalline perfection . Topaz is used mainly as gemstones, It is the most common irradiated gem on the market. High energy such as neutrons, have enough energy to produce color centers . Irradiation is most often carried out in nuclear reactors (high-energy neutrons). Irradiation of topaz in the Egyptian research reactor (ETRR-2) by neutrons changes its cloudy white color to a reddish pink which could be changed to blue by heating

Nuclear reactor with several cores

International Nuclear Information System (INIS)

Swars, H.

1977-01-01

Several sodium-cooled cores in separate vessels with removable closures are placed in a common reactor tank. Each individual vessel is protected against the consequences of an accident in the relevant core. Maintenance devices and inlet and outlet pipes for the coolant are also arranged within the reactor tank. The individual vessels are all enclosed by coolant in a way that in case of emergency cooling or refuelling each core can be continued to be cooled by means of the coolant loops of the other cores. (HP) [de

Smiling faces and cash bonuses: Exploring common affective coding across positive and negative emotional and motivational stimuli using fMRI.

Science.gov (United States)

Park, Haeme R P; Kostandyan, Mariam; Boehler, C Nico; Krebs, Ruth M

2018-06-01

Although it is clear that emotional and motivational manipulations yield a strong influence on cognition and behaviour, these domains have mostly been investigated in independent research lines. Therefore, it remains poorly understood how far these affective manipulations overlap in terms of their underlying neural activations, especially in light of previous findings that suggest a shared valence mechanism across multiple affective processing domains (e.g., monetary incentives, primary rewards, emotional events). This is particularly interesting considering the commonality between emotional and motivational constructs in terms of their basic affective nature (positive vs. negative), but dissociations in terms of instrumentality, in that only reward-related stimuli are typically associated with performance-contingent outcomes. Here, we aimed to examine potential common neural processes triggered by emotional and motivational stimuli in matched tasks within participants using functional magnetic resonance imaging (fMRI). Across tasks, we found shared valence effects in the ventromedial prefrontal cortex and left inferior frontal gyrus (part of dorsolateral prefrontal cortex), with increased activity for positive and negative stimuli, respectively. Despite this commonality, emotion and reward tasks featured differential behavioural patterns in that negative valence effects (performance costs) were exclusive to emotional stimuli, while positive valence effects (performance benefits) were only observed for reward-related stimuli. Overall, our data suggest a common affective coding mechanism across different task domains and support the idea that monetary incentives entail signed basic valence signals, above and beyond the instruction to perform both gain and loss trials as accurately as possible to maximise the outcome.

Concept and basic performance of an in-pile experimental reactor for fast breeder reactors using fast driver core

International Nuclear Information System (INIS)

Obara, Toru; Sekimoto, Hiroshi

1997-01-01

The possibility of an in-pile experimental reactor for fast breeder reactors using a fast driver core is investigated. The driver core is composed of a particle bed with diluted fuel. The results of various basic analyses show that this reactor could perform as follows: (1) power peaking at the outer boundary of test core does not take place for large test core; (2) the radial power distribution in test fuel pin is expected to be the same as a real reactor; (3) the experiments with short half width pulse is possible; (4) for the ordinary MOX core, enough heating-up is possible for core damage experiments; (5) the positive effects after power burst can be seen directly. These are difficult for conventional thermal in-pile experimental reactors in large power excursion experiments. They are very attractive advantages in the in-pile experiments for fast breeder reactors. (author)

Can commonly used profile planes be used to evaluate changes in lower lip position?

Science.gov (United States)

Buschang, Peter H; Fretty, Kimberly; Campbell, Phillip M

2011-07-01

To determine the validity of five profile planes commonly used to describe the horizontal changes of the lower lip during orthodontic treatment. Pretreatment and posttreatment cephalograms of 79 patients (12.4 ± 2.8 years of age) were evaluated. Lower lip (labrale inferiorus) changes over time were measured relative to the Rickett's E-line, Steiner's S1-line, Burstone's B-line, Sushner's S2-line, and Holdaway's H-line. As an independent measure of actual horizontal lip changes, the labrale inferiorus was measured relative to a stable reference plane registered on the sella and oriented on the SN-7°. The lower lip actually moved anteriorly 2.35 ± 3.35 mm during orthodontic treatment; the five profile planes indicated that the lower lip moved to a more retrusive, posterior position. The five profile planes also showed no statistically significant sex differences in terms of the treatment changes that occurred, while the actual lip changes showed that males exhibited significantly greater changes than females. Actual treatment changes showed that the lower lip moved to a more protrusive position with nonextraction than with extraction treatments, changes that were not evident based on the five profile lines. While lip changes based on the five profile planes demonstrated moderately high to high intercorrelations ranging from 0.81 to 0.97, they showed only weak correlations (r planes measured similar aspects of positional change, none of them closely reflected the actual lower lip changes that occurred. These planes should not be used to measure changes in lip position that occur during treatment.

Positive predictive value of cholescintigraphy in common bile duct obstruction

International Nuclear Information System (INIS)

Lecklitner, M.L.; Austin, A.R.; Benedetto, A.R.; Growcock, G.W.

1986-01-01

Technetium-99m DISIDA imaging was employed in 400 patients to differentiate obstruction of the common bile duct from medical and other surgical causes of hyperbilirubinemia. Sequential anterior images demonstrated variable degrees of liver uptake, yet there was no evidence of intrabiliary or extrabiliary radioactivity for at least 4 hr after injection in 25 patients. Twenty-three patients were surgically documented to have complete obstruction of the common bile duct. One patient had hepatitis, and another had sickle cell crisis without bile duct obstruction. The remaining patients had either partial or no obstruction of the common bile duct. We conclude that the presence of liver uptake without evident biliary excretion by 4 hr on cholescintigraphy is highly sensitive and predictive of total obstruction of the common bile duct

Micro processor based research reactor instrumentation and control system

International Nuclear Information System (INIS)

Hyde, W.K.

1987-01-01

The system consists of a Control System Computer (CSC) incorporated into a Reactor Control Console (RCC) and a Data Acquisition and Control Unit (DAC) adjacent to the reactor. The CSC has a high resolution color graphics CRT monitor which provides real-time graphic simulation of the reactor and a number of bar graphs displaying strategic parameters of the reactor system. In addition, abnormal or dangerous conditions are displayed. The CSC is equipped with two printers eliminating manual logging of reactor data. The reactor display and pulse mode display may also be printed. Historical data is saved in the system's large capacity memory and may be replayed and/or printed. Because of the CSC's inherent high speed math capability, raw reactor data will be quickly converted and displayed in real-time. Data can be presented in meaningful engineering units. The DAC provides a high speed data acquisition and control capability adjacent to the reactor. It continuously collects data from the reactor system, concentrates the data into a database and transmits it to the CSC when requested. Data transmission is over one of two data trunks to the CSC. The secondary trunk is used if the primary trunk fails. The data trunks drastically reduce the wiring requirements between the reactor and the Control Console. During steady-state operation of the reactor, operator commands to adjust the rod positions is transmitted from the CSC to the DAC which re-issues the commands to the drive mechanisms. In the automatic mode, the DAC will control the position of the rods via a PID algorithm. The system is independently monitored by two or more safety computers. Their function is to monitor the power level, the rate of change of power and fuel temperature of the reactor and to independently shut the reactor down in the event of a potentially dangerous (scram) condition. (author)

System assessment of helical reactors in comparison with tokamaks

International Nuclear Information System (INIS)

Yamazaki, K.; Imagawa, S.; Muroga, T.; Sagara, A.; Okamura, S.

2002-10-01

A comparative assessment of tokamak and helical reactors has been performed using equivalent physics/engineering model and common costing model. Higher-temperature plasma operation is required in tokamak reactors to increase bootstrap current fraction and to reduce current-drive (CD) power. In helical systems, lower-temperature operation is feasible and desirable to reduce helical ripple transport. The capital cost of helical reactor is rather high, however, the cost of electricity (COE) is almost same as that of tokamak reactor because of smaller re-circulation power (no CD power) and less-frequent blanket replacement (lower neutron wall loading). The standard LHD-type helical reactor with 5% beta value is economically equivalent to the standard tokamak with 3% beta. The COE of lower-aspect ratio helical reactor is on the same level of high-β N tokamak reactors. (author)

MDEP Common Position CP-EPRWG-04. Common position on EPR containment heat removal system in accident conditions

International Nuclear Information System (INIS)

2015-01-01

The importance of the integrity of the containment as a fundamental barrier to protect the people and environment against the effects of a nuclear accident is well established. In this regard, an essential objective is that the necessity for off-site counter-measures to reduce radiological consequences be limited or even eliminated. The design should provide engineering means to address those sequences which would otherwise lead to large or early releases, even in case of severe external hazards. The plant shall be designed so that it can be brought into a controlled and stable state and the containment function can be maintained, under accident conditions in which there is a significant amount of radioactive material in the containment, i.e. resulting from severe degradation of the reactor core. It is expected that due consideration to these requirements is to be given while tailoring long term loss of electrical power mitigation strategies. In order to reliably maintain the containment barrier, the regulators believe that: - safety features specifically designed for fulfilling safety functions required in core melt accidents shall be independent to the extent reasonably practicable from the Systems, Structures and Components (SSC) of the other levels of defense; - safety features specifically designed for fulfilling safety functions required in core melt accidents shall be safety classified and adequately qualified for the core melt accident environmental conditions for the time frame for which they are required to operate. In the light of the Fukushima Daiichi accident, the regulators believe that those safety features shall be designed with an adequate margin as compared to the levels of natural hazards considered for the site hazard evaluation; - the systems and components necessary for ensuring the containment function in a core melt accident shall have reliability commensurate with the function that they are required to fulfil. This may require redundancy of

Homogeneous SLOWPOKE reactors for replacing SLOWPOKE-2 research reactors and the production of radioisotopes

International Nuclear Information System (INIS)

Bonin, H.W.; Hilborn, J.W.; Carlin, G.E.; Gagnon, R.; Busatta, P.

2014-01-01

Inspired from the inherently safe SLOWPOKE-2 research reactor, the Homogeneous SLOWPOKE reactor was conceived with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors when they reach end-of-core life to continue their missions of neutron activation analysis and neutron radiography at universities, and to produce radioisotopes such as 99 Mo for medical applications. A homogeneous reactor core allows a much simpler extraction of radioisotopes (such as 99 Mo) for applications in industry and nuclear medicine. The 20 kW Homogeneous SLOWPOKE reactor was modelled using both the deterministic WIMS-AECL and the probabilistic MCNP 5 reactor simulation codes. The homogeneous fuel mixture was a dilute aqueous solution of Uranyl Sulfate (UO 2 SO 4 ) with 994.2 g of 235 U (enrichment at 20%) providing an excess reactivity at operating temperature (40 o C) of 3.8 mk for a molality determined as 1.46 mol kg -1 for a Zircaloy-2 reactor vessel. Because this reactor is intended to replace the core of SLOWPOKE-2 reactors, the Homogeneous SLOWPOKE reactor core had a height about twice its diameter. The reactor could be controlled by mechanical absorber rods in the beryllium reflector, chemical control in the core, or a combination of both. The safety of the Homogeneous SLOWPOKE reactor was analysed for both normal operation and transient conditions. Thermal-hydraulics calculations used COMSOL Multiphysics and the results showed that natural convection was sufficient to ensure adequate reactor cooling in all situations. The most severe transient simulated resulted from a 5.87 mk step positive reactivity insertion to the reactor in operation at critical and at steady state at 20 o C. Peak temperature and power were determined as 83 o C and 546 kW, respectively, reached 5.1 s after the reactivity insertion. However, the power fell rapidly to values below 20 kW some 35 s after the peak and remained below that value thereafter. Both the temperature and void coefficients are

Homogeneous SLOWPOKE reactors for replacing SLOWPOKE-2 research reactors and the production of radioisotopes

Energy Technology Data Exchange (ETDEWEB)

Bonin, H.W., E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada); Hilborn, J.W. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Carlin, G.E. [Ontario Power Generation, Toronto, Ontario (Canada); Gagnon, R.; Busatta, P. [Canadian Forces (Canada)

2014-07-01

Inspired from the inherently safe SLOWPOKE-2 research reactor, the Homogeneous SLOWPOKE reactor was conceived with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors when they reach end-of-core life to continue their missions of neutron activation analysis and neutron radiography at universities, and to produce radioisotopes such as {sup 99}Mo for medical applications. A homogeneous reactor core allows a much simpler extraction of radioisotopes (such as {sup 99}Mo) for applications in industry and nuclear medicine. The 20 kW Homogeneous SLOWPOKE reactor was modelled using both the deterministic WIMS-AECL and the probabilistic MCNP 5 reactor simulation codes. The homogeneous fuel mixture was a dilute aqueous solution of Uranyl Sulfate (UO{sub 2}SO{sub 4}) with 994.2 g of {sup 235}U (enrichment at 20%) providing an excess reactivity at operating temperature (40 {sup o}C) of 3.8 mk for a molality determined as 1.46 mol kg{sup -1} for a Zircaloy-2 reactor vessel. Because this reactor is intended to replace the core of SLOWPOKE-2 reactors, the Homogeneous SLOWPOKE reactor core had a height about twice its diameter. The reactor could be controlled by mechanical absorber rods in the beryllium reflector, chemical control in the core, or a combination of both. The safety of the Homogeneous SLOWPOKE reactor was analysed for both normal operation and transient conditions. Thermal-hydraulics calculations used COMSOL Multiphysics and the results showed that natural convection was sufficient to ensure adequate reactor cooling in all situations. The most severe transient simulated resulted from a 5.87 mk step positive reactivity insertion to the reactor in operation at critical and at steady state at 20 {sup o}C. Peak temperature and power were determined as 83 {sup o}C and 546 kW, respectively, reached 5.1 s after the reactivity insertion. However, the power fell rapidly to values below 20 kW some 35 s after the peak and remained below that value thereafter. Both the

Radiological walkover survey at the site of the world's first nuclear reactor using the NAVSTAR global positioning system

International Nuclear Information System (INIS)

Sims, J.; Barry, C.; Efird, C.

1994-01-01

A radiological walkover survey was conducted by Bechtel National, Inc. (BNI) in the spring of 1994 at Site A, the site of the world's first nuclear reactor pilot plant facility. The objective of this survey was to provide comprehensive information regarding surface radiation levels to support identification and documentation of unknown contaminated areas and selection of future sample locations, and provide information for health and safety coverage during field investigations. The technique used represented a new application integrating the radiation measuring instrumentation with the global positioning system (GPS) in real-time differential mode. The integrated mapping and radiation survey system (MARSS) developed by J.S. Bland Associates, Inc. (JSB) simultaneously recorded direct radiation, exposure rate, and position data. Site A is in the Palos Forest Preserve, about 20 miles southwest of downtown Chicago, Illinois. The topography of the site is relatively flat (usually less than 5 % grade), with a maximum grade of 10%. Prior to the survey, the site was staked on 100-ft. centers and roughly 75% of the site was cleared of dense vegetation including trees less than 6 in. in diameter. The survey covered approximately 2 acres each day, including background determinations, source checks, calibrations, and downloading of integrated position data and radiological measurements

Optimization geometries of a vortex gliding-arc reactor for partial oxidation of methane

International Nuclear Information System (INIS)

Guofeng, Xu; Xinwei, Ding

2012-01-01

The effects of the geometry of gliding-arc reactor – such as distance between the electrodes, outlet diameter, and inlet position – on the reactor characteristics (methane conversion, hydrogen yield, and energy efficiency) have not been fully investigated. In this paper, AC gliding-arc reactors including the vortex flow configuration are designed to produce hydrogen from the methane by partial oxidation. The influence of vortex flow configuration on the reactor characteristics is also studied by varying the inlet position. When the inlet of the gliding-arc reactor is positioned close to the outlet, reverse vortex flow reactor (RVFR), the maximum energy efficiency reaches 50% and the yields of hydrogen and carbon monoxide are 40% and 65%, respectively. As the distance between electrodes increases from 5 mm to 15 mm, both hydrogen yield and energy efficiency increase approximately 10% for the RVFR. The energy efficiency and hydrogen yield are highest when the ratio of the outlet diameter to the inner diameter is 0.5 for the RVFR. Experimental results indicate that the flow field in the plasma reactor has an important influence on the reactor performance. Furthermore, hydrogen production increases as the number of feed gas flows in contact with the plasma zone increases. -- Highlights: ► Gliding-arc reactors were designed to produce hydrogen for studying the characteristics of the vortex flow reactor. ► Hydrogen yield of reverse vortex flow reactor was 10% higher than that of forward vortex flow reactor. ► Maximum energy efficiency was 50% for reverse vortex flow reactor. ► If discharge power was supplied to the reactors, the reactor performance increased with increasing distance between electrodes. ► Optimum ratio of the outlet and inner diameter was 1/2.

Radiolytic generation of gases in reactors

International Nuclear Information System (INIS)

Ramshesh, V.; Venkateswarlu, K.S.

1988-01-01

Water or heavy water is used in different circuits in a reactor. Their most common use is as a moderator and/or as a coolant. Light water is used at other places such as in end shield, calandria vault etc., In the process they are exposed to intense ionizing radiation and undergo radiolytic degradation. The molecular produts of radiolysis are hydrogen, hydrogen peroxide and oxygen. As is commonly known if hydrogen is formed beyond a certain level, in the presence of oxygen it may lead to combustion or even explosion. Thus one should comprehend the basic principles of radiolysis and see whether the concentration of these gases under various conditions can be worked out. This report attempts to analyse in depth the radiolytic generation of gases in reactor systems. (author). 3 tabs

A review of the UK fast reactor programme

International Nuclear Information System (INIS)

Picker, C.; Ainsworth, K.F.

1998-01-01

The general position with regard to nuclear power and fast reactors in the UK during 1996 is described. The main UK Government-funded fast reactor research and development programme was concluded in 1993, to be replaced by a smaller programme which is mainly funded and managed by British Nuclear Fuels plc. The main focus of this programme sustains the UK participation in the European Fast Reactor (EFR) collaboration and the broader international links built-up over the previous decades. The status of fast reactor studies made in the UK in 1996 is outlined and, with respect to the Prototype Fast Reactor at Dounreay, a report of progress with the closure studies, fuel reprocessing and decommissioning activities is provided. (author)

Defense against common-mode failures in protection system design

International Nuclear Information System (INIS)

Wyman, R.H.; Johnson, G.L.

1997-01-01

The introduction of digital instrumentation and control into reactor safety systems creates a heightened concern about common-mode failure. This paper discusses the concern and methods to cope with the concern. Common-mode failures have been a ''fact-of-life'' in existing systems. The informal introduction of defense-in-depth and diversity (D-in-D ampersand D)-coupled with the fact that hardware common-mode failures are often distributed in time-has allowed systems to deal with past common-mode failures. However, identical software operating in identical redundant systems presents the potential for simultaneous failure. Consequently, the use of digital systems raises the concern about common-mode failure to a new level. A more methodical approach to mitigating common-mode failure is needed to address these concerns. Purposeful introduction of D-in-D ampersand D has been used as a defense against common-mode failure in reactor protection systems. At least two diverse systems are provided to mitigate any potential initiating event. Additionally, diverse displays and controls are provided to allow the operator to monitor plant status and manually initiate engineered safety features. A special form of conimon-mode failure analysis called ''defense-in-depth and diversity analysis'' has been developed to identify possible conimon-mode failure vulnerabilities in digital systems. An overview of this analysis technique is provided

Subcriticality monitoring method for reactor

International Nuclear Information System (INIS)

Ueda, Makoto.

1991-01-01

The present invention accurately monitors the reactor subcriticality and ensures the critical safety, irrespective of the presence or absence of artificial neutron sources. That is, when the subcriticality is monitored upon reactivity changing operation which causes reactivity change to the reactor during shutdown, neutron monitors are disposed at a plurality of monitoring positions. Then, neutron counting ratio before and after conducting the reactivity changing operation is determined. The subcriticality of the reactor is monitored by the ratio and the state of scattering of the ratio of neutron counting rate between each of the neutron monitors. With such procedures, signals of the neutron monitors are used, the characteristic that the change of the signals depend on the change of the neutron multiplication of the reactor core can be utilized whether artificial neutron sources (external neutron sources) are disposed or not. Accordingly, the subcriticality can be monitored more reliably. (I.S.)

Emergency cooling system for nuclear reactors

International Nuclear Information System (INIS)

Frisch, E.; Andrews, H.N.

1976-01-01

Upon the occasion of loss of coolant in a nuclear reactor as when a coolant supply or return line breaks, or both lines break, borated liquid coolant from an emergency source is supplied in an amount to absorb heat being generated in the reactor even after the control rods have been inserted. The liquid coolant flows from pressurized storage vessels outside the reactor to an internal manifold from which it is distributed to unused control rod guide thimbles in the reactor fuel assemblies. Since the guide thimbles are mounted at predetermined positions relative to heat generating fuel elements in the fuel assemblies, holes bored at selected locations in the guide thimble walls, sprays the coolant against the reactor fuel elements which continue to dissipate heat but at a reduced level. The cooling water evaporates upon contacting the fuel rods thereby removing the maximum amount of heat (970 BTU per pound of water) and after heat absorption will leave the reactor in the form of steam through the break which is the cause of the accident to help assure immediate core cooldown

Reactor antineutrino monitoring with a plastic scintillator array as a new safeguards method

OpenAIRE

Oguri, S.; Kuroda, Y.; Kato, Y.; Nakata, R.; Inoue, Y.; Ito, C.; Minowa, M.

2014-01-01

We developed a segmented reactor-antineutrino detector made of plastic scintillators for application as a tool in nuclear safeguards inspection and performed mostly unmanned field operations at a commercial power plant reactor. At a position outside the reactor building, we measured the difference in reactor antineutrino flux above the ground when the reactor was active and inactive.

Overview of the fast reactors fuels program

International Nuclear Information System (INIS)

Evans, E.A.; Cox, C.M.; Hayward, B.R.; Rice, L.H.; Yoshikawa, H.H.

1980-04-01

Each nation involved in LMFBR development has its unique energy strategies which consider energy growth projections, uranium resources, capital costs, and plant operational requirements. Common to all of these strategies is a history of fast reactor experience which dates back to the days of the Manhatten Project and includes the CLEMENTINE Reactor, which generated a few watts, LAMPRE, EBR-I, EBR-II, FERMI, SEFOR, FFTF, BR-1, -2, -5, -10, BOR-60, BN-350, BN-600, JOYO, RAPSODIE, Phenix, KNK-II, DFR, and PFR. Fast reactors under design or construction include PEC, CRBR, SuperPhenix, SNR-300, MONJU, and Madras (India). The parallel fuels and materials evolution has fully supported this reactor development. It has involved cermets, molten plutonium alloy, plutonium oxide, uranium metal or alloy, uranium oxide, and mixed uranium-plutonium oxides and carbides

Italian position paper on the safety analysis of liquid metal fast breeder reactors as related to sodium fires. The PEC reactor

International Nuclear Information System (INIS)

Gerosa, A.

1983-01-01

To obtain a deep understanding of physical phenomena and engineering problems connected to sodium fires, and to optimize the utilization of human and financial resources available, CNEN (now ENEA) has decided to join the French Commissariat a l'Energie Atomique (CEA) in the realization of a Franco-Italian experimental programme on sodium fires, named ESMERALDA. As for design preventions for PEC reactor (a fast flux, liquid metal cooled, fuel element testing reactor) fundamental choices were made taking into account all available knowledge, but with particular reference to the results of CEA's previous experiments on sodium fires. More detailed design analysis will be possible in the future, based on experimental results coming from the ESMERALDA programme

Pressurized heavy-water reactor safety

International Nuclear Information System (INIS)

Pease, L.; Wilson, R.

1977-09-01

CANDU-PWR type reactors routinely release small amounts of radioactive liquids and gases and large quantities of low-grade waste heat. Radioactive emissions are usually below 1% of the derived release limits based on ICRP limits. Waste heat is common to all power plants and is not foreseen as a problem in Canadian conditions. Risk analysis shows a very low accident probability for CANDU type reactors. Multiple barriers to release of radionuclides, quality assurance, control, and inspection, containment systems, the shutdown system, the ECCS, and leak-before-break design, would all combine to mitigate the effects of an accident. (E.C.B.)

Core arrangement in BWR type reactors

International Nuclear Information System (INIS)

Asano, Masayuki.

1981-01-01

Purpose: To decrease the number of fuel assemblies whose locations are to be changed upon fuel exchange, as well as unify the power distribution in the core by arranging, in a chess board configuration, a plurality pattern of unit reactor lattices each containing fuel assemblies of different burnup degrees in orthogonal positions to each other. Constitution: A first pattern of unit reactor lattice is formed by disposing fuel assemblies of burnup degree 1 and fuel assemblies of burnup degree 3 at orthogonal positions to each other. A second pattern of unit reactor lattice is formed by disposing fuel assemblies of burnup degree 2 and fuel assemblies of burnup degree 1 at orthogonal positions to each other. The unit lattices each in such a dispositions are arranged in a chess board arrangement. Since, the fuel assemblies of the burnup degree 1 in the first pattern unit lattices proceed to the burnup degree 2 and the fuel assemblies of the burnup degree 2 in the second pattern unit lattices proceed to the burnup degree 3 up to the fuel exchange stage, fuel exchange and movement have only to be made, not for those fuel assemblies, but for another half of the fuel assemblies. (Kawakami, Y.)

Study of startup conditions of a pulsed annular reactor

International Nuclear Information System (INIS)

Silva, Mario Augusto Bezerra da

2003-10-01

A new concept of reactor, which combines features of pulsed and stationary reactors, was proposed so as to produce intense neutronic fluxes. Such a reactor, known as VICHFPR (Very Intense Continuous High Flux Pulsed Reactor), consists of a subcritical core with an annular geometry and pulsed by a rotating reflector which acts as a reactivity modulator as it produces a short pulse (approximately equal to 1 ms) of high intensity, guiding the region near the pulser to super-prompt critical state. This dissertation intends to analyze the startup conditions of a Pulsed Annular Reactor. The evolution of the neutron pulse intensity is analyzed when the reactivity modulator is brought upwards according to a helicoidal path from its initial position (far away from the core), when the multiplication factor has a subcritical value, up to the final position (near the core), in which a super-prompt critical state is reached. Part of the analysis is based on the variation of neutron reflection, which is a uniform function of the exit and reflection angles between the core and the modulator. It must be emphasized that this work is an approximation of the real situation. As the initial and final reactor parameters are known, a programming code in Fortran is worked out to provide the multiplication factor and the flux intensity evolution. According to the results obtained with this code, the conditions under which the modulator must be lifted up during the startup are established. Basically, these conditions are related to the analysis of the rising and the rotation velocities, the reflector saving and the initial distance between the reactor and the modulator. The Pulsed Annular Reactor startup was divided into three stages. Because of its negative reactivity in the first two stages, the neutron multiplication is not large, while the last one, having a positive reactivity, shows an intense multiplication as is usually expected when handling pulsed systems. This last stage is quite

Study of the self-regulating properties of a WWER reactor

International Nuclear Information System (INIS)

Filo, J.; Trnkusz, J.; Polak, V.

1979-01-01

The results of a self-regulation experiment carried out on the V-1 reactor in Czechoslovakia in the period of the start-up are presented. The kinetic state of the reactor was modified by varying the position of the automatic control rods, the power of the turbogenerators and by switching off a main pump in the primary circuit, on power levels of 35, 55, 75 and 90%. The important thermal parameters of the reactor, the electric power of the turbogenerators, the neutron flux, the position of the automatic control rod group and the concentration of the boric acid in the coolant have been measured. (R.J.)

Support structure for reactor core constituent element

International Nuclear Information System (INIS)

Aida, Yasuhiko.

1993-01-01

A connection pipe having an entrance nozzle inserted therein as a reactor core constituent element is protruded above the upper surface of a reactor core support plate. A through hole is disposed to the protruding portion of the connection pipe. The through hole and a through hole disposed to the reactor core support plate are connected by a communication pipe. A shear rod is disposed in a horizontal portion at the inside of the communication pipe and is supported by a spring horizontally movably. Further, a groove is disposed at a position of the entrance nozzle opposing to the shear rod. The shear rod is urged out of the communication pipe by the pressure of the high pressure plenum and the top end portion of the shear rod is inserted to the groove of the entrance nozzle during operation. Accordingly, the shear rod is positioned in a state where it is extended from the through hole of the communication pipe to the groove of the entrance nozzle. This can mechanically constrain the rising of the reactor core constituent elements by the shear rod upon occurrence of earthquakes. (I.N.)

Fuel transfer system for a nuclear reactor

International Nuclear Information System (INIS)

Katz, L.R.; Marshall, J.R.; Desmarchais, W.E.

1977-01-01

Disclosed is a fuel transfer system for moving nuclear reactor fuel assemblies from a new fuel storage pit to a containment area containing the nuclear reactor, and for transferring spent fuel assemblies under water from the reactor to a spent fuel storage area. The system includes an underwater track which extends through a wall dividing the fuel building from the reactor containment and a car on the track serves as the vehicle for moving fuel assemblies between these two areas. The car is driven by a motor and linkage extending from an operating deck to a chain belt drive on the car. A housing pivotally mounted at its center on the car is hydraulically actuated to vertically receive a fuel assembly which then is rotated to a horizontal position to permit movement through the wall between the containment and fuel building areas. Return to the vertical position provides for fuel assembly removal and the reverse process is repeated when transferring an assembly in the opposite direction. Limit switches used in controlling operation of the system are designed to be replaced from the operating deck when necessary by tools designed for this purpose. 5 claims, 8 figures

Lessons from feedback of safety operating experience for reactor physics

International Nuclear Information System (INIS)

Suchomel, J.; Rapavy, S.

1999-01-01

Analyses of events in WWER operations as a part of safety experience feedback provide a valuable source of lessons for reactor physics. Examples of events from Bohunice operation will be shown such as events with inadequate approach to criticality, positive reactivity insertions, expulsion of a control rod from shut-down reactor, problems with reactor protection system and control rods. (Authors)

Positioning and locking device for fuel pin to grid attachment

International Nuclear Information System (INIS)

Frick, T.M.; Wineman, A.L.

1976-01-01

A positioning and locking device for fuel pin to grid attachment provides an inexpensive means of positively positioning and locking the individual fuel pins which make up the driver fuel assemblies used in nuclear reactors. The device can be adapted for use with a currently used attachment grid assembly design and insures that the pins remain in their proper position throughout the in-reactor life of the assembly. This device also simplifies fuel bundle assembly in that a complete row of fuel pins can be added to the bundle during each step of assembly. 8 claims, 8 drawing figures

A digital position-indication system for control rods

International Nuclear Information System (INIS)

Nishizawa, Yukio; Hayakawa, Toshifumi

1979-01-01

Systems that detect and indicate the position of the control rods that regulate the thermal output of a nuclear reactor play a particularly important role in monitoring its operational status. Conventionally, control rod position indication in pressurized water reactors has been of the analog type, utilizing the principle of the differential transformer. The present digital system was developed with the objective of achieving greater stability, greater accuracy, and higher reliability. The article gives a general description of the system and describes its advantages. (author)

Peptidolytic microbial community of methanogenic reactors from two modified UASBs of brewery industries

Directory of Open Access Journals (Sweden)

C. Díaz

2010-10-01

Full Text Available We studied the peptide-degrading anaerobic communities of methanogenic reactors from two mesophilic full-scale modified upflow anaerobic sludge blanket (UASB reactors treating brewery wastewater in Colombia. Most probable number (MPN counts varied between 7.1 x 10(8 and 6.6 x 10(9 bacteria/g volatile suspended solids VSS (Methanogenic Reactor 1 and 7.2 x 10(6 and 6.4 x 10(7 bacteria/g (VSS (Methanogenic Reactor 2. Metabolites detected in the highest positive MPN dilutions in both reactors were mostly acetate, propionate, isovalerate and, in some cases, negligible concentrations of butyrate. Using the highest positive dilutions of MPN counts, 50 dominant strains were isolated from both reactors, and 12 strains were selected for sequencing their 16S rRNA gene based on their phenotypic characteristics. The small-subunit rRNA gene sequences indicated that these strains were affiliated to the families Propionibacteriaceae, Clostridiaceae and Syntrophomonadaceae in the low G + C gram-positive group and Desulfovibrio spp. in the class d-Proteobacteria. The main metabolites detected in the highest positive dilutions of MPN and the presence of Syntrophomonadaceae indicate the effect of the syntrophic associations on the bioconversion of these substrates in methanogenic reactors. Additionally, the potential utilization of external electron acceptors for the complete degradation of amino acids by Clostridium strains confirms the relevance of these acceptors in the transformation of peptides and amino acids in these systems.

Job task and functional analysis of the Division of Reactor Projects, office of Nuclear Reactor Regulation. Final report

Energy Technology Data Exchange (ETDEWEB)

Morzinski, J.A.; Gilmore, W.; Hahn, H.A.

1998-07-10

A job task and functional analysis was recently completed for the positions that make up the regional Divisions of Reactor Projects. Among the conclusions of that analysis was a recommendation to clarify roles and responsibilities among site, regional, and headquarters personnel. As that analysis did not cover headquarters personnel, a similar analysis was undertaken of three headquarters positions within the Division of Reactor Projects: Licensing Assistants, Project Managers, and Project Directors. The goals of this analysis were to systematically evaluate the tasks performed by these headquarters personnel to determine job training requirements, to account for variations due to division/regional assignment or differences in several experience categories, and to determine how, and by which positions, certain functions are best performed. The results of this analysis include recommendations for training and for job design. Data to support this analysis was collected by a survey instrument and through several sets of focus group meetings with representatives from each position.

Testing of research reactor fuel in the high flux reactor (Petten)

International Nuclear Information System (INIS)

Guidez, J.; Markgraf, J.W.; Sordon, G.; Wijtsma, F.J.; Thijssen, P.J.M.; Hendriks, J.A.

1999-01-01

The two types of fuel most frequently used by the main research reactors are metallic: highly enriched uranium (>90%) and silicide low enriched uranium ( 3 . However, a need exists for research on new reactor fuel. This would permit some plants to convert without losses in flux or in cycle length and would allow new reactor projects to achieve higher possibilities especially in fluxes. In these cases research is made either on silicide with higher density, or on other types of fuel (UMo, etc.). In all cases when new fuel is proposed, there is a need, for safety reasons, to test it, especially regarding the mechanical evolution due to burn-up (swelling, etc.). Initially, such tests are often made with separate plates, but lately, using entire elements. Destructive examinations are often necessary. For this type of test, the High Flux Reactor, located in Petten (The Netherlands) has many specific advantages: a large core, providing a variety of interesting positions with high fluence rate; a downward coolant flow simplifies the engineering of the device; there exists easy access with all handling possibilities to the hot-cells; the high number of operating days (>280 days/year), together with the high flux, gives a possibility to reach quickly the high burn-up needs; an experienced engineering department capable of translating specific requirements to tailor-made experimental devices; a well equipped hot-cell laboratory on site to perform all necessary measurements (swelling, γ-scanning, profilometry) and all destructive examinations. In conclusion, the HFR reactor readily permits experimental research on specific fuels used for research reactors with all the necessary facilities on the Petten site. (author)

Water feeding method upon reactor isolation

International Nuclear Information System (INIS)

Sasaki, Koichi; Takahara, Kuniaki; Hamamura, Kenji; Arakawa, Masahiro.

1990-01-01

The present invention concerns a method of feeding water upon reactor isolation in a plural loop type reactor having a plurality of reactor cooling systems. Water can be injected to a plurality of pools even if the pressure between the pools is not balanced and the water level in the reactor cooling system is optimally controlled. That is, water can be injected in accordance with the amount required for each of the pools by setting the opening of a turbine inlet steam control valve to somewhat higher than the cooling system pressure of the highest pressure loop. Water feeding devices upon reactor isolation were required by the same number as that for the reactor cooling systems. Whereas since pumps and turbines are used in common without worsening the water injection controllability to each of the loops according to the method of this invention and, accordingly, the cost performance can be improved. Further, since the opening degree of the turbine inlet steam control valve is controlled while making the difference pressure constant between the turbine inlet pressure and the pump exhaust pressure, the amount of the turbine exhausted steams can be reduced and, further, water injection controllability of the flow rate control valve in the injection line is improved. (I.S.)

Venting device for nuclear reactor container

International Nuclear Information System (INIS)

Yamashita, Masahiro; Ogata, Ken-ichi.

1994-01-01

An airtight vessel of a venting device of a nuclear reactor container is connected with a reactor container by way of a communication pipeline. A feed water tank is disposed at a position higher than the liquid surface of scrubbing water in the airtight vessel for supplying scrubbing water to the airtight vessel. In addition, a scrubbing water storage tank is disposed at a position hither than the feed water tank for supplying scrubbing water to the feed water tank. Storage water in the feed water tank is introduced into the airtight vessel by the predetermined opening operation of a valve by the pressure exerted on the liquid surface and the own weight of the storage water. Further, the storage water in the scrubbing water storage tank is led into the feed water tank by the water head pressure. The scrubbing water for keeping the performance of the venting device of the reactor container can be supplied by a highly reliable method without using AC power source or the like as a driving source. (I.N.)

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

International Nuclear Information System (INIS)

Faghihi, F.; Mirvakili, S.M.

2009-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2009-06-15

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

Hold-down device for nuclear reactor

International Nuclear Information System (INIS)

Leclercq, J.; Bonnamour, M.

1984-01-01

The present device can be used in nuclear reactors and more particularly in pressurized water reactors consisting of coupled fuel assemblies, certain of which are equipped with non-displaceable elements carried by an unsertable member. The device comprises the unsertable member provided with at least two sets of springs which transmit the load of an upper structure common to the fuel assemblies ajacent that which supports the unsertable member. The device is used to hold-down fuel assemblies which are subjected to the forces of circulating coolant [fr

Photocatalytic reactors for treating water pollution with solar illumination. II: a simplified analysis for flow reactors

Energy Technology Data Exchange (ETDEWEB)

Sagawe, G.; Bahnemann, D. [Inst. fuer Technische Chemie, Univ. Hannover, Hannover (Germany); Brandi, R.J.; Cassano, A.E. [INTEC (Univ. Nacional del Litoral and CONICET), Santa Fe (Argentina)

2003-07-01

Very frequently outgoing streams of real wastewaters do not have a definite and constant composition. Additionally, when the degradation process makes use of solar irradiation, the photon flux is hardly constant. These two factors strongly militate against the use of very elaborate, exact models for analyzing the performance of the employed reactors. In these cases, approximate methods may be the most practical approach. One possible way is presented in this paper. The observed photonic efficiency concept developed in a previous contribution (sagawe et al., 2002a) is applied to continuous reactors for both steady state and transient operations of photocatalytic reactions applied to wastewaters decontamination processes. For this reactor the local observed photonic efficiency, defined at each reactor longitudinal position, is the convenient property to express the concentration spatial evolution. It is also shown that the description of the reactor performance employing a mass balance can be done in a rather simple way introducing a mass-moving coordinate transformation that remodel the mass inventory and permits working with simpler ordinary differential equations. (orig.)

Radiation shield for nuclear reactors

International Nuclear Information System (INIS)

Weissenfluh, J.A.

1978-01-01

A shield for use with nuclear reactor systems to attenuate radiation resulting from reactor operation is described. The shield comprises a container preferably of a thin, flexible or elastic material, which may be in the form of a bag, a mattress, a toroidal segment or toroid or the like filled with radiation attenuating liuid. Means are provided in the container for filling and draining the container in place. Due to its flexibility, the shield readily conforms to irregularities in surfaces with which it may be in contact in a shielding position

NEUTRONIC REACTOR CONTROL ROD DRIVE APPARATUS

Science.gov (United States)

Oakes, L.C.; Walker, C.S.

1959-12-15

ABS>A suspension mechanism between a vertically movable nuclear reactor control rod and a rod extension, which also provides information for the operator or an automatic control signal, is described. A spring connects the rod extension to a drive shift. The extension of the spring indicates whether (1) the rod is at rest on the reactor, (2) the rod and extension are suspended, or (3) the extension alone is suspended, the spring controlling a 3-position electrical switch.

Application of the neutron noise technique for measurement of reactivity for subcritical reactor RA-4

International Nuclear Information System (INIS)

Orso, J; Marenzana, A

2012-01-01

Reactor core RA-4 is divided into two parts that come together to start reactor. The reactor with core separate has the largest subcritical condition, this condition is more secure and therefore the reactor shutdown. In this paper measurements are made of the decay constant of the neutron prompt ' P ', using the α-Rossi and α-Feynman methods to calculate the reactivity of the reactor core for different positions. Both techniques are compared and reactivity is obtained for several position of the reactor core using the α-Rossi technical which is obtained a function that gives the reactivity depending on the separation of the core length. Both techniques are verified using a no multiplicative system. Reactivity values for different position of the core obtained by α-Rossi technique are: $[0 cm] = (-11+/-1) dollar, $[3 cm] = (-7+/-1) dollar, $[3.5 cm] (-5.5+/-0.8) dollar, $[4.2 cm] = (-3.8+/-0.3) dollar y $[4.5] = (-3.0+/-0.1) dollar (author)

Reactor theory and power reactors. 1. Calculational methods for reactors. 2. Reactor kinetics

International Nuclear Information System (INIS)

Henry, A.F.

1980-01-01

Various methods for calculation of neutron flux in power reactors are discussed. Some mathematical models used to describe transients in nuclear reactors and techniques for the reactor kinetics' relevant equations solution are also presented

Report on the Survey of the Design Review of New Reactor Applications. Volume 4: Reactor Coolant and Associated Systems

International Nuclear Information System (INIS)

Downey, Steven; Monninger, John; Nevalainen, Janne; Joyer, Philippe; Koley, Jaharlal; Kawamura, Tomonori; Chung, Yeon-Ki; Haluska, Ladislav; Persic, Andreja; Reierson, Craig; Monninger, John; Choi, Young-Joon; )

2017-01-01

At the tenth meeting of the Committee on Nuclear Regulatory Activities (CNRA) Working Group on the Regulation of New Reactors (WGRNR) in March 2013, the Working Group agreed to present the responses to the Second Phase, or Design Phase, of the licensing process survey as a multi-volume text. As such, each report will focus on one of the eleven general technical categories covered in the survey. The general technical categories were selected to conform to the topics covered in the International Atomic Energy Agency (IAEA) Safety Guide GS-G-4.1. This report provides a discussion of the survey responses related to the Reactor Coolant and Associated Systems category. The Reactor Coolant and Associated Systems category includes the following technical topics: overpressure protection, reactor coolant pressure boundary, reactor vessel, and design of the reactor coolant system. For each technical topic, the member countries described the information provided by the applicant, the scope and level of detail of the technical review, the technical basis for granting regulatory authorisation, the skill sets required and the level of effort needed to perform the review. Based on a comparison of the information provided by the member countries in response to the survey, the following observations were made: - Although the description of the information provided by the applicant differs in scope and level of detail among the member countries that provided responses, there are similarities in the information that is required. - All of the technical topics covered in the survey are reviewed in some manner by all of the regulatory authorities that provided responses. - It is common to consider operating experience and lessons learnt from the current fleet during the review process. - The most commonly and consistently identified technical expertise needed to perform design reviews related to this category are mechanical engineering and materials engineering. The complete survey

Operational experience at the AFRRI-TRIGA reactor facility (1972-1974)

Energy Technology Data Exchange (ETDEWEB)

McKenzie, J L [Armed Forces Radiobiology Research Institute, Bethesda, MD (United States)

1974-07-01

The Armed Forces Radiobiology Research Institute operates a TRIGA Mark-F Reactor which has a movable core, and the capability to operate in the steady state mode up to a maximum power level of one megawatt and in the pulse mode up to a maximum peak power of 2600 MW (10 millisecond pulse). The reactor experienced three operational incidents during the period from February 1972 to February 1974, and two of these incidents were reportable to the Atomic Energy Commission. The first incident consisted of a failure of a weld at the top of the tri-flute on an instrumented fuel element which allowed the tri-flute to move up about one-half inch from its normal position. The instrumented fuel element was removed from the reactor core and replaced with a new instrumented fuel element. The second incident consisted of a malfunction of the reactor core position safety interlock which resulted in the lead shield doors closing around the reactor core shroud. The lead shield doors did not make contact with the reactor core shroud and therefore no damage occurred. The incident was reported to the Atomic Energy Commission. The third incident consisted of a failure of the threaded connector on the top of the transient control rod which allowed the transient control rod to separate from the connecting rod and drop to the bottom of the guide tube. The damaged transient control rod was removed from the guide tube and a new transient rod was installed in the reactor core. This incident was reported to the Atomic Energy Commission. A modification was made to Exposure Room 2 which consisted of placing panels, painted with gadolinium oxide paint, on the walls, ceiling, and reactor core tank projection. This resulted in the {sup 41}Ar production rate and the effluent release to the environment being reduced by a factor of 10 to 20, depending upon the position of the reactor core. (author)

Operational experience at the AFRRI-TRIGA reactor facility (1972-1974)

International Nuclear Information System (INIS)

McKenzie, J.L.

1974-01-01

The Armed Forces Radiobiology Research Institute operates a TRIGA Mark-F Reactor which has a movable core, and the capability to operate in the steady state mode up to a maximum power level of one megawatt and in the pulse mode up to a maximum peak power of 2600 MW (10 millisecond pulse). The reactor experienced three operational incidents during the period from February 1972 to February 1974, and two of these incidents were reportable to the Atomic Energy Commission. The first incident consisted of a failure of a weld at the top of the tri-flute on an instrumented fuel element which allowed the tri-flute to move up about one-half inch from its normal position. The instrumented fuel element was removed from the reactor core and replaced with a new instrumented fuel element. The second incident consisted of a malfunction of the reactor core position safety interlock which resulted in the lead shield doors closing around the reactor core shroud. The lead shield doors did not make contact with the reactor core shroud and therefore no damage occurred. The incident was reported to the Atomic Energy Commission. The third incident consisted of a failure of the threaded connector on the top of the transient control rod which allowed the transient control rod to separate from the connecting rod and drop to the bottom of the guide tube. The damaged transient control rod was removed from the guide tube and a new transient rod was installed in the reactor core. This incident was reported to the Atomic Energy Commission. A modification was made to Exposure Room 2 which consisted of placing panels, painted with gadolinium oxide paint, on the walls, ceiling, and reactor core tank projection. This resulted in the 41 Ar production rate and the effluent release to the environment being reduced by a factor of 10 to 20, depending upon the position of the reactor core. (author)

Nuclear reactor refuelable in space

International Nuclear Information System (INIS)

El-Genk, M.S.; Buden, D.; Mims, J.E.

1992-01-01

This patent describes a gas cooled nuclear reactor suitable for use in space. It comprises a lightweight structure comprising a plurality of at least three sections, each sector comprising a container for a reactor core separate and distinct from the reactor cores of the other sectors, each sector being capable of operating on its own and in cooperation with one or more of the other sectors and each sector having a common juncture with every other structure; and means associated with each sector independently introducing gas coolant into and extracting coolant from each sector to cool the core therein, wherein in event of failure of the cooling system of a core in a sector, one or more of the other sectors comprise means for conducting heat away from the failed sector core and means for convecting the heat away, and wherein operation of the one or more other sectors is maintained

A charge regulating system for turbo-generator gas-cooled high-temperature reactor power stations

International Nuclear Information System (INIS)

Braytenbah, A.S.; Jaegtnes, K.O.

1975-01-01

The invention relates to a regulating system for gas-cooled high-temperature reactors power stations (helium coolant), equipped with several steam-boilers, each of which deriving heat from a corresponding cooling-gas flow circulating in the reactor, so as to feed superheated steam into a main common steam-manifold and re-superheated steam into a re-superheated hot common manifold [fr

Optimal operating parameters of the reactor oscillator in the channel of 6.5/10 MW reactor; Odredjivanje optimalnih radnih tacaka za reaktorski oscilator u kanalu na reaktoru 6,5/10 MW

Energy Technology Data Exchange (ETDEWEB)

Lolic, B; Zecevic, V [The Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1961-07-01

Operating conditions for the reactor oscillator in the central vertical experimental channel (VK5) in the RA reactor were studied during 1960. Channel VK5 was chosen because the sensitivity of the reactor is highest in this case. The central vertical experimental channel is placed in the center of the core and its bottom is placed 200 mm from the bottom of the reactor core. Diameter of the channel is is 110 mm and its length 5706 mm. During operation of the reactor oscillator with total modulation of the reactor power, it is very important to determine the oscillator operating point and the oscillation amplitude in such a way to avoid any change in reactor power level. Positive reactivity changes originating from oscillations of the samples should be compensated by the negative reactivity changes so that the effect should be nil. Operating points of the reactor oscillator are in the middle of the straight part of the figure showing the reactivity change dependent on the position of the absorber.

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.)

The European pressurized water reactor

International Nuclear Information System (INIS)

Leny, J.C.

1993-01-01

The present state of development of the European Pressurized Water Reactor (EPR) is outlined. During the so-called harmonization phase, the French and German utilities drew up their common requirements and evaluated the reactor concept developed until then with respect to these requirements. A main result of the harmonization phase was the issue, in September 1993, of the 'EPR Conceptual Safety Feature Review File' to be jointly assessed by the safety authorities in France and Germany. The safety objectives to be met by the EPR are specified in the second part of the paper, and some details of the primary and secondary side safety systems are given. (orig.) [de

Space dependence of reactivity parameters on reactor dynamic perturbation measurements

International Nuclear Information System (INIS)

Maletti, R.; Ziegenbein, D.

1985-01-01

Practical application of reactor-dynamic perturbation measurements for on-power determination of differential reactivity weight of control rods and power coefficients of reactivity has shown a significant dependence of parameters on the position of outcore detectors. The space dependence of neutron flux signal in the core of a VVER-440-type reactor was measured by means of 60 self-powered neutron detectors. The greatest neutron flux alterations are located close to moved control rods and in height of the perturbation position. By means of computations, detector positions can be found in the core in which the one-point model is almost valid. (author)

Operation experience at the Neuherberg Research Reactor (FRN) with several modifications of reactor components

Energy Technology Data Exchange (ETDEWEB)

Demmeler, M; Rau, G [Gesellschaft fuer Strahlen- und Umweltforschung mbH, Neuherberg (Germany)

1974-07-01

Since the first full power operation in September 1972 up till now (Dec. 1973) the TRIGA Mark III reactor FRN has run more than 500 MWh in steady state operation and has been pulsed for 265 times. During startup experiments, neutron- and gamma-flux mapping has been performed with special technical devices in the core and in several irradiation positions, mainly in the thermal column and in the exposure room. Furthermore reactivity values of each fuel element have been measured at full power of 1 MW, thus enabling a more accurate burnup calculation. Troubles with the rotary specimen rack occurred at power rates above 280 kW; here, the lazy susan stuck, caused by thermal stress. Thus it will be replaced by a hydraulic-operated type, which has been developed at the TRIGA reactor Heidelberg. In order to increase irradiation capacity, a new core configuration has been set up a few months ago, replacing several fuel-reflector-elements by irradiation tubes within the grid-plate positions E-22, G-2, G-17 and G-36. Four additional fuel elements had to be inserted to compensate for the resulting reactivity losses. The original plan of regaining sufficient excess-reactivity by inserting a fuel element in grid-plate position A-l failed because of local boiling in the center of the core by 1 MW-operation. Experiments at the reactor started with the begin of routine-operation in September 1973. Up till now, a total of 450 neutron- and gamma- irradiations have been performed, mainly for neutron-activations. (author)

Advances in fast reactor technology. Proceedings of the 30. meeting of the International Working Group on Fast Reactors

International Nuclear Information System (INIS)

1998-04-01

Individual States were largely responsible for early developments in experimental and prototype liquid metal fast reactors (LMFRs). However, for development of advanced LMFRs, international co-operation plays an important role. The IAEA seeks to promote such co-operation. For R and D incorporating innovative features, international co-operation allows pooling of resources and expertise in areas of common interest. Information on experience gained from R and D, and from the operation and construction of fast reactors, has been reviewed periodically by the International Working Group on Fast Reactors (IWGFR). These proceedings contain updated a new information on the status of LMFR development, as reported at the 30th meeting of the IWGFR, held in Beijing, China, from 13 to 16 May 1997

Advances in fast reactor technology. Proceedings of the 30. meeting of the International Working Group on Fast Reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-04-01

Individual States were largely responsible for early developments in experimental and prototype liquid metal fast reactors (LMFRs). However, for development of advanced LMFRs, international co-operation plays an important role. The IAEA seeks to promote such co-operation. For R and D incorporating innovative features, international co-operation allows pooling of resources and expertise in areas of common interest. Information on experience gained from R and D, and from the operation and construction of fast reactors, has been reviewed periodically by the International Working Group on Fast Reactors (IWGFR). These proceedings contain updated a new information on the status of LMFR development, as reported at the 30th meeting of the IWGFR, held in Beijing, China, from 13 to 16 May 1997. Refs,figs,tabs.

Education and Training on ISIS Research Reactor

International Nuclear Information System (INIS)

Foulon, F.; Badeau, G.; Lescop, B.; Wohleber, X.

2013-01-01

In the frame of academic and vocational programs the National Institute for Nuclear Science and Technology uses the ISIS research reactor as a major tool to ensure a practical and comprehensive understanding of the nuclear reactor physics, principles and operation. A large set of training courses have been developed on ISIS, optimising both the content of the courses and the pedagogical approach. Programs with duration ranging from 3 hours (introduction to reactor operation) to 24 hours (full program for the future operators of research reactors) are carried out on ISIS reactor. The reactor is operated about 350 hours/year for education and training, about 40 % of the courses being carried out in English. Thus, every year about 400 trainees attend training courses on ISIS reactor. We present here the ISIS research reactor and the practical courses that have been developed on ISIS reactor. Emphasis is given to the pedagogical method which is used to focus on the operational and safety aspects, both in normal and incidental operation. We will present the curricula of the academic and vocational courses in which the practical courses are integrated, the courses being targeted to a wide public, including operators of research reactors, engineers involved in the design and operation of nuclear reactors as well as staff of the regulatory body. We address the very positive impact of the courses on the development of the competences and skills of participants. Finally, we describe the Internet Reactor Laboratories (IRL) that are under development and will consist in broadcasting the training courses via internet to remote facilities or institutions

Education and Training on ISIS Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Foulon, F.; Badeau, G.; Lescop, B.; Wohleber, X. [French Atomic Energy and Alternative Energies Commission, Paris (France)

2013-07-01

In the frame of academic and vocational programs the National Institute for Nuclear Science and Technology uses the ISIS research reactor as a major tool to ensure a practical and comprehensive understanding of the nuclear reactor physics, principles and operation. A large set of training courses have been developed on ISIS, optimising both the content of the courses and the pedagogical approach. Programs with duration ranging from 3 hours (introduction to reactor operation) to 24 hours (full program for the future operators of research reactors) are carried out on ISIS reactor. The reactor is operated about 350 hours/year for education and training, about 40 % of the courses being carried out in English. Thus, every year about 400 trainees attend training courses on ISIS reactor. We present here the ISIS research reactor and the practical courses that have been developed on ISIS reactor. Emphasis is given to the pedagogical method which is used to focus on the operational and safety aspects, both in normal and incidental operation. We will present the curricula of the academic and vocational courses in which the practical courses are integrated, the courses being targeted to a wide public, including operators of research reactors, engineers involved in the design and operation of nuclear reactors as well as staff of the regulatory body. We address the very positive impact of the courses on the development of the competences and skills of participants. Finally, we describe the Internet Reactor Laboratories (IRL) that are under development and will consist in broadcasting the training courses via internet to remote facilities or institutions.

Power control device in nuclear reactor

International Nuclear Information System (INIS)

Koyama, Kazuaki.

1981-01-01

Purpose: To enable smooth power changes in power conditioning systems by calculating forecast values for the neutron flux distribution and power distribution and by controlling the driving speed of control rods so as to correspond the forecast values with aimed values. Constitution: Control rod position is detected by a position detector and sent to a control computer as the position information. At the same time, the neutron flux distribution information is obtained by the neutron monitors, the power distribution information is obtained by a reactor power computer and they are outputted to the control computer. The control computer calculates the forecast values for the neutron flux distribution and the reactor power distribution from the information, and compares them with the aimed values from a setter and then outputs control signals so as to correspond the forecast values with the aimed values. The control rods can be inserted in appropriate velocity by the control signals. (Horiuchi, T.)

Validation of neutron flux redistribution factors in JSI TRIGA reactor due to control rod movements

International Nuclear Information System (INIS)

Kaiba, Tanja; Žerovnik, Gašper; Jazbec, Anže; Štancar, Žiga; Barbot, Loïc; Fourmentel, Damien; Snoj, Luka

2015-01-01

For efficient utilization of research reactors, such as TRIGA Mark II reactor in Ljubljana, it is important to know neutron flux distribution in the reactor as accurately as possible. The focus of this study is on the neutron flux redistributions due to control rod movements. For analyzing neutron flux redistributions, Monte Carlo calculations of fission rate distributions with the JSI TRIGA reactor model at different control rod configurations have been performed. Sensitivity of the detector response due to control rod movement have been studied. Optimal radial and axial positions of the detector have been determined. Measurements of the axial neutron flux distribution using the CEA manufactured fission chambers have been performed. The experiments at different control rod positions were conducted and compared with the MCNP calculations for a fixed detector axial position. In the future, simultaneous on-line measurements with multiple fission chambers will be performed inside the reactor core for a more accurate on-line power monitoring system. - Highlights: • Neutron flux redistribution due to control rod movement in JSI TRIGA has been studied. • Detector response sensitivity to the control rod position has been minimized. • Optimal radial and axial detector positions have been determined

Software verification and validation methodology for advanced digital reactor protection system using diverse dual processors to prevent common mode failure

International Nuclear Information System (INIS)

Son, Ki Chang; Shin, Hyun Kook; Lee, Nam Hoon; Baek, Seung Min; Kim, Hang Bae

2001-01-01

The Advanced Digital Reactor Protection System (ADRPS) with diverse dual processors is being developed by the National Research Lab of KOPEC for ADRPS development. One of the ADRPS goals is to develop digital Plant Protection System (PPS) free of Common Mode Failure (CMF). To prevent CMF, the principle of diversity is applied to both hardware design and software design. For the hardware diversity, two different types of CPUs are used for Bistable Processor and Local Coincidence Logic Processor. The VME based Single Board Computers (SBC) are used for the CPU hardware platforms. The QNX Operating System (OS) and the VxWorks OS are used for software diversity. Rigorous Software Verification and Validation (V and V) is also required to prevent CMF. In this paper, software V and V methodology for the ADRPS is described to enhance the ADRPS software reliability and to assure high quality of the ADRPS software

Fuel burnup analysis for the Moroccan TRIGA research reactor

International Nuclear Information System (INIS)

El Bakkari, B.; El Bardouni, T.; Nacir, B.; El Younoussi, C.; Boulaich, Y.; Boukhal, H.; Zoubair, M.

2013-01-01

Highlights: ► A fuel burnup analysis of the 2 MW TRIGA MARK II Moroccan research reactor was established. ► Burnup calculations were done by means of the in-house developed burnup code BUCAL1. ► BUCAL1 uses the MCNP tallies directly in the calculation of the isotopic inventories. ► The reactor life time was found to be 3360 MW h considering full power operating conditions. ► Power factors and fluxes of the in-core irradiation positions are strongly affected by burnup. -- Abstract: The fundamental advantage and main reason to use Monte Carlo methods for burnup calculations is the possibility to generate extremely accurate burnup dependent one group cross-sections and neutron fluxes for arbitrary core and fuel geometries. Yet, a set of values determined for a material at a given position and time remains accurate only in a local region, in which neutron spectrum and flux vary weakly — and only for a limited period of time, during which changes of the local isotopic composition are minor. This paper presents the approach of fuel burnup evaluation used at the Moroccan TRIGA MARK II research reactor. The approach is essentially based upon the utilization of BUCAL1, an in-house developed burnup code. BUCAL1 is a FORTRAN computer code designed to aid in analysis, prediction, and optimization of fuel burnup performance in nuclear reactors. The code was developed to incorporate the neutron absorption reaction tally information generated directly by MCNP5 code in the calculation of fissioned or neutron-transmuted isotopes for multi-fueled regions. The fuel cycle length and changes in several core parameters such as: core excess reactivity, control rods position, fluxes at the irradiation positions, axial and radial power factors and other parameters are estimated. Besides, this study gives valuable insight into the behavior of the reactor and will ensure better utilization and operation of the reactor during its life-time and it will allow the establishment of

Flexible Conversion Ratio Fast Reactor Systems Evaluation

Energy Technology Data Exchange (ETDEWEB)

Neil Todreas; Pavel Hejzlar

2008-06-30

Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores reated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcme the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better themal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor.

Flexible Conversion Ratio Fast Reactor Systems Evaluation

International Nuclear Information System (INIS)

Neil Todreas; Pavel Hejzlar

2008-01-01

Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores treated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcome the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better thermal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor

Safety Research Experiment Facility Project. Conceptual design report. Volume VII. Reactor cooling

International Nuclear Information System (INIS)

1975-12-01

The Reactor Cooling System (RCS) will provide the required cooling during test operations of the Safety Research Experiment Facility (SAREF) reactor. The RCS transfers the reactor energy generated in the core to a closed-loop water storage system located completely inside the reactor containment building. After the reactor core has cooled to a safe level, the stored heat is rejected through intermediate heat exchangers to a common forced-draft evaporative cooling tower. The RCS is comprised of three independent cooling loops of which any two can remove sufficient heat from the core to prevent structural damage to the system components

Modeling issues associated with production reactor safety assessment

International Nuclear Information System (INIS)

Stack, D.W.; Thomas, W.R.

1990-01-01

This paper describes several Probabilistic Safety Assessment (PSA) modeling issues that are related to the unique design and operation of the production reactors. The identification of initiating events and determination of a set of success criteria for the production reactors is of concern because of their unique design. The modeling of accident recovery must take into account the unique operation of these reactors. Finally, a more thorough search and evaluation of common-cause events is required to account for combinations of unique design features and operation that might otherwise not be included in the PSA. It is expected that most of these modeling issues also would be encountered when modeling some of the other more unique reactor and nonreactor facilities that are part of the DOE nuclear materials production complex. 9 refs., 2 figs

Overview of the fast reactors fuels program. [LMFBR

Energy Technology Data Exchange (ETDEWEB)

Evans, E.A.; Cox, C.M.; Hayward, B.R.; Rice, L.H.; Yoshikawa, H.H.

1980-04-01

Each nation involved in LMFBR development has its unique energy strategies which consider energy growth projections, uranium resources, capital costs, and plant operational requirements. Common to all of these strategies is a history of fast reactor experience which dates back to the days of the Manhatten Project and includes the CLEMENTINE Reactor, which generated a few watts, LAMPRE, EBR-I, EBR-II, FERMI, SEFOR, FFTF, BR-1, -2, -5, -10, BOR-60, BN-350, BN-600, JOYO, RAPSODIE, Phenix, KNK-II, DFR, and PFR. Fast reactors under design or construction include PEC, CRBR, SuperPhenix, SNR-300, MONJU, and Madras (India). The parallel fuels and materials evolution has fully supported this reactor development. It has involved cermets, molten plutonium alloy, plutonium oxide, uranium metal or alloy, uranium oxide, and mixed uranium-plutonium oxides and carbides.

Regulation for installation and operation of reactor

International Nuclear Information System (INIS)

1977-01-01

Concerning the description of an application for the approval of installation of a reactor, stipulated in Article 23 paragraph 2 of the Law for Regulation of Nuclear Source Materials, Nuclear Fuel Materials and Reactors (hereinafter referred to as the Law), the following items must be written. Namely, the heat output of the reactor in Article 23 paragraph 2 item 3 of the Law, the position, structure and facilities of the reactor facilities, described according to the stipulated classifications, the work plan, nuclear fuel materials employed, and the disposal of spent fuel. Concerning an application for the approval of a reactor installed aboard a foreign ship, stipulations are made separately. Description of an application for the approval of change of the heat output of a reactor and others should include the stipulated items. When it is wished to undergo inspection of the construction and performance of reactor facilities, an application for that end including the required items should be filed. Various safety measures preventing personnel from being exposed to radiation should be taken. When a foreign atomic-powered ship tries to enter a Japanese port, the stipulated necessary informations should be reported 60 days before such ship actually enters the Japanese port. A chief technician of reactors should take and pass the official examination. (Rikitake, Y.)

Intrinsically Safe and Economical Reactor (ISER)

International Nuclear Information System (INIS)

Wakabayashi, Hiroaki; Asahi, Yoshiro

1991-01-01

The Intrinsically Safe and Economical Reactor (ISER) is designed based on the principle of a process inherent ultimate safe reactor, PIUS, a so-called inherently safe reactor (ISR). ISER has been developed joingly by the members of the Kanagawa Institute of Technology, the University of Tokyo, the Japan Atomic Energy Research Institute (JAERI) and several industrial firms in Japan. This paper describes the requirements for the next generation of power reactor, the safety design philosphy of ISR and ISER, the controllability of ISER and the results of analyses of some of the design-based accidents (DBA) of ISER, namely station blackout, accidents in which the pressurizer relief valve becomes jammed and stuck in open position and tube breaks in the steam generator. It is concluded that the ISER can ensure a wide range of contraollabitily and fuel integrity for all the analysed DBAs. (orig.)

Coolant clean-up system in nuclear reactor

International Nuclear Information System (INIS)

Tsuburaya, Hirobumi; Akita, Minoru; Shiraishi, Tadashi; Kinoshita, Shoichiro; Okura, Minoru; Tsuji, Akio.

1987-01-01

Purpose: To ensure a sufficient urging pressure at the inlet of a coolant clean-up system pump in a nuclear reactor and eliminate radioactive contaminations to the pump. Constitution: Coolant clean-up system (CUW) pump in a nuclear reactor is disposed to the downstream of a filtration desalter and, for compensating the insufficiency of the urging pressure at the pump inlet, the reactor water intake port to the clean-up system is disposed to the downstream of the after-heat removing pump and the heat exchanger. By compensating the net positive suction head (NPSH) of the clean-up system from the residual heat removing system, the problems of insufficient NPSH for the CUW pump upon reactor shut-down can be dissolved and, accordingly, the reactor clean-up system can be arranged in the order of the heat exchanger, clean-up device and pump. Thus, the CUW pump acts on reactor water after cleaned-up in the clean-up device to reduce the radioactivity contamination to the pump. (Kawakami, Y.)

Limit regulation system for pressurized water nuclear reactors

International Nuclear Information System (INIS)

Aleite, W.; Bock, H.W.

1976-01-01

Described is a limit regulation system for a pressurized water nuclear reactor in combination with a steam generating system connected to a turbine, the nuclear reactor having control rods as well as an operational regulation system and a protective system, which includes reactor power limiting means operatively associated with the control rods for positioning the same and having response values between operating ranges of the operational regulation system, on the one hand, and response values of the protective system, on the other hand, and a live steam-minimal pressure regulation system cooperating with the reactor power limiting means and operatively connected to a steam inlet valve to the turbine for controlling the same

Cooperation in reactor design evaluation and licensing

International Nuclear Information System (INIS)

Kaufer, B.; Wasylyk, A.

2014-01-01

In January 2007 the World Nuclear Association (WNA) established the Cooperation in Reactor Design Evaluation and Licensing (CORDEL) Working Group with the aim of stimulating a dialogue between the nuclear industry (including reactor vendors, operators and utilities) and nuclear regulators (national and international organisations) on the benefits and means of achieving a worldwide convergence of reactor safety standards for reactor designs. From the time of its inception to the present, CORDEL has evolved from a group of experts discussing how to achieve international standardisation in nuclear safety design to an established and recognised working group dedicated to analysing and forging common understandings in key areas as input to major decisions on nuclear energy policy. This paper will review the general directions and activities CORDEL plans to undertake during the next five-year period, including its general strategy, activities, priorities and interactions with its customers in order to meet its objectives. (author)

Cooperation in reactor design evaluation and licensing

Energy Technology Data Exchange (ETDEWEB)

Kaufer, B.; Wasylyk, A. [World Nuclear Association, London (United Kingdom)

2014-07-01

In January 2007 the World Nuclear Association (WNA) established the Cooperation in Reactor Design Evaluation and Licensing (CORDEL) Working Group with the aim of stimulating a dialogue between the nuclear industry (including reactor vendors, operators and utilities) and nuclear regulators (national and international organisations) on the benefits and means of achieving a worldwide convergence of reactor safety standards for reactor designs. From the time of its inception to the present, CORDEL has evolved from a group of experts discussing how to achieve international standardisation in nuclear safety design to an established and recognised working group dedicated to analysing and forging common understandings in key areas as input to major decisions on nuclear energy policy. This paper will review the general directions and activities CORDEL plans to undertake during the next five-year period, including its general strategy, activities, priorities and interactions with its customers in order to meet its objectives. (author)

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

Directory of Open Access Journals (Sweden)

Imam Mahmoud M.

2002-01-01

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

Sodium fires at fast reactors: RF status report

International Nuclear Information System (INIS)

Bagdasarov, Yu.E.; Buksha, Yu.K.; Drobyshev, A.V.; Zybin, V.A.; Ivanenko, V.N.; Kardash, D.Yu.; Kulikov, E.V.; Yagodkin, I.V.

1996-01-01

Scientific and engineering studies carried out in Russian Federation since 1992 up to 1996 in the sodium fire area and their main results are described. A review of activities on modification of the computer codes BOX and AERO developed at IPPE for calculating sodium fire consequences is given. Results of analysis of possible accidental situations at currently designed BN-800 reactor NPP with the use of these codes are presented. Sodium leaks occurring at our domestic fast reactors are briefly analyzed. Experimental work performed are described. Results of comparative analysis of common-cause and sodium fire hazards for fast reactor NPP are presented. (author)

Water cooled reactor technology: Safety research abstracts no. 1

International Nuclear Information System (INIS)

1990-01-01

The Commission of the European Communities, the International Atomic Energy Agency and the Nuclear Energy Agency of the OECD publish these Nuclear Safety Research Abstracts within the framework of their efforts to enhance the safety of nuclear power plants and to promote the exchange of research information. The abstracts are of nuclear safety related research projects for: pressurized light water cooled and moderated reactors (PWRs); boiling light water cooled and moderated reactors (BWRs); light water cooled and graphite moderated reactors (LWGRs); pressurized heavy water cooled and moderated reactors (PHWRs); gas cooled graphite moderated reactors (GCRs). Abstracts of nuclear safety research projects for fast breeder reactors are published independently by the Nuclear Energy Agency of the OECD and are not included in this joint publication. The intention of the collaborating international organizations is to publish such a document biannually. Work has been undertaken to develop a common computerized system with on-line access to the stored information

Fuel handling grapple for nuclear reactor plants

International Nuclear Information System (INIS)

Rousar, D.L.

1992-01-01

This patent describes a fuel handling system for nuclear reactor plants. It comprises: a reactor vessel having an openable top and removable cover and containing therein, submerged in water substantially filling the reactor vessel, a fuel core including a multiplicity of fuel bundles formed of groups of sealed tube elements enclosing fissionable fuel assembled into units, the fuel handling system consisting essentially of the combination of: a fuel bundle handling platform movable over the open top of the reactor vessel; a fuel bundle handling mast extendable downward from the platform with a lower end projecting into the open top reactor vessel to the fuel core submerged in water; a grapple head mounted on the lower end of the mast provided with grapple means comprising complementary hooks which pivot inward toward each other to securely grasp a bail handle of a nuclear reactor fuel bundle and pivot backward away from each other to release a bail handle; the grapple means having a hollow cylindrical support shaft fixed within the grapple head with hollow cylindrical sleeves rotatably mounted and fixed in longitudinal axial position on the support shaft and each sleeve having complementary hooks secured thereto whereby each hook pivots with the rotation of the sleeve secured thereto; and the hollow cylindrical support shaft being provided with complementary orifices on opposite sides of its hollow cylindrical and intermediate to the sleeves mounted thereon whereby the orifices on both sides of the hollow cylindrical support shaft are vertically aligned providing a direct in-line optical viewing path downward there-through and a remote operator positioned above the grapple means can observe from overhead the area immediately below the grapple hooks

Reactor container

International Nuclear Information System (INIS)

Hidaka, Masataka; Hatamiya, Shigeo; Kawasaki, Terufumi; Fukui, Toru; Suzuki, Hiroaki; Kataoka, Yoshiyuki; Kawabe, Ryuhei; Murase, Michio; Naito, Masanori.

1990-01-01

In order to suppress the pressure elevation in a reactor container due to high temperature and high pressure steams jetted out upon pipeway rupture accidents in the reactor container, the steams are introduced to a pressure suppression chamber for condensating them in stored coolants. However, the ability for suppressing the pressure elevation and steam coagulation are deteriorated due to the presence of inactive incondensible gases. Then, there are disposed a vent channel for introducing the steams in a dry well to a pressure suppression chamber in the reactor pressure vessel, a closed space disposed at the position lower than a usual liquid level, a first channel having an inlet in the pressure suppression chamber and an exit in the closed space and a second means connected by way of a backflow checking means for preventing the flow directing to the closed space. The first paths are present by plurality, a portion of which constitutes a syphon. The incondensible gases and the steams are discharged to the dry well at high pressure by using the difference of the water head for a long cooling time after the pipeway rupture accident. Then, safety can be improved without using dynamic equipments as driving source. (N.H.)

A worldwide survey of fast breeder reactors

International Nuclear Information System (INIS)

Hennies, H.H.

1986-01-01

While the completion of the SNR 300 was accompanied by manifold discussions on questions relevant to safety and energy policies in the Federal Republic of Germany and as a result considerable scheduling delays and exceeding of budgets were recorded, breeder reactor technology has been progressing worldwide. The transition from the development phase with small trial reactors to the construction and operation of large performance reactors was completed systematically, in particular in France and the Soviet Union. Even though the uranium supply situation does not make a short-term and comprehensive employment of fast breeder reactors essential, technology has meanwhile been advanced to such a level and extensive operating experience is on hand to enable the construction and safe operation of fast breeder reactors. A positive answer has long been found to the question of the realization of a breeding rate to guarantee the breeding effect. There remain now the endeavors to achieve a reduction in investment and fuel cycle costs. (orig.) [de

Conceptual design of fusion experimental reactor (FER)

International Nuclear Information System (INIS)

1984-03-01

A conceptual design study (option C) has been carried out for the fusion experimental reactor (FER). In addition to design of the tokamak reactor and associated systems based on the reference design specifications, feasibility of a water-shield reactor concept was examined as a topical study. The design study for the reference tokamak reactor has produced a reactor concept for the FER, along with major R D items for the concept, based on close examinations on thermal design, electromagnetics, neutronics and remote maintenance. Particular efforts have been directed to the area of electromagnetics. Detailed analyses with close simulation models have been performed on PF coil arrangements and configurations, shell effects of the blanket for plasma position unstability, feedback control, and eddy currents during disruptions. The major design specifications are as follows; Peak fusion power 437 MW Major radius 5.5 m Minor radius 1.1 m Plasma elongation 1.5 Plasma current 5.3 MA Toroidal beta 4 % Field on axis 5.7 T (author)

Factors affecting the minimum capital cost of a tokamak reactor

International Nuclear Information System (INIS)

Hancox, R.

1981-01-01

The Mk IIA Culham conceptual tokamak reactor design is a 2500 MWe steady-state reactor developed on the basis of a cost optimisation. A revised 1200 MWe conceptual design, the Mk IIB, used a lower wall loading and lower thermodynamic efficiency. A detailed costing of the Mk IIB design, however, showed it to have an unacceptably high capital cost. Since this high cost is a common characteristic of many fusion reactor designs, the cost optimisation of the Mk II design has been reconsidered. (author)

Reactor shutdown device

International Nuclear Information System (INIS)

Ito, Masahiko

1990-01-01

The object of the present invention is to reliably shutdown an LMFBR type reactor upon accident of the reactor. That is, curie point magnetic member is made annular so that it can be moved between the outer circumference of an electromagnet and the position above the electromagnet. This enables to enlarge the curie point magnetic member since it is no more necessary to be inserted it in a guide tube. Accordingly, attracting force upon normal operation is increased to remarkably improve the reliability against erronerous scram, etc. Further, since a required gap is formed between the curie point magnetic member and the electromagnet and the heat of coolants is efficiently transmitted to the curie point magnetic member, rapid scram is possible. Further, a position support mechanism is disposed to a part of a control element or at the inner side of the guiding tube for urging and actuating the armature to make it protrude above the top of the guiding tube. With such a constitution, since the armature can be adsorbed without inserting the curie point magnetic member and the electromagnet guide tube, the same effect as in the case of inserting them can be obtained. (I.S.)

Breeding blankets for thermonuclear reactors

International Nuclear Information System (INIS)

Rocaboy, Alain.

1982-06-01

Materials with structures suitable for this purpose are studied. A bibliographic review of the main solid and liquid lithiated compounds is then presented. Erosion, dimensioning and maintenance problems associated with the limiter and the first wall of the reactor are studied from the point of view of the constraints they impose on the design of the blankets. Detailed studies of the main solid and liquid blanket concepts enable the best technological compromises to be determined for the indispensable functions of the blanket to be assured under acceptable conditions. Our analysis leads to four classes of solution, which cannot at this stage be considered as final recommendations, but which indicate what sort of solutions it is worthwhile exploring and comparing in order to be in a position to suggest a realistic blanket at the time when plasma control is sufficiently good for power reactors to be envisaged. Some considerations on the general architecture of the reactor are indicated. Energy storage with pulsed reactors is discussed in the appendix, and a first approach made to minimizing the total tritium recovery [fr

Modernization of reactor instrumentation for research reactors at Trombay

International Nuclear Information System (INIS)

Darbhe, M.D.; Chaudhuri, H.

1989-01-01

The three research reactors at Trombay, viz., Apsara, Cirus and Zerlina were commissioned in 1956, 1960 and 1961 respectively. The nuclear instrumentation designs were based on the vacuum tube technology, which was prevalent during those days. The effect of component obsolescence of critical components like vacuum tubes, magnetic amplifiers and sensitrol meter relays was strongly felt since early 1970s. Also, the failure rates of the units were observed to show an increasing trend due to ageing and lack of good quality indigenous spares. Hence it was proposed to replace the nuclear instrumentation units for the three reactors, with those employing modern, state of the art solid state devices, keeping indigenous content as high as practicable. The work started in 1977 with the preparations of specifications and the project was scheduled to be completed in 1981. The project was divided into two phases. The Phase I comprising of nuclear channels common to all reactors and Phase II consisting exclusively of regulating system units of Cirus. The salient stages of project progress and completion were: (i) Fabrication and testing of final design prototypes was completed by end of 1982. (ii) Commissioning of new units at Apsara was completed in January 1984. (iii) Commissioning of new units at Cirus was completed in September 1984. An account of experience in all these stages and problems encountered is given. (author). 6 figs

Time changes of vertical profile of neutron fluence rate in LVR-15 reactor

International Nuclear Information System (INIS)

Viererbl, L.; Stehno, J.; Erben, O.; Lahodova, Z.; Marek, M.

2003-01-01

The LVR-15 reactor is a light water research type reactor, which is situated, in Nuclear Research Institute, Rez near Prague. The reactor is used as a multipurpose facility. For some experiments and material productions, e.g. for homogeneity of silicon resistance in production of radiation doped silicon, the time changes of vertical profile of neutron fluence rate are particularly important. The assembly used for silicon irradiation has two self-powered neutron detectors installed in a vertical irradiation channel in LVR-15 reactor. Vertical profile of thermal neutron fluence rate was automatically scanned during reactor operation. The results of measurements made in 2002 and 2003 with these detectors are presented. A set of vertical profile measurements was made during two 21-days reactor cycles. During the cycle the vertical profile slightly changes both in the position of its maximum and in the shape. The time dependences of the position of profile maximum and the profile width at half maximum during the cycle are given. (author)

The evolution of Canadian research reactors, 1942 to 1992

International Nuclear Information System (INIS)

Lidstone, R.F.

1993-01-01

This report reviews the evolution of Canadian research reactor design over the past fifty years. Canadian research reactors have played a central role in the development of CANDU power reactors and the creation of an international business based on the production of medical and industrial radioisotopes. A focus on neutronic efficiency, which stemmed from the incentive for direct use of Canada's abundant uranium resources, has resulted in several generations of multipurpose reactors with a common heritage of neutronic efficiency. Each successive nuclear design has been engineered with the best available resources to extrapolate slightly from the current technology base to meet specific program requirements and to incorporate the flexibility needed to accommodate evolving priorities. (Author) 19 refs., 2 tabs., 2 figs

Investigation reactor D-2201 polypropylene production unit using nuclear technique

International Nuclear Information System (INIS)

Wibisono; Sugiharto; Jefri Simanjuntak

2016-01-01

D-2201 reactor is a unit in the polypropylene production process at Pertamina Refinery Unit III Plaju. Reactor with a capacity of 45 kilo liter is not operated in normal operation condition. The validity of liquid level indicator on the unit is doubtful when refers to the production quality. Gamma source of 150 mCi Cobalt-60 and a scintillation detector had been used to scan the outer wall of the reactor to detect the liquid level during operation with a capacity of 40 %. Measurements were made along the reactor walls with 25 mm scan resolution and 5 seconds time sampling. Experiment result shows that the liquid level at the position of 40 % and at normal level position are not observed. Investigation did not find the liquid level above normal. D-2201 is diagnose not normal operating condition diagnosed with liquid abundant passed the recommended limits. Investigation advised to repair or to calibrate the liquid level indicator which is currently installed. (author)

Remaining life diagnosis method and device for nuclear reactor

International Nuclear Information System (INIS)

Yamamoto, Michiyoshi.

1996-01-01

A neutron flux measuring means is inserted from the outside of a reactor pressure vessel during reactor operation to forecast neutron-degradation of materials of incore structural components in the vicinity of portions to be measured based on the measured values, and the remaining life of the reactor is diagnosed by the forecast degraded state. In this case, the neutron fluxes to be measured are desirably fast and/or medium neutron fluxes. As the positions where the measuring means is to be inserted, for example, the vicinity of the structural components at the periphery of the fuel assembly is selected. Aging degradation characteristics of the structural components are determined by using the aging degradation data for the structural materials. The remaining life is analyzed based on obtained aging degradation characteristics and stress evaluation data of the incore structural components at portions to be measured. Neutron irradiation amount of structural components at predetermined positions can be recognized accurately, and appropriate countermeasures can be taken depending on the forecast remaining life thereby enabling to improve the reliability of the reactor. (N.H.)

A reactor core with accurately positioned fuel-batteries

International Nuclear Information System (INIS)

Borrman, B.E.

1976-01-01

A reactor core of containing a grid for a plurality of fuel batteries each of which is constituted by several claddings containing fuel-rods, said grid comprising square members mainly and being located at the core upper-end, each square member surrounding a group of four fuel batteries, spring-contacts being mounted between the fuel batteries and the grid, slots being provided between the batteries for the four arms of a centrally mounted cross-slaped control-rod, each slot being provided at the grid-level, with a flexible spacing device, the overall spacing of whork determining the (a+2b)- dimension is equal to, or higher than, the largest thickness of arm D of the above-mentioned control-rod, said spacing device constituting one of the control-rails the fuel batteries fixed to the fuel-element envelope, as well as the control-rails fixed to the grid, characterized in that each battery control-rail forms a closing surface at right angles to the wall of the adjacent battery and directed toward the grid nearest surface in contact with the above-mentioned control-rail. (author)

Reactor fueling system

International Nuclear Information System (INIS)

Hattori, Noriaki; Hirano, Haruyoshi.

1983-01-01

Purpose: To optimally position a fuel catcher by mounting a television camera to a fuel catching portion and judging video images by the use of a computer or the like. Constitution: A television camera is mounted to the lower end of a fuel catching mechanism for handling nuclear fuels and a fuel assembly disposed within a reactor core or a fuel storage pool is observed directly from above to judge the position for the fuel assembly by means of video signals. Then, the relative deviation between the actual position of the fuel catcher and that set in a memory device is determined and the positional correction is carried out automatically so as to reduce the determined deviation to zero. This enables to catch the fuel assembly without failure and improves the efficiency for the fuel exchange operation. (Moriyama, K.)

Nuclear reactor physics course for reactor operators

International Nuclear Information System (INIS)

Baeten, P.

2006-01-01

The education and training of nuclear reactor operators is important to guarantee the safe operation of present and future nuclear reactors. Therefore, a course on basic 'Nuclear reactor physics' in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The aim of the basic course on 'Nuclear Reactor Physics for reactor operators' is to provide the reactor operators with a basic understanding of the main concepts relevant to nuclear reactors. Seen the education level of the participants, mathematical derivations are simplified and reduced to a minimum, but not completely eliminated

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.)

Ageing problems and renovation programme of ET-RR-1 reactor

International Nuclear Information System (INIS)

Khattab, M.S.; Sultan, M.A.

1995-01-01

Based on Practical Experience gained from interfacing ageing systems in addition to operating new systems, current problems could be deduced whenever in-service inspection are carried out. This paper summarizes the in-service inspection made, and the proposed programme of rehabilitation of mechanical system in the ET-RR-1 research reactor at Inshass. Exchangeable experience in solving common problems in similar reactors play an important role in the effectiveness of such rehabilitation programme. The paper summarizes also the modernization of control, measuring and radiation monitoring system already carried out at the reactor. (orig.)

Isothermal temperature reactivity coefficient measurement in TRIGA reactor

International Nuclear Information System (INIS)

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

2002-01-01

Direct measurement of an isothermal temperature reactivity coefficient at room temperatures in TRIGA Mark II research reactor at Jozef Stefan Institute in Ljubljana is presented. Temperature reactivity coefficient was measured in the temperature range between 15 o C and 25 o C. All reactivity measurements were performed at almost zero reactor power to reduce or completely eliminate nuclear heating. Slow and steady temperature decrease was controlled using the reactor tank cooling system. In this way the temperatures of fuel, of moderator and of coolant were kept in equilibrium throughout the measurements. It was found out that TRIGA reactor core loaded with standard fuel elements with stainless steel cladding has small positive isothermal temperature reactivity coefficient in this temperature range.(author)

Assessment of benefits of research reactors in less developed countries. A case study of the Dalat reactor in Vietnam

International Nuclear Information System (INIS)

Hien, P.D.

1999-01-01

The analysis of data on nuclear research reactor (NRR) and socio-economic conditions across countries reveals highly significant relationships of reactor power with GDP and R and D expenditure. The trends revealed can be used as preliminary guides for feasibility assessment of investment in a NRR. Concerning reactor performance, i.e. the number of reactor operation days per year, the covariation with R and D expenditure is most significant, but moderate, implying that there are other controlling factors, e.g. the engagement of country in nuclear power development. Thus, the size of the R and D fund is a most significant indicator to look at in reactor planning. Unfortunately, the lack of adequate R and D funding is a common and chronic problem in less developed countries. As NRR is among the biggest R and D investment in less developed countries, adequate cost benefit assessment is rightfully required. In the case of Vietnam, during 15 years of operation of a 500 kW NRR 2300 Ci of radioisotopes were delivered and 45,000 samples were analysed for multielemental compositions. From a pure financial viewpoint these figures would still be insignificant to justify the investment. However, the impact of the reactor on the technological development seems not to be a matter of pro and cons. The status of reactor utilization and lessons learned are presented and discussed. (author)

Assessment of benefits of research reactors in less developed countries. A case study of the Dalat reactor in Vietnam

Energy Technology Data Exchange (ETDEWEB)

Hien, P.D. [Vietnam Atomic Energy Agency, Hanoi (Viet Nam)

1999-08-01

The analysis of data on nuclear research reactor (NRR) and socio-economic conditions across countries reveals highly significant relationships of reactor power with GDP and R and D expenditure. The trends revealed can be used as preliminary guides for feasibility assessment of investment in a NRR. Concerning reactor performance, i.e. the number of reactor operation days per year, the covariation with R and D expenditure is most significant, but moderate, implying that there are other controlling factors, e.g. the engagement of country in nuclear power development. Thus, the size of the R and D fund is a most significant indicator to look at in reactor planning. Unfortunately, the lack of adequate R and D funding is a common and chronic problem in less developed countries. As NRR is among the biggest R and D investment in less developed countries, adequate cost benefit assessment is rightfully required. In the case of Vietnam, during 15 years of operation of a 500 kW NRR 2300 Ci of radioisotopes were delivered and 45,000 samples were analysed for multielemental compositions. From a pure financial viewpoint these figures would still be insignificant to justify the investment. However, the impact of the reactor on the technological development seems not to be a matter of pro and cons. The status of reactor utilization and lessons learned are presented and discussed. (author)

Evaluation of neutron exposure conditions for the Buffalo Reactor

International Nuclear Information System (INIS)

Lippincott, E.P.; Kellogg, L.S.; McElroy, W.N.; Baldwin, C.A.

1984-04-01

The light water test reactor at the Nuclear Science and Technology Facility of the State University of New York at Buffalo is currently being used to irradiate specimens in in-core positions for NRC-sponsored metallurgical tests. It is important that the neutron exposures for these Buffalo tests be consistent with those determined for related irradiations in the BSR and ORR reactor at ORNL. Therefore, HEDL National Reactor Dosimetry Center dosimetry procedures and ORNL calculational procedures were used for an evaluation of typical test conditions

Nuclear reactors

International Nuclear Information System (INIS)

Barre, Bertrand

2015-10-01

After some remarks on the nuclear fuel, on the chain reaction control, on fuel loading and unloading, this article proposes descriptions of the design, principles and operations of different types of nuclear reactors as well as comments on their presence and use in different countries: pressurized water reactors (design of the primary and secondary circuits, volume and chemistry control, backup injection circuits), boiling water reactors, heavy water reactors, graphite and boiling water reactors, graphite-gas reactors, fast breeder reactors, and fourth generation reactors (definition, fast breeding). For these last ones, six concepts are presented: sodium-cooled fast reactor, lead-cooled fast reactor, gas-cooled fast reactor, high temperature gas-cooled reactor, supercritical water-cooled reactor, and molten salt reactor

Application of Cherenkov light observation to reactor measurements (1). Estimation of reactor power from Cherenkov light intensity

International Nuclear Information System (INIS)

Yamamoto, Keiichi; Takeuchi, Tomoaki; Kimura, Nobuaki; Ohtsuka, Noriaki; Tsuchiya, Kunihiko; Sano, Tadafumi; Nakajima, Ken; Homma, Ryohei; Kosuge, Fumiaki

2015-01-01

Development of the reactor measurement system was started to obtain the real-time in-core nuclear and thermal information, where the quantitative measurement of brightness of Cherenkov light was investigated. The system would be applied as a monitoring system in severe accidents and for the advanced operation management technology in existing LWRs. The calculation and the observation were performed to obtain the quantity of the Cherenkov light caused by the gamma and beta rays emitted from the fuels in the core of Kyoto University Research Reactor. The results indicate that the real-time reactor power can be estimated from the brightness of the Cherenkov light observed by a CCD camera. This method can also work for the estimation of the burn-up of spent fuels at commercial reactors. Since the observed brightness value of the Cherenkov light was influenced by the camera position, the optical observation method should be improved to achieve high accuracy observation. (author)

Physics design of fast reactor safety test facilities for in-pile experiments

International Nuclear Information System (INIS)

Travelli, A.; Matos, J.E.; Snelgrove, J.L.; Shaftman, D.H.; Tzanos, C.P.; Lam, S.K.; Pennington, E.M.; Woodruff, W.L.

1976-01-01

A determined effort to identify and resolve current Fast Breeder Reactor safety testing needs has recently resulted in a number of conceptual designs for FBR safety test facilities which are very complex and diverse both in their features and in their purpose. The paper discusses the physics foundations common to most fast reactor safety test facilities and the constraints which they impose on the design. The logical evolution, features, and capabilities of several major conceptual designs are discussed on the basis of this common background

Radiation detector for use in nuclear reactors

International Nuclear Information System (INIS)

Cisco, T.C.; Grimaila, A.G.

1981-01-01

A multi-sensor radiation detection system for removable insertion into a nuclear reactor is described in which one conductor of all the sensors is a single, common element. This single common element is contained within a tubular metallic sheath and in crosssection comprises a multiple radial armed metallic conductor having a star shaped cross-section dimensioned to form wedgeshaped compartments throughout the active radiation detecting length of the metallic sheath

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)

Results of theoretical studies on power noise of PWR type reactors

International Nuclear Information System (INIS)

Meyer, K.

1979-12-01

For evaluating noise measurements of neutron flux in power reactors position-dependent effects are taken into account. Based on an effective one-group diffusion model results are obtained for reactor noise due to vibrating control elements and disturbances in the coolant flow

PERIGEE computer codes for reactor simulation in 3 dimensions, using 1 or 2 neutron velocity groups

International Nuclear Information System (INIS)

Olson, A.P.

1964-02-01

PERIGEE is a code written in SNAP for the G-20 computer. It solves the one- or two-group neutron diffusion equations by finite-difference methods on a three-dimensional, uniform mesh having a common spacing in the two directions normal to the fuel channels. The positions of mesh points along a fuel channel, relative to points in adjacent channels, may correspond to either NPD or CANDU fuel bundle positions. The extrapolated flux boundary may be specified in sufficient detail to represent a tapered or stepped circumferential reflector, a variable axial length and, for a reactor with axis horizontal, a variable moderator level and a variable plane bottom surface equivalent to the CANDU dump structure. The neutron flux may be normalized to give a specified power output from the hottest fuel bundle or hottest channel, or to give a total thermal power limited by the turbine and generator. Reactor operation may be simulated in finite time steps, taking into account any fuel shifts, any changes in moderator level and the change in nuclear properties of the fuel with increasing irradiation. The appropriate properties are obtained by interpolation from tables supplied for as many as 8 types of fuel bundle. The mean fuel exit burnup can be calculated at equilibrium for a reactor in which the exit burnups for two zones may be adjusted to give radial power flattening and the fuelling schedules may be designed to give axial power flattening in one or both zones. (author)

Experiment calculated ascertainment of factors affecting the energy release in IGR reactor core

International Nuclear Information System (INIS)

Kurpesheva, A.M.; Zhotabayev, Zh.R.

2006-01-01

Full text: At present energy supply resources problem is important. Nuclear reactors can, of course, solve this problem, but at the same time there is another issue, concerning safety exploitation of nuclear reactors. That is why, for the last seven years, such experiments as 'Investigation of the processes, conducting severe accidents with core melting' are being carried out at our IGR (impulse graphite reactor) reactor. Leaving out other difficulties of such experiments, it is necessary to notice, that such experiments require more accurate IGR core energy release calculations. The final aim of the present research is verification and correction of the existing method or creation of new method of IGR core energy release calculation. IGR reactor is unique and there is no the same reactor in the world. Therefore, application of the other research reactor methods here is quite useful. This work is based on evaluation of factors affecting core energy release (physical weight of experimental device, different configuration of reactor core, i.e. location of absorbers, initial temperature of core, etc), as well as interference of absorbers group. As it is known, energy release is a value of integral reactor power. During experiments with rays, Reactor power depends on currents of ion production chambers (IPC), located round the core. It is worth to notice that each ion production chamber (IPC) in the same start-up has its own ratio coefficient between IPC current and reactor present power. This task is complicated due to 'IPC current - reactor power' ratio coefficients, that change continuously, probably, because of new loading of experimental facility and different position of control rods. That is why, in order to try about reactor power, before every start-up, we have to re-determine the 'IPC current - reactor power' ratio coefficients for each ion production chamber (IPC). Therefore, the present work will investigate the behavior of ratio coefficient within the

Probabilistic evaluation of multiple failures for steam generators tubes by common mode

International Nuclear Information System (INIS)

Bloch, M.; Pierrey, J.L.; Dussarte, D.

1987-11-01

The reactor safety can be affected when systems or components are subject to phenomena conducting at a wear nontake in account in the conception. This paper presents a methodology which takes in account the non simultaneous failures resulting of this situation. To illustrate this purpose, we give an evaluation of risk of multiple failures for steam generators tubes by common mode (stress corrosion) when the reactor is in normal operation [fr

Nuclear reactor core servicing apparatus

International Nuclear Information System (INIS)

Andrea, C.

1977-01-01

Disclosed is an improved core servicing apparatus for a nuclear reactor of the type having a reactor vessel, a vessel head having a head penetration therethrough, a removable plug adapted to fit in the head penetration, and a core of the type having an array of elongated assemblies. The improved core servicing apparatus comprises a plurality of support columns suspended from the removable plug and extending downward toward the nuclear core, rigid support means carried by each of the support columns, and a plurality of servicing means for each of the support columns for servicing a plurality of assemblies. Each of the plurality of servicing means for each of the support columns is fixedly supported in a fixed array from the rigid support means. Means are provided for rotating the rigid support means and servicing means between condensed and expanded positions. When in the condensed position, the rigid support means and servicing means lie completely within the coextensive boundaries of the plug, and when in the expanded position, some of the rigid support means and servicing means lie without the coextensive boundaries of the plug

Applications of the Dow TRIGA research reactor

International Nuclear Information System (INIS)

Kocher, C.W.; Quinn, T.J.; Krueger, D.A.

1982-01-01

The Dow TRIGA Mark I reactor is a one-hundred kilowatt nuclear reactor installed by General Atomics using the Torrey Pines reactor console, seventy-five used stainless-steel clad fuel elements and one new aluminium clad fuel element. The reactor is equipped with a forty-position rotating Lazy Susan in the reflector, a pneumatic transfer system with its terminal in the F-ring of the core, and a central thimble which can be used for irradiation of samples in the center of the core or which can be emptied of the shielding water to produce a beam of neutrons and gamma rays in the area atop the pool. Samples can also be irradiated in or near the core. There is no provision for pulsing this TRIGA reactor. The neutron activation analysis program uses the Dow TRIGA reactor as a source of thermal neutrons and a Kaman A711 generator as a source of 14-MeV neutrons. The associated counting equipment includes one Gel(Li) detector and two Nal(Tl) detectors, each using a 100-position sample changer and all interfaced to a Tracor-Northern TN-11 data acquisition and computing system, one Ge(Li) detector and its TN-11 system for the pneumatic transfer system and the beam tube experiments, and two NaKTl)detectors with a TN-4000 system used with the Kaman neutron generator. The activation analysis program gets samples from all parts of the manufacturing and research efforts at Dow: raw materials, intermediates, products, effluents, research samples, samples from customers who use Dow products, and environmental samples. This presentation is devoted to the progress made in the past year on the pneumatic transfer system and the renewed work on prompt gamma-ray spectroscopy including the extensive process of method validation

Evaluation of common mode failure of safety functions for limiting fault events

International Nuclear Information System (INIS)

Rezendes, J.P.; Hyde, A.W.

2004-01-01

The draft U.S. Nuclear Regulatory Commission (NRC) policy on digital protection system software requires all Advanced Light Water Reactors (ALWRs) to be evaluated assuming a hypothetical common mode failure (CMF) which incapacitates the normal automatic initiation of safety functions. The System 80 + ALWR has been evaluated for such hypothetical conditions. The results show that the diverse automatic and manual protective systems in System 80 + provide ample safety performance margins relative to core coolability, offsite radiological releases. Reactor Coolant System (RCS) pressurization and containment integrity. This deterministic evaluation served to quantify the significant inherent safety margins in the System 80 + Standard Plant design even in the event of this extremely low probability scenario of a common mode failure. (author)

Upgrading and modernization of the high flux reactor Petten

International Nuclear Information System (INIS)

Ahlf, J.

1992-01-01

The High Flux Reactor (HFR) at Petten, The Netherlands, owned by the European Communities and operated by the Netherlands Energy Research Foundation, is a water-cooled and moderated, multipurpose research reactor of the closed-tank in-pool type, operated at 45 MW. Performance upgrading comprised two power increases from 20 MW via 30 MW to 45 MW, providing more and higher rated irradiation positions in the tank. With the replacement of the original reactor vessel the experimental capabilities of the reactor were improved. Better pool side facilities and the introduction of a large cross-section, double, beam tube were implemented. Additional major installation upgrading activities consisted of the replacement of the primary and the pool heat exchangers, replacement of the beryllium reflector elements, extension of the overpower protection systems and upgrading of the nuclear instrumentation as well as the guaranteed power supply. Control room upgrading is in progress. A full new safety analysis, as well as the introduction of a comprehensive Quality Assurance system, are summarized under software upgrading. Continuous modernization and upgrading also takes place of equipment for fuel and structural materials irradiations for fission reactors and future fusion machines. In parallel, all supporting services, as well as the management structure for large irradiation programmes, have been developed. Presently the reactor is operating at about 275 full power days per year with an average utilization of the irradiation positions of 70 to 80%. (orig.)

Fuel exchange device for FBR type reactor

International Nuclear Information System (INIS)

Onuki, Koji.

1993-01-01

The device of the present invention can provide fresh fuels with a rotational angle aligned with the direction in the reactor core, so that the fresh fuels can be inserted being aligned with apertures of the reactor core even if a self orientation mechanism should fail to operate. That is, a rotational angle detection means (1) detects the rotational angle of fresh fuels before insertion to the reactor core. A fuel rotational angle control means (2) controls the rotational angle of the fresh fuels by comparing the detection result of the means (1) and the data for the insertion position of the reactor core. A fuel rotation means (3) compensates the rotational angel of the fresh fuels based on the control signal from the means (2). In this way, when the fresh fuels are inserted to the reactor core, the fresh fuels set at the same angle as that for the aperture of the reactor core. Accordingly, even if the self orientation mechanism should not operate, the fresh fuels can be inserted smoothly. As a result, it is possible to save loss time upon fuel exchange and mitigate operator's burden during operation. (I.S.)

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.)

Making better use of research reactors

International Nuclear Information System (INIS)

1964-01-01

Some 250 research reactors are in operation in the world today, and there are problems in putting them to the most fruitful use. The difficulties - of trained manpower, of auxiliary equipment, of satisfactory research programmes, of co-ordination, between the various disciplines - are common to all users. But as is only to be expected, they press more heavily on the newly-established centres, particularly those in the developing countries which are lacking in long experience in research and usually severely limited as to technical manpower and money. The IAEA has been turning its attention to this question for the past three or four years - ever since, in fact, its early assistance missions and other field operations brought it into close contact with the operations of numerous Member States. The task of providing assistance and advice in this matter is growing. Many centres have been building research reactors under bilateral arrangements; with the completion of their projects this form of aid usually ends, and they look to IAEA for help in operating the reactors. Although some critics consider that difficulties have been caused by premature construction of research reactors, before well-founded programmes of nuclear research had been developed in the countries concerned, several valid motives have led to the establishment of some of these centres at an early stage. A research reactor often provides an effective stimulant for scientific research in the country. It is a remarkably versatile tool for workers in many fields of science and technology. There have been instances where the establishment of a research reactor has had a great impact on the scientific education of a country and has led to a salutary reappraisal and reforms. A reactor is sometimes considered to be a particularly effective means of retaining in the country men trained in the nuclear field. This particular problem is common to most countries. In fact, it is a feature of the present age that

Position display device

International Nuclear Information System (INIS)

Nishizawa, Yukio.

1974-01-01

Object: To provide a device capable of easily and quickly reading mutual mounting relations of control bodies such as control rods mounted on a nuclear reactor and positions to which the control bodies are driven. Structure: A scanning circuit is provided to scan positions of controllably mounted control bodies such as control rods. Values detected by scanning the positions are converted into character signals according to the values and converted into preranked color signals. The character signals and color signals are stored in a memory circuit by synchronous signals in synchronism with the scanning in the scanning circuit. Outputs of the memory circuit are displayed by a display unit such as a color Braun tube in accordance with the synchronous signals to provide color representations according to positions to which control bodies are driven in the same positional relation as the mounting of the control bodies. (Kamimura, M.)

Tests of Neutron Spectrum Calculations with the Help of Foil Measurements in a D{sub 2}O and in an H{sub 2}O-Moderated Reactor and in Reactor Shields of Concrete an Iron

Energy Technology Data Exchange (ETDEWEB)

Nilsson, R; Aalto, E

1964-09-15

Foil measurements covering the fast, epithermal and thermal neutron energy regions have been made in the centre of the Swedish D{sub 2}O-moderated reactor R1, in the pool reactor R2-0, and in different positions in reactor shields of iron, magnetite concrete and ordinary concrete. Neutron spectra have also been calculated for most of these positions, often with the help of a numerical integration of the Boltzmann equation. The measurements and the calculated spectra are presented.

Status of the advanced boiling water reactor and simplified boiling water reactor

International Nuclear Information System (INIS)

Smith, P.F.

1992-01-01

This paper reports that the excess of U.S. electrical generating capacity which has existed for the past 15 years is coming to an end as we enter the 1990s. Environmental and energy security issues associated with fossil fuels are kindling renewed interest in the nuclear option. The importance of these issues are underscored by the National Energy Strategy (NES) which calls for actions which are designed to ensure that the nuclear power option is available to utilities. Utilities, utility associations, and nuclear suppliers, under the leadership of the Nuclear Power Oversight Committee (NPOC), have jointly developed a 14 point strategic plan aimed at establishing a predictable regulatory environment, standardized and pre-licensed Advanced Light Water Reactor (ALWR) nuclear plants, resolving the long-term waste management issue, and other enabling conditions. GE is participating in this national effort and GE's family of advanced nuclear power plants feature two new reactor designs, developed on a common technology base, aimed at providing a new generation of nuclear plants to provide safe, clean, economical electricity to the world's utilities in the 1990s and beyond. Together, the large-size (1300 MWe) Advanced Boiling Water Reactor (ABWR) and the small-size (600 MWe) Simplified Boiling Water Reactor (SBWR) are innovative, near-term candidates for expanding electrical generating capacity in the U.S. and worldwide. Both possess the features necessary to do so safely, reliably, and economically

Nolinear stability analysis of nuclear reactors : expansion methods for stability domains

International Nuclear Information System (INIS)

Yang, Chae Yong

1992-02-01

Two constructive methods for estimating asymptotic stability domains of nonlinear reactor models are developed in this study: an improved Chang and Thorp's method based on expansion of a Lyapunov function and a new method based on expansion of any positive definite function. The methods are established on the concept of stability definitions of Lyapunov itself. The first method provides a sequence of stability regions that eventually approaches the exact stability domain, but requires many expansions in order to obtain the entire stability region because the starting Lyapunov function usually corresponds to a small stability region and because most dynamic systems are stiff. The second method (new method) requires only a positive definite function and thus it is easy to come up with a starting region. From a large starting region, the entire stability region is estimated effectively after sufficient iterations. It is particularly useful for stiff systems. The methods are applied to several nonlinear reactor models known in the literature: one-temperature feedback model, two-temperature feedback model, and xenon dynamics model, and the results are compared. A reactor feedback model for a pressurized water reactor (PWR) considering fuel and moderator temperature effects is developed and the nonlinear stability regions are estimated for the various values of design parameters by using the new method. The steady-state properties of the nonlinear reactor system are analyzed via bifurcation theory. The analysis of nonlinear phenomena is carried out for the various forms of reactivity feedback coefficients that are both temperature- (or power-) independent and dependent. If one of two temperature coefficients is positive, unstable limit cycles or multiplicity of the steady-state solutions appear when the other temperature coefficient exceeds a certain critical value. As an example, even though the fuel temperature coefficient is negative, if the moderator temperature

Structural materials issues for the next generation fission reactors

Science.gov (United States)

Chant, I.; Murty, K. L.

2010-09-01

Generation-IV reactor design concepts envisioned thus far cater to a common goal of providing safer, longer lasting, proliferation-resistant, and economically viable nuclear power plants. The foremost consideration in the successful development and deployment of Gen-W reactor systems is the performance and reliability issues involving structural materials for both in-core and out-of-core applications. The structural materials need to endure much higher temperatures, higher neutron doses, and extremely corrosive environments, which are beyond the experience of the current nuclear power plants. Materials under active consideration for use in different reactor components include various ferritic/martensitic steels, austenitic stainless steels, nickel-base superalloys, ceramics, composites, etc. This article addresses the material requirements for these advanced fission reactor types, specifically addressing structural materials issues depending on the specific application areas.

Power distribution forecasting device for reactors

International Nuclear Information System (INIS)

Tsukii, Makoto

1981-01-01

Purpose: To save expensive calculations on the forecasting of reactor power distribution. Constitution: Core status (CSD) such as entire coolant flow rate, pressures in the reactor, temperatures at the outlet and inlet and positions for control rods are inputted into a power distribution calculation device to calculate the power distribution based on physical models intermittently. Further, present power distribution is calculated based on in-core neutron flux measured values and CSD in a process control computer. Further, the ratio of the calculation results of the latter to those of the former is calculated, stored and inputted into a correction device to correct the forecast power distribution obtained by the power distribution calculation device. This enables to forecast the power distribution with excellent responsivity in the reactor site. (Furukawa, Y.)

Quick release latch for reactor scram

International Nuclear Information System (INIS)

Johnson, M.L.; Shawver, B.M.

1976-01-01

A simple, reliable, and fast-acting means for releasing a control element and allowing it to be inserted rapidly into the core region of a nuclear reactor for scram purposes is described. A latch mechanism grips a coupling head on a nuclear control element to connect the control element to the control drive assembly. The latch mechanism is closed by tensioning a cable or rod with an actuator. The control element is released by de-energizing the actuator, providing fail-safe, rapid release of the control element to effect reactor shutdown. A sensing rod provides indication that the control element is properly positioned in the latch. Two embodiments are illustrated, one involving a collet-type latch mechanism, the other a pliers-type latch mechanism with the actuator located inside the reactor vessel

Quick release latch for reactor scram

International Nuclear Information System (INIS)

Johnson, M.L.; Shawver, B.M.

1975-01-01

A simple, reliable, and fast-acting means for releasing a control element and allowing it to be inserted rapidly into the core region of a nuclear reactor for scram purposes is described. A latch mechanism grips a coupling head on a nuclear control element to connect the control element to the control drive assembly. The latch mechanism is closed by tensioning a cable or rod with an actuator. The control element is released by de-energizing the actuator, providing fail-safe, rapid release of the control element to effect reactor shutdown. A sensing rod provides indication that the control element is properly positioned in the latch. Two embodiments are illustrated, one involving a collet-type latch mechanism, the other a pliers-type latch mechanism with the actuator located inside the reactor vessel

Nuclear reactor melt-retention structure to mitigate direct containment heating

Science.gov (United States)

Tutu, Narinder K.; Ginsberg, Theodore; Klages, John R.

1991-01-01

A light water nuclear reactor melt-retention structure to mitigate the extent of direct containment heating of the reactor containment building. The structure includes a retention chamber for retaining molten core material away from the upper regions of the reactor containment building when a severe accident causes the bottom of the pressure vessel of the reactor to fail and discharge such molten material under high pressure through the reactor cavity into the retention chamber. In combination with the melt-retention chamber there is provided a passageway that includes molten core droplet deflector vanes and has gas vent means in its upper surface, which means are operable to deflect molten core droplets into the retention chamber while allowing high pressure steam and gases to be vented into the upper regions of the containment building. A plurality of platforms are mounted within the passageway and the melt-retention structure to direct the flow of molten core material and help retain it within the melt-retention chamber. In addition, ribs are mounted at spaced positions on the floor of the melt-retention chamber, and grid means are positioned at the entrance side of the retention chamber. The grid means develop gas back pressure that helps separate the molten core droplets from discharged high pressure steam and gases, thereby forcing the steam and gases to vent into the upper regions of the reactor containment building.

Method of operating water cooled reactor with blanket

International Nuclear Information System (INIS)

Suzuki, Katsuo.

1988-01-01

Purpose: To increase the production amount of fissionable plutonium by increasing the burnup degree of blanket fuels in a water cooled reactor with blanket. Method: Incore insertion assemblies comprising water elimination rods, fertile material rods or burnable poison rods are inserted to those fuel assemblies at the central portion of the reactor core that are situated at the positions not inserted with control rods in the earlier half of the operation cycle, while the incore reactor insertion assemblies are withdrawn at the latter half of the operation cycle of a nuclear reactor. As a result, it is possible to increase the power share of the blanket fuels and increase the fuel burnup degree to thereby increase the production amount of fissionable plutonium. Furthermore, at the initial stage of the cycle, the excess reactivity of the reactor can be suppressed to decrease the reactivity control share on the control rod. At the final stage of the cycle, the excess reactivity of the reactor core can be increased to improve the cycle life. (Kamimura, M.)

Development of Bundle Position-Wise Linear Model for Predicting the Pressure Tube Diametral Creep in CANDU Reactors

International Nuclear Information System (INIS)

Lee, Jae Yong; Na, Man Gyun

2011-01-01

Diametral creep of the pressure tube (PT) is one of the principal aging mechanisms governing the heat transfer and hydraulic degradation of a heat transport system. PT diametral creep leads to diametral expansion that affects the thermal hydraulic characteristics of the coolant channels and the critical heat flux. Therefore, it is essential to predict the PT diametral creep in CANDU reactors, which is caused mainly by fast neutron irradiation, reactor coolant temperature and so forth. The currently used PT diametral creep prediction model considers the complex interactions between the effects of temperature and fast neutron flux on the deformation of PT zirconium alloys. The model assumes that long-term steady-state deformation consists of separable, additive components from thermal creep, irradiation creep and irradiation growth. This is a mechanistic model based on measured data. However, this model has high prediction uncertainty. Recently, a statistical error modeling method was developed using plant inspection data from the Bruce B CANDU reactor. The aim of this study was to develop a bundle position-wise linear model (BPLM) to predict PT diametral creep employing previously measured PT diameters and HTS operating conditions. There are twelve bundles in a fuel channel and for each bundle, a linear model was developed by using the dependent variables, such as the fast neutron fluxes and the bundle temperatures. The training data set was selected using the subtractive clustering method. The data of 39 channels that consist of 80 percent of a total of 49 measured channels from Units 2, 3 and 4 were used to develop the BPLM models. The remaining 10 channels' data were used to test the developed BPLM models. The BPLM was optimized by the maximum likelihood estimation method. The developed BPLM to predict PT diametral creep was verified using the operating data gathered from the Units 2,3 and 4 in Korea. Two error components for the BPLM, which are the epistemic

Temperature monitoring of gas-cooled reactors

International Nuclear Information System (INIS)

Kaiser, G.E.

1977-01-01

The present paper deals with questions like : a) Why temperature monitoring in high-temperature reactors at all. b) How are the measuring positions arranged and how are the measurements designed. c) What technique of temperature measurement is applied. (RW) [de

Design Status and Applications of Small reactors without On-site Refuelling

International Nuclear Information System (INIS)

Kuznetsov, V.

2006-01-01

Small reactors without on-site refuelling are the reactors that can operate without reloading and shuffling of fuel for a reasonably long period, consistent with plant economy and considerations of energy security, with no fresh or spent fuel being stored at a site during reactor operation. Such reactors could simplify the implementation of safeguards and provide certain guarantees of sovereignty to those countries that would prefer to lease fuel from a foreign vendor or, perhaps, an international fuel cycle centre. About 30 concepts of such reactors are being analyzed or developed in 6 IAEA Member States. They cover all principle reactor lines: water cooled, fast gas cooled, sodium cooled, lead or lead bismuth cooled and molten salt cooled reactors. An increased refuelling interval could be achieved with reduced core power density, burnable absorbers, or high conversion ratio. The design goals for small reactors without on-site refuelling, inter alia, include: difficult unauthorized access to fuel; design provisions to facilitate the implementation of safeguards; capability to survive all postulated accident scenarios without requiring emergency response in the public domain; economic competitiveness for anticipated market conditions and applications; the capability to achieve higher manufacturing quality through factory mass production, design standardization and common basis for design certification; and a flexibility in siting and applications. Such reactors are often considered in conjunction with fuel or NPP leasing Small reactors without on-site refuelling have many common technology development issues related to the provision of lifetime core operation, economic competitiveness, high level of safety and proliferation resistance. Reestablishment of a practice of licensing by test and establishment of legal provisions and the insurance scheme for a transit of fuel loads or factory fabricated reactors through the territory of a third country are mentioned as

Transient thermal characteristics of a core channel in a molten salt reactor

International Nuclear Information System (INIS)

Sakashita, H.; Ishiguro, R.; Sugiyama, K.

1987-01-01

The present paper deals with the thermal characteristics of Molten Salt Reactor (MSR). Analyses of the fundamental behavior of internal heat generating fluid and graphite contiguous to the fluid are performed. As a result, it is known that the transient thermal characteristics of MSR differ fundamentally from those of a solid-fuel reactor, and the simplified method of thermal analysis which is commonly used for solid-fuel reactors gives optimistic predictions than the actual phenomena. (author)

Radionuclide trap for liquid metal cooled reactors

International Nuclear Information System (INIS)

McGuire, J.C.; Brehm, W.F.

1978-10-01

At liquid metal cooled reactor operating temperatures, radioactive corrosion product transport and deposition in the primary system will be sufficiently high to limit access time for maintenance of system components. A radionuclide trap has been developed to aid in controlling radioactivity transport. This is a device which is located above the reactor core and which acts as a getter, physically immobilizing radioactive corrosion products, particularly 54 Mn. Nickel is the getter material used. It is most effective at temperatures above 450 0 C and effectiveness increases with increasing temperature. Prototype traps have been tested in sodium loops for 40,000 hours at reactor primary temperatures and sodium velocities. Several possible in-reactor trap sites were considered but a location within the top of each driver assembly was chosen as the most convenient and effective. In this position the trap is changed each time fuel is changed

The reactor trainer: state-of-the-art classroom learning

International Nuclear Information System (INIS)

Stater, R.G.

1996-01-01

The Reactor Trainer is a professional, PC based, graphically enhanced, training resource specifically developed and customized for Class Room teaching of and learning about, reactor behavior. This unique, and focused, learning-target sets The Trainer apart from the panorama of the more common PC plant simulator. Its educational scope extends along a logical learning path, starting with important fundamental behavioral concepts of delayed neutrons, neutron multiplying factors, and reactor rate, moving to simple reactor transients in real time, and culminating with more complex operational evolutions. The Trainer empowers the Instructor with a dynamic Class Room demonstrator and the student with a superior hands-on learning tool. The Trainer's versatility encompasses a wide variety of educational needs, including initial operator training, requalification training, Shift Technical Advisor training, and other advanced or specialized training. In addition, The Reactor Trainer enhances prerequisite preparation of operator candidates for full-scale control room training and, in so doing, PC economics relieves full-scale simulator hours. (author)

Evaluation of the impact of a committed site on fusion reactor development

International Nuclear Information System (INIS)

Reid, R.L.; Nagy, A.

1979-01-01

The technical and economic merits of a committed fusion site for development of tokamak, mirror, and EBT reactor from ignition through demo phases were evaluated. Schedule compression resulting from evolving several reactor concepts and/or phases on a committed site as opposed to sequential use of independent sites was estimated. Land, water, and electrical power requirements for a committed fusion site were determined. A conceptual plot plan for siting three fusion reactors on a committed site was configured. Reactor support equipment common to the various concepts was identified as candidates for sharing. Licensing issues for fusion plants were briefly addressed

Description of the Triton reactor

International Nuclear Information System (INIS)

1967-09-01

The Triton reactor is an enriched uranium pool type reactor. It began operation in 1959, after a divergence made on the June 30 the same year. Devoted to studies of radiation protection, its core can be displaced in the longitudinal direction. The pool can be separated in two unequal compartments by a wall. The Triton core is placed in a small compartment, the Nereide core in the big compartment. A third compartment without water is called Naiade II, is separated by a concrete wall in which is made a window closed by an aluminium plate (2.50 m x 2.70 m). The Naiade II hole is useful for protection experiments using the Nereide core. After a complete refitting, the power of the triton reactor that reached progressively from 1.2 MW to 2 MW, then 3 MW has reached in August 1965 6.5 MW. The reactor has been specialized in irradiations in fix position, the core become fix, the nereide core has been hung mobile. Since it has been used for structure materials irradiation, for radioelements fabrication and fundamental research. The following descriptions are valid for the period after August 1965 [fr

Nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

This patent describes an improved liquid metal nuclear reactor construction comprising: (a) a nuclear reactor core having a bottom platform support structure; (b) a reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core; (c) a containment structure surrounding the reactor vessel and having a sidewall spaced outwardly from the reactor vessel side wall and having a base mat spaced below the reactor vessel bottom end wall; (d) a central small diameter post anchored to the containment structure base mat and extending upwardly to the reactor vessel to axially fix the bottom end wall of the reactor vessel and provide a center column support for the lower end of the reactor core; (e) annular support structure disposed in the reactor vessel on the bottom end wall and extending about the lower end of the core; (f) structural support means disposed between the containment structure base mat and bottom end of the reactor vessel wall and cooperating for supporting the reactor vessel at its bottom end wall on the containment structure base mat to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event; (g) a bed of insulating material disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall; freely expand radially from the central post as it heats up while providing continuous support thereof; (h) a deck supported upon the wall of the containment vessel above the top open end of the reactor vessel; and (i) extendible and retractable coupling means extending between the deck and the top open end of the reactor vessel and flexibly and sealably interconnecting the reactor vessel at its top end to the deck

Nuclear reactor melt-retention structure to mitigate direct containment heating

International Nuclear Information System (INIS)

Tutu, N.K.; Ginsberg, T.; Klages, J.R.

1991-01-01

This patent describes a nuclear reactor melt-retention structure that functions to retain molten core material within a melt retention chamber to mitigate the extent of direct containment heating. The structure being adapted to be positioned within or adjacent to a pressurized or boiling water nuclear reactor containment building at a location such that at least a portion of the melt retention structure is lower than and to one side of the nuclear reactor pressure vessel, and such that the structure is adjacent to a gas escape channel means that communicates between the reactor cavity and the containment building of the reactor. It comprises a melt-retention chamber, wall means defining a passageway extending between the reactor cavity underneath the reactor pressure vessel and one side of the chamber, the passageway including vent means extending through an upper wall portion thereof. The vent means being in communication with the upper region of the reactor containment building, whereby gas and steam discharged from the reactor pressure vessel are vented through the passageway and vent means into the gas-escape channel means and the reactor containment building

Nuclear reactor container

International Nuclear Information System (INIS)

Shioiri, Akio.

1992-01-01

In a nuclear reactor container, a vent tube communication port is disposed to a pressure suppression pool at a position higher than the pool water therein for communication with an upper dry well, and the upper end opening of a dry well communication pipe is disposed at a position higher than the communication port. When condensate return pipeline is ruptured in the upper dry well, water in a water source pool is injected to the pressure vessel and partially discharged out of the ruptured port and a depressurization valve connected to the pressure vessel to the inside of the upper dry well. The discharged water stays in the upper dry well and, when the water level reaches the height of the vent tube communication port, it flows into the pressure suppression pool. Even in a state that the entire amount of water in the water source pool is supplied, since water does not reach the upper opening port of the dry well communication pipe, water does not flow into a lower dry well. Accordingly, the motor of a control rod drives disposed in the lower dry well can be prevented from submerging. The reactor core can be cooled more reliably, to improve the reliability of the pressure suppression function. (N.H.)

Assessment of Loss-of-Coolant Effect on Pressurized Heavy Water Reactors

International Nuclear Information System (INIS)

Kim, Won Young; Park, Joo Hwan; Kim, Bong Ghi

2009-01-01

A CANDU reactor is a heavy-water-moderated, natural uranium fuelled reactor with a pressure tube. The reactor contains a horizontal cylindrical vessel (calandria) and each pressure tube is isolated from the heavy-water moderator in a calandria. This allows the moderator system to be operated of a high-pressure and of a high-temperature coolant in pressure tube. This causes the pressurized liquid coolant in the channel to void and therefore give rise to a reactivity transient in the event of a break or fault in the coolant circuit. In particular, all CANDU reactors are well known to have a positive void reactivity coefficient and thus this phenomenon may lead to a positive feedback, which can cause a large power pulse. We assess the loss-of-coolant effect by coolant void reactivity versus fuel burnup, four factor parameters for fresh fuel and equilibrium fuel, reactivity change due to the change of coolant density and reactivity change in the case of half- and full-core coolant

First Study of Helium Gas Purification System as Primary Coolant of Co-Generation Reactor

International Nuclear Information System (INIS)

Piping Supriatna

2009-01-01

The technological progress of NPP Generation-I on 1950’s, Generation-II, Generation-III recently on going, and Generation-IV which will be implemented on next year 2025, concept of nuclear power technology implementation not only for generate electrical energy, but also for other application which called cogeneration reactor. Commonly the type of this reactor is High Temperature Reactor (HTR), which have other capabilities like Hydrogen production, desalination, Enhanced Oil Recovery (EOR), etc. The cogeneration reactor (HTR) produce thermal output higher than commonly Nuclear Power Plant, and need special Heat Exchanger with helium gas as coolant. In order to preserve heat transfer with high efficiency, constant purity of the gas must be maintained as well as possible, especially contamination from its impurities. In this report has been done study for design concept of HTR primary coolant gas purification system, including methodology by sampling He gas from Primary Coolant and purification by using Physical Helium Splitting Membrane. The examination has been designed in physical simulator by using heater as reactor core. The result of study show that the of Primary Coolant Gas Purification System is enable to be implemented on cogeneration reactor. (author)

A sustainability analysis of the Brazilian multipurpose reactor project

International Nuclear Information System (INIS)

Obadia, I.J.; Perrotta, J.A.

2010-01-01

The project of a new research reactor in Brazil for radioisotope production, support of the nuclear energy program and scientific research has received a positive sign of the government and is starting to be developed by the Brazilian Commission of Nuclear Energy. International Atomic Energy Agency points out that the implementation of a new research reactor is a major undertaking for a country, requiring an analysis to identify to which extent the conditions of the national nuclear program are proper and adequate to lead to a sustainable research reactor life cycle. This paper introduces the Brazilian Multipurpose Reactor Project (RMB) and describes the sustainability analysis performed, which has shown that the national nuclear infrastructure presents a very favourable condition to the implementation of the RMB project as well as to provide a sustainable life cycle for this new research reactor. (author)

Coolant make-up device for BWR type reactor

International Nuclear Information System (INIS)

Sasagawa, Hiroshi.

1994-01-01

In a coolant make-up device, an opening of a pressure equalizing pipeline in a pressure vessel is disposed in coolants above a reactor core and below a usual fluctuation range of a reactor vessel water level. Further, a float check valve is disposed to the pressure equalizing pipeline for preventing coolants in the pressure vessel flowing into the pipeline. If the water level in the pressure vessel is lowered than the setting position for the float check valve, the float drops by its own weight to open the opening of the pressure equalizing pipeline. Then, steams in the pressure vessel are flown into the pipeline, to equalize the pressure between a coolant storage tank and the pressure vessel of the reactor. Coolants in the coolant storage tank is injected to the pressure vessel by way of the water injection pipeline due to the difference of the pressure head between the water level in the coolants storage tank and the water level in the pressure vessel. If the coolants are lowered than the setting position for the float check value, the float check valve does not close unless the water level is recovered to the setting position for the float valve and, accordingly, the coolant make-up is continued. (N.H.)

Proposal on experience learning of a nuclear reactor for children in future. A basic concept on a nuclear reactor facility for demonstration and education

International Nuclear Information System (INIS)

Murata, Takashi; Yoshiki, Nobuya; Kinehara, Yoshiki; Nakagawa, Haruo

2001-01-01

The Science Council of Japan indicates in a proposal on R and D on nuclear energy forward the 21st Century that it is important to expand the educational object on nuclear energy from colleges and gradual schools to elementary, middle high schools. And, the Committee of Japan Nuclear Energy Industries also proposed that as an effort forward security of reliability and popularization of knowledge, completeness of learning chance on energy and nuclear energy in education such as usage of general learning time, concept on establishment of educational reactor for demonstration and experience, is essential. Here was described on a concept on establishment of nuclear reactor for demonstration and experience at objectives of common national peoples, which was based on results of searches and investigations carried out by authors and aimed to supply to a field to grow up a literary adequately and widely capable of judging various information on the peoples by focusing to effectiveness of empirical learning as a method of promoting corrective understanding of common citizens on high class technical system and by establishment of the reactor aiming at general education on nuclear energy at a place easily accessible by common citizens, such as large city. (G.K.)

Proposal on experience learning of a nuclear reactor for children in future. A basic concept on a nuclear reactor facility for demonstration and education

Energy Technology Data Exchange (ETDEWEB)

Murata, Takashi [Kyoto Univ., Graduate School of Energy Science, Kyoto (Japan); Yoshiki, Nobuya; Kinehara, Yoshiki; Nakagawa, Haruo

2001-12-01

The Science Council of Japan indicates in a proposal on R and D on nuclear energy forward the 21st Century that it is important to expand the educational object on nuclear energy from colleges and gradual schools to elementary, middle high schools. And, the Committee of Japan Nuclear Energy Industries also proposed that as an effort forward security of reliability and popularization of knowledge, completeness of learning chance on energy and nuclear energy in education such as usage of general learning time, concept on establishment of educational reactor for demonstration and experience, is essential. Here was described on a concept on establishment of nuclear reactor for demonstration and experience at objectives of common national peoples, which was based on results of searches and investigations carried out by authors and aimed to supply to a field to grow up a literary adequately and widely capable of judging various information on the peoples by focusing to effectiveness of empirical learning as a method of promoting corrective understanding of common citizens on high class technical system and by establishment of the reactor aiming at general education on nuclear energy at a place easily accessible by common citizens, such as large city. (G.K.)

Reactor calculations and nuclear information

International Nuclear Information System (INIS)

Lang, D.W.

1977-12-01

The relationship of sets of nuclear parameters and the macroscopic reactor quantities that can be calculated from them is examined. The framework of the study is similar to that of Usachev and Bobkov. The analysis is generalised and some properties required by common sense are demonstrated. The form of calculation permits revision of the parameter set. It is argued that any discrepancy between a calculation and measurement of a macroscopic quantity is more useful when applied directly to prediction of other macroscopic quantities than to revision of the parameter set. The mathematical technique outlined is seen to describe common engineering practice. (Author)

Safety upgrades to the NRU research reactor

International Nuclear Information System (INIS)

DeAbreu, B.; Mark, J.M.; Mutterback, E.J.

1998-01-01

The NRU (National Research Universal) Reactor is a 135 MW thermal research facility located at Chalk River Laboratories, and is owned and operated by Atomic Energy of Canada Limited. One of the largest and most versatile research reactors in the world, it serves as the R and D workhorse for Canada's CANDU business while at the same time filling the role as one of the world's major producers of medical radioisotopes. AECL plans to extend operation of the NRU reactor to approximately the year 2005 when a new replacement, the Irradiation Research Facility (IRF) will be available. To achieve this, AECL has undertaken a program of safety reassessment and upgrades to enhance the level of safety consistent with modem requirements. An engineering assessment/inspection of critical systems, equipment and components was completed and seven major safety upgrades are being designed and installed. These upgrades will significantly reduce the reactor's vulnerability to common mode failures and external hazards, with particular emphasis on seismic protection. The scheduled completion date for the project is 1999 December at a cost approximately twice the annual operating cost. All work on the NRU upgrade project is planned and integrated into the regular operating cycles of the reactor; no major outages are anticipated. This paper describes the safety upgrades and discusses the technical and managerial challenges involved in extending the operating life of the NRU reactor. (author)

Condensation During Nuclear Reactor Loca

International Nuclear Information System (INIS)

Rihan, Y.; Teamah, M.; Sorour, M.; Soliman, S.

2008-01-01

Two-phase channel flow with condensation is a common phenomenon occurs in a number of nuclear reactor accident scenarios. It also plays an important role during the operation of the safety coolant injection systems in advanced nuclear reactors. Semiempirical correlations and simple models based on the analogy between heat and mass transfer processes have been previously applied. Rigorous models, compatible with the state-of-the-art numerical algorithms used in thermal-hydraulic computer codes, are scare, and are of great interest. The objective of this research is to develop a method for modeling condensation, with noncondensable gases, compatible with the state-of-the-art numerical methods for the solution of multi-phase field equations. A methodology for modeling condensation, based on the stagnant film theory, and compatible with the reviewed numerical algorithms, is developed. The model treats the coupling between the heat and mass transfer processes, and allows for an implicit treatment of the mass and momentum exchange terms as the gas-liquid interphase, without iterations. The developed model was used in the application of loss of coolant in pressurized water reactor accidents

New applications of neutron noise theory in power reactor physics

Energy Technology Data Exchange (ETDEWEB)

Arzhanov, Vasiliy

2000-04-01

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

New applications of neutron noise theory in power reactor physics

International Nuclear Information System (INIS)

Arzhanov, Vasiliy

2000-04-01

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

Regional groundwater flow model for C, K. L. and P reactor areas, Savannah River Site, Aiken, SC

Energy Technology Data Exchange (ETDEWEB)

Flach, G.P.

2000-02-11

A regional groundwater flow model encompassing approximately 100 mi2 surrounding the C, K, L, and P reactor areas has been developed. The reactor flow model is designed to meet the planning objectives outlined in the General Groundwater Strategy for Reactor Area Projects by providing a common framework for analyzing groundwater flow, contaminant migration and remedial alternatives within the Reactor Projects team of the Environmental Restoration Department. The model provides a quantitative understanding of groundwater flow on a regional scale within the near surface aquifers and deeper semi-confined to confined aquifers. The model incorporates historical and current field characterization data up through Spring 1999. Model preprocessing is automated so that future updates and modifications can be performed quickly and efficiently. The CKLP regional reactor model can be used to guide characterization, perform scoping analyses of contaminant transport, and serve as a common base for subsequent finer-scale transport and remedial/feasibility models for each reactor area.

Regional groundwater flow model for C, K. L. and P reactor areas, Savannah River Site, Aiken, SC

International Nuclear Information System (INIS)

Flach, G.P.

2000-01-01

A regional groundwater flow model encompassing approximately 100 mi2 surrounding the C, K, L, and P reactor areas has been developed. The reactor flow model is designed to meet the planning objectives outlined in the General Groundwater Strategy for Reactor Area Projects by providing a common framework for analyzing groundwater flow, contaminant migration and remedial alternatives within the Reactor Projects team of the Environmental Restoration Department. The model provides a quantitative understanding of groundwater flow on a regional scale within the near surface aquifers and deeper semi-confined to confined aquifers. The model incorporates historical and current field characterization data up through Spring 1999. Model preprocessing is automated so that future updates and modifications can be performed quickly and efficiently. The CKLP regional reactor model can be used to guide characterization, perform scoping analyses of contaminant transport, and serve as a common base for subsequent finer-scale transport and remedial/feasibility models for each reactor area

The evolution of Canadian research reactors: 1942 to 1992

International Nuclear Information System (INIS)

Lidstone, R.F.

1992-01-01

Research reactors have played a central role in the development of the Canadian nuclear program. Major benefits arising from the operation of Canada's research reactors include the development and demonstration of Canada deuterium uranium (CANDU) power reactors and the creation of an international business based on the production of medical and industrial radioisotopes. Insofar as practicable, nuclear engineers have based each design on the evolutionary development of appropriate technology for Canada. This approach has resulted in several generations of multipurpose reactors with a common heritage of neutronic efficiency. A focus on neutronic efficiency stemmed from the incentive for direct use of Canada's abundant uranium resources; the desired reliance on natural-uranium fuels promoted interest in efficient moderators, especially heavy water, and structural materials with a low affinity for absorbing neutrons, such as aluminum and zirconium alloys

Mode of operation of a nuclear reactor

International Nuclear Information System (INIS)

Morita, T.

1976-01-01

A method is proposed for the operation of a nuclear reactor guaranteeing an essentially symmetrical axial power distribution during normal operation by controlling the changes occuring in the reactor power partly by variation of the concentration of the neutron-absorbing element and partly by variation of the control rod positions. The representative parameters are recorded for the upper and lower half and adjusted to a predetermined reference value. In using this method, the axial power peals are reduced and power losses avoided. (RW) [de

Nuclear detectors for in-core power-reactors

International Nuclear Information System (INIS)

Duchene, Jean; Verdant, Robert.

1979-12-01

Nuclear reactor control is commonly obtained through neutronic measurements, ex-core and in-core. In large size reactors flux instabilities may take place. For a good monitoring of them, local in-core power measurements become particularly useful. This paper intends to review the questions about neutronic sensors with could be used in-core. A historical account about methods is given first, from early power reactors with brief description of each system. Sensors presently used (ionization fission chambers, self-powered detectors) are then considered and also those which could be developped such as gamma thermometers. Their physical basis, main characteristics and operation modes are detailed. Preliminary tests and works needed for an extension of their life-time are indicated. As an example present irradiation tests at the CEA are then proposed. Two tables will help comparing the characteristics of each type in terms of its precise purpose: fuel monitoring, safety or power control. Finally a table summarizes the kind of sensors mounted on working power reactors and another one is a review of characteristics for some detectors from obtainable commercial sheets [fr

Driver options and burn cycle selection based on power reactor considerations

International Nuclear Information System (INIS)

Ehst, D.A.

1983-01-01

Reactor implications for noninductive current drive are presented based on a number of studies. First, the lower hybrid driver for the STARFIRE reactor is discussed and the disadvantages of this driver are reviewed. Next, the results of an extensive search for a better current driver are presented. A large number of alternatives were compared in a common context, the DEMO reactor, in order to examine their suitability on a standard basis. Finally, the methodology of a study, currently in progress, is described. The goals of this last study are to compare tokamak reactor designs optimized for operation under different burn cycles, in order to assess the actual benefits and costs of pulsed versus steady-state operation. (author)

Driver options and burn-cycle selection based on power-reactor considerations

International Nuclear Information System (INIS)

Ehst, D.A.

1983-04-01

Reactor implications for noninductive current drive are presented based on a number of studies. First, the lower hybrid driver for the STARFIRE reactor is discussed and the disadvantages of this driver are reviewed. Next, the results of an extensive search for a better current driver are presented. A large number of alternatives were compared in a common context, the DEMO reactor, in order to examine their suitability on a standard basis. Finally, the methodology of a study, currently in progress, is described. The goals of this last study are to compare tokamak reactor designs optimized for operation under different burn cycles, in order to assess the actual benefits and costs of pulsed versus steady-state operation

Comparative analysis of a hypothetical coolant loss accident in an LMFB reactor with the use of various calculation models for a common reference problem

International Nuclear Information System (INIS)

Royl, P.

1979-01-01

The results of a comparative analysis of the initial and dismantling stages of a hypothetical loss of flow accident in an LMFB reactor are presented. The analyses were made for a common reference problem with four different calculation models (CARMEN/KADIS, SURDYN, CAPRI/KADIS and FRAX). The reference core is described specifically, as are the differences in its geometrical disposition in the models, the static and transient conditions before and after the start of boiling and during dismantling. The differences in the models used for simulating the boiling and the dismantling are compared. The structure of the core, as well as the calculation conditions and hypotheses, were intentionally designed so that the accident would culminate, in all cases, in an energetic hydrodynamic dismantling stage

Primary cooling system for BWR type reactor

International Nuclear Information System (INIS)

Ibe, Eishi; Takahashi, Masanori; Aoki, Yasuko

1993-01-01

The present invention effectively uses information from a plurality of sensors in order to suppress corrosion circumstance of a nuclear reactor. That is, a predetermined general water quality factor at a predetermined position is determined as a standard index. A concentration of a water quality improver is controlled such that the index is within an aimed range. For this purpose, the entire sensor groups disposed in a primary coolant system of a nuclear reactor are divided into a plural systems of sensor groups each disposed on every different positions. Then, a predetermined sensor group (standard sensor group) is connected to a computing device and a data base so that it is always monitored for calculating and estimating the standard index. Only oxidative ingredient in water at the measuring point is noted, and a concentration distribution which agrees with an actually measured value of oxidative ingredients is extracted from data base and used as a correct concentration distribution. With such procedures, reactor water quality can be estimated accurately while compensating erroneous factors of individual sensors. Even when a new sensor is used, it is not necessary to greatly change control logic. (I.S.)

Design Engineering of Safety Parameter Display for Kartini Reactor

International Nuclear Information System (INIS)

Yoyok Dwi Setyo Pambudi; Suharyo Widagdo; Aliq Zuhdi; Darlis

2003-01-01

Modification information parameter of Reactor Kartini has been conducted. This program use compiler LabView 5.0 that compatible and relevant with National Instrument card. These card mostly use in reactor instrumentation and industry. The information that been display is control rod position, fuel temperature, and water pH. The development in this research is control rod figure that can move up and down as the output from reactor power. Beside this is a module-module security figure that can light if there is a warning signal. Maximum high of control rod is 100 cm. (author)

Generation IV reactors: reactor concepts

International Nuclear Information System (INIS)

Cardonnier, J.L.; Dumaz, P.; Antoni, O.; Arnoux, P.; Bergeron, A.; Renault, C.; Rimpault, G.; Delpech, M.; Garnier, J.C.; Anzieu, P.; Francois, G.; Lecomte, M.

2003-01-01

Liquid metal reactor concept looks promising because of its hard neutron spectrum. Sodium reactors benefit a large feedback experience in Japan and in France. Lead reactors have serious assets concerning safety but they require a great effort in technological research to overcome the corrosion issue and they lack a leader country to develop this innovative technology. In molten salt reactor concept, salt is both the nuclear fuel and the coolant fluid. The high exit temperature of the primary salt (700 Celsius degrees) allows a high energy efficiency (44%). Furthermore molten salts have interesting specificities concerning the transmutation of actinides: they are almost insensitive to irradiation damage, some salts can dissolve large quantities of actinides and they are compatible with most reprocessing processes based on pyro-chemistry. Supercritical water reactor concept is based on operating temperature and pressure conditions that infers water to be beyond its critical point. In this range water gets some useful characteristics: - boiling crisis is no more possible because liquid and vapour phase can not coexist, - a high heat transfer coefficient due to the low thermal conductivity of supercritical water, and - a high global energy efficiency due to the high temperature of water. Gas-cooled fast reactors combining hard neutron spectrum and closed fuel cycle open the way to a high valorization of natural uranium while minimizing ultimate radioactive wastes and proliferation risks. Very high temperature gas-cooled reactor concept is developed in the prospect of producing hydrogen from no-fossil fuels in large scale. This use implies a reactor producing helium over 1000 Celsius degrees. (A.C.)

Nuclear reactor with a fixed system of neutron poison, which can be burnt up, introduced into the reactor core

International Nuclear Information System (INIS)

Mueller, E.; Roegler, H.J.; Wickert, M.

1985-01-01

The fixed system consists of neutron poison which can be burnt up, in an uneven distribution, and with adjustable absorber rods for output control, which are driven into the reactor core from the side along the fuel elements. There is an excess of neutron poison which can be burnt up, overall, on the side of the reactor core away from the absorber rods. The reactor core is free of neutron poison which can be burnt up on the side where the absorber rods are driven in, so that the ratio of maximum to mean power density with reference to a possible absorber rod positions is less than for homogeneous distribution of the neutron poison which can be burnt up. (orig./HP) [de

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

International Nuclear Information System (INIS)

Nguyen Hoang Hai; Do Quang Binh

2011-01-01

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

Fast neutron reactors: the safety point of view

International Nuclear Information System (INIS)

Laverie, M.; Avenas, M.

1984-01-01

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

Economic evaluation of reprocessing. Indicative UK position

International Nuclear Information System (INIS)

1979-05-01

This paper, which also appears as an Appendix to the final Working Group 4 report, forms part of the overall economic assessment of reprocessing. The indicative national position and illustrative ''phase diagram'' for the United Kingdom is presented. Under conditions pertaining in the United Kingdom the diagram suggests that: if uranium prices rise rapidly the fast reactor would become economic in the decade 1990-2000, if uranium prices rise more slowly, the fast reactor would become economic in the decade 2000-2010

Russian Federation: Passive Safety Components for Lead-Cooled Reactor Facilities

International Nuclear Information System (INIS)

Sarkulov, M.K.

2015-01-01

There is a specific range of engineered features used traditionally in nuclear technology. As a rule, main reactivity control systems use conventional active actuators with solid-body control members and/or liquid systems with active injection of liquid absorber. Other operation principles are normally chosen for additional systems. Currently, the traditional approach to improving the reliability of a reactor facility suggests an increase in the number of safety components and systems which provide for mutual assurance or assist each other. There is a great variety of additional reactivity control members designed for the reactor facility control and shutdown, including hydrodynamic members in the form of rods (acting from the coolant flow); floating-type members (absorbers and displacers); storage-type and liquid members (used in separate channels); bulk members (pebble absorber); gas-based members (with a gas absorber); shape-memory members and others. Hydrodynamic systems were introduced at Beloyarsk NPP Units 1 and 2 and proposed for use in other facility designs, Gases and bulk materials have not been commonly accepted: the former because of the high cost of high-efficiency gaseous absorbers, and the latter because of the complecated monitoring of the bulk material position. It is rather difficult and not always necessary to use the same engineering approaches in new lead-cooled reactor facilities as in traditional ones. Similarly to the development of traditional safety systems, passive safety components (devices) shall be designed according to the essential requirements of the nuclear regulations of the Russian Federation

Reference design for the standard mirror hybrid reactor

Energy Technology Data Exchange (ETDEWEB)

Bender, D.J.; Fink, J.H.; Galloway, T.R.; Kastenberg, W.E.; Lee, J.D.; Devoto, R.S.; Neef, W.S. Jr.; Schultz, K.R.; Culver, D.W.; Rao, S.B.; Rao, S.R.

1978-05-22

This report describes the results of a two-year study by Lawrence Livermore Laboratory and General Atomic Co. to develop a conceptual design for the standard (minimum-B) mirror hybrid reactor. The reactor parameters have been chosen to minimize the cost of producing nuclear fuel (/sup 239/Pu) for consumption in fission power reactors (light water reactors). The deuterium-tritium plasma produces approximately 400 MW of fusion power with a plasma Q of 0.64. The fast-fission blanket, which is fueled with depleted uranium and lithium, generates sufficient tritium to run the reactor, has a blanket energy multiplication of M = 10.4, and has a net fissile breeding ratio of Pu/n = 1.51. The reactor has a net electrical output of 600 MWe, a fissile production of 2000 kg of plutonium per year (at a capacity factor of 0.74), and a net plant efficiency of 0.18. The plasma-containment field is generated by a Yin-Yang magnet using NbTi superconductor, and the neutral beam system uses positive-ion acceleration with beam direct conversion. The spherical blanket is based on gas-cooled fast reactor technology. The fusion components, blanket, and primary heat-transfer loop components are all contained within a prestressed-concrete reactor vessel, which provides magnet restraint and supports the primary heat-transfer loop and the blanket.

Reference design for the standard mirror hybrid reactor

International Nuclear Information System (INIS)

Bender, D.J.; Fink, J.H.; Galloway, T.R.; Kastenberg, W.E.; Lee, J.D.; Devoto, R.S.; Neef, W.S. Jr.; Schultz, K.R.; Culver, D.W.; Rao, S.B.; Rao, S.R.

1978-01-01

This report describes the results of a two-year study by Lawrence Livermore Laboratory and General Atomic Co. to develop a conceptual design for the standard (minimum-B) mirror hybrid reactor. The reactor parameters have been chosen to minimize the cost of producing nuclear fuel ( 239 Pu) for consumption in fission power reactors (light water reactors). The deuterium-tritium plasma produces approximately 400 MW of fusion power with a plasma Q of 0.64. The fast-fission blanket, which is fueled with depleted uranium and lithium, generates sufficient tritium to run the reactor, has a blanket energy multiplication of M = 10.4, and has a net fissile breeding ratio of Pu/n = 1.51. The reactor has a net electrical output of 600 MWe, a fissile production of 2000 kg of plutonium per year (at a capacity factor of 0.74), and a net plant efficiency of 0.18. The plasma-containment field is generated by a Yin-Yang magnet using NbTi superconductor, and the neutral beam system uses positive-ion acceleration with beam direct conversion. The spherical blanket is based on gas-cooled fast reactor technology. The fusion components, blanket, and primary heat-transfer loop components are all contained within a prestressed-concrete reactor vessel, which provides magnet restraint and supports the primary heat-transfer loop and the blanket

Core construction in a pressure tube type heavy water reactor

International Nuclear Information System (INIS)

Ueda, Makoto; Aoki, Katsutada.

1975-01-01

Object: To replace a centrally positioned fuel assembly of a fuel assembly unit with a reactor controlling machinery to decrease a distance between the fuel assemblies thereby saving use of heavy water and enhancing economy. Structure: A centrally positioned fuel assembly of a fuel assembly unit, which is composed of a plurality of fuel assemblies orderly arranged in lattice fashion, is replaced with a reactor controlling members such as control rods, poison tubes and the like to provide an arrangement of lattice-free type fuel assembly, thus reducing the pitch as small as possible. (Kamimura, M.)

Reactor TRIGA PUSPATI (RTP) spent fuel pool conceptual design

International Nuclear Information System (INIS)

Mohd Fazli Zakaria; Tonny Lanyau; Ahmad Nabil Ab Rahim

2010-01-01

Reactor TRIGA PUSPATI (RTP) is the one and only research reactor in Malaysia that has been safely operated and maintained since 1982. In order to enhance technical capabilities and competencies especially in nuclear reactor engineering a feasibility study on RTP power upgrading was proposed to serve future needs for advance nuclear science and technology in the country with the capability of designing and develop reactor system. The need of a Spent Fuel Pool begins with the discharge of spent fuel elements from RTP for temporary storage that includes all activities related to the storage of fuel until it is either sent for reprocessed or sent for final disposal. To support RTP power upgrading there will be major RTP systems replacement such as reactor components and a new temporary storage pool for fuel elements. The spent fuel pool is needed for temporarily store the irradiated fuel elements to accommodate a new reactor core structure. Spent fuel management has always been one of the most important stages in the nuclear fuel cycle and considered among the most common problems to all countries with nuclear reactors. The output of this paper will provide sufficient information to show the Spent Fuel Pool can be design and build with the adequate and reasonable safety assurance to support newly upgraded TRIGA PUSPATI TRIGA Research Reactor. (author)

Grey water treatment in UASB reactor at ambient temperature.

Science.gov (United States)

Elmitwalli, T A; Shalabi, M; Wendland, C; Otterpohl, R

2007-01-01

In this paper, the feasibility of grey water treatment in a UASB reactor was investigated. The batch recirculation experiments showed that a maximum total-COD removal of 79% can be obtained in grey-water treatment in the UASB reactor. The continuous operational results of a UASB reactor treating grey water at different hydraulic retention time (HRT) of 20, 12 and 8 hours at ambient temperature (14-24 degrees C) showed that 31-41% of total COD was removed. These results were significantly higher than that achieved by a septic tank (11-14%), the most common system for grey water pre-treatment, at HRT of 2-3 days. The relatively lower removal of total COD in the UASB reactor was mainly due to a higher amount of colloidal COD in the grey water, as compared to that reported in domestic wastewater. The grey water had a limited amount of nitrogen, which was mainly in particulate form (80-90%). The UASB reactor removed 24-36% and 10-24% of total nitrogen and total phosphorus, respectively, in the grey water, due to particulate nutrients removal by physical entrapment and sedimentation. The sludge characteristics of the UASB reactor showed that the system had stable performance and the recommended HRT for the reactor is 12 hours.

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

Science.gov (United States)

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

2018-02-01

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

Reactor instrumentation renewal of the TRIGA reactor Vienna, Austria

International Nuclear Information System (INIS)

Boeck, H.; Weiss, H.; Hood, W.E.; Hyde, W.K.

1992-01-01

The TRIGA Mark-II reactor at the Atominstitut in Vienna, Austria is replacing its twenty-four year old instrumentation system with a microprocessor based control system supplied by General Atomics. Ageing components, new governmental safety requirements and a need for state of the art instrumentation for training students has spurred the demand for new reactor instrumentation. In Austria a government appointed expert is assigned the responsibility of reviewing the proposed installation and verifying all safety aspects. After a positive review, final assembly and checkout of the instrumentation system may commence. The instrumentation system consists of three basic modules: the control system console, the data acquisition console and the NH-1000 wide range channel. Digital communications greatly reduce interwiring requirements. Hardwired safety channels are independent of computer control, thus, the instrumentation system in no way relies on any computer intervention for safety function. In addition, both the CSC and DAC computers are continuously monitored for proper operation via watchdog circuits which are capable of shutting down the reactor in the event of computer malfunction. Safety channels include two interlocked NMP-1000 multi-range linear channels for steady state mode, an NPP-1000 linear safety channel for pulse mode and a set of three independent fuel temperature monitoring channels. The microprocessor controlled wide range NM- 1000 digital neutron monitor (fission chamber based) functions as a startup/operational channel, and provides all power level related Interlocks. The Atominstitut TRIGA reactor is configured for four modes of operation: manual mode, automatic mode (servo control), pulsing mode and square wave mode. Control of the standard control rods is via stepping motor control rod drives, which offers the operator the choice of which control rods are operated by the servo system in automatic and square wave model. (author)

Fuel bundle for nuclear reactor

International Nuclear Information System (INIS)

Long, J.W.; Flora, B.S.; Ford, K.L.

1977-01-01

The invention concerns a new, simple and inexpensive system for assembling and dismantling a nuclear reactor fuel bundle. Several fuel rods are fitted in parallel rows between two retaining plates which secure the fuel rods in position and which are maintained in an assembled position by means of several stays fixed to the two end plates. The invention particularly refers to an improved apparatus for fixing the stays to the upper plate by using locking fittings secured to rotating sleeves which are applied against this plate [fr

An approach for coupled-code multiphysics core simulations from a common input

International Nuclear Information System (INIS)

Schmidt, Rodney; Belcourt, Kenneth; Hooper, Russell; Pawlowski, Roger; Clarno, Kevin; Simunovic, Srdjan; Slattery, Stuart; Turner, John; Palmtag, Scott

2015-01-01

Highlights: • We describe an approach for coupled-code multiphysics reactor core simulations. • The approach can enable tight coupling of distinct physics codes with a common input. • Multi-code multiphysics coupling and parallel data transfer issues are explained. • The common input approach and how the information is processed is described. • Capabilities are demonstrated on an eigenvalue and power distribution calculation. - Abstract: This paper describes an approach for coupled-code multiphysics reactor core simulations that is being developed by the Virtual Environment for Reactor Applications (VERA) project in the Consortium for Advanced Simulation of Light-Water Reactors (CASL). In this approach a user creates a single problem description, called the “VERAIn” common input file, to define and setup the desired coupled-code reactor core simulation. A preprocessing step accepts the VERAIn file and generates a set of fully consistent input files for the different physics codes being coupled. The problem is then solved using a single-executable coupled-code simulation tool applicable to the problem, which is built using VERA infrastructure software tools and the set of physics codes required for the problem of interest. The approach is demonstrated by performing an eigenvalue and power distribution calculation of a typical three-dimensional 17 × 17 assembly with thermal–hydraulic and fuel temperature feedback. All neutronics aspects of the problem (cross-section calculation, neutron transport, power release) are solved using the Insilico code suite and are fully coupled to a thermal–hydraulic analysis calculated by the Cobra-TF (CTF) code. The single-executable coupled-code (Insilico-CTF) simulation tool is created using several VERA tools, including LIME (Lightweight Integrating Multiphysics Environment for coupling codes), DTK (Data Transfer Kit), Trilinos, and TriBITS. Parallel calculations are performed on the Titan supercomputer at Oak

Instrumentation device at the outside of reactor and method of using the same

International Nuclear Information System (INIS)

Ichige, Masayuki.

1997-01-01

The present invention provides instrumentation device at the outside of a reactor capable of measuring conditions of inside of a reactor, such as the power of the reactor, distribution of voids or water level while considering hysteresis of neutrons or γ-rays from the inside to the outside of the reactor. Namely, a plurality of radiation detectors are disposed being elongated in vertical direction at a predetermined distance on the outer circumference of a reactor pressure vessel. The detectors detect intensity of radiation rays and the detection time at a plurality of positions of the outer side of the reactor. An amplifier amplifies the detected signals. A signal processing device determines the positions and the time of the emitted radiation rays based on the amplified detected signals. An analysis device analyzes spacial distribution and time distribution of the energy and the intensity of the radiation rays (neutron or γ-rays) based on the signals of predetermined radiation rays at the outer side of the reactor. Then, spacial and time variation components and the power distribution, water level, change of the water level, void distribution are calculated while considering decay of the radiation rays based on the distribution of material densities of incore structures. (I.S.)

Nuclear reactor fuel assembly

International Nuclear Information System (INIS)

1975-01-01

A description is given of a nuclear reactor fuel assembly comprising a cluster of fuel elements supported by transversal grids so that their axes are parallel to and at a distance from each other, in order to establish interstices for the axial flow of a coolant. At least one of the interstices is occupied by an axial duct reserved for an auxiliary cooling fluid and is fitted with side holes through which the auxiliary cooling fluid is sprayed into the cluster. Deflectors extend as from a transversal grid in a position opposite the holes to deflect the cooling fluid jet towards those parts of the fuel elements that are not accessible to the auxiliary coolant. This assembly is intended for reactors cooled by light or heavy water [fr

Microcomputer-based equipment-control and data-acquisition system for fission-reactor reactivity-worth measurements

International Nuclear Information System (INIS)

McDowell, W.P.; Bucher, R.G.

1980-01-01

Material reactivity-worth measurements are one of the major classes of experiments conducted on the Zero Power research reactors (ZPR) at Argonne National Laboratory. These measurements require the monitoring of the position of a servo control element as a sample material is positioned at various locations in a critical reactor configuration. In order to guarantee operational reliability and increase experimental flexibility for these measurements, the obsolete hardware-based control unit has been replaced with a microcomputer based equipment control and data acquisition system. This system is based on an S-100 bus, dual floppy disk computer with custom built cards to interface with the experimental system. To measure reactivity worths, the system accurately positions samples in the reactor core and acquires data on the position of the servo control element. The data are then analyzed to determine statistical adequacy. The paper covers both the hardware and software aspects of the design

Microcomputer-based equipment-control and data-acquisition system for fission-reactor reactivity-worth measurements

Energy Technology Data Exchange (ETDEWEB)

McDowell, W.P.; Bucher, R.G.

1980-01-01

Material reactivity-worth measurements are one of the major classes of experiments conducted on the Zero Power research reactors (ZPR) at Argonne National Laboratory. These measurements require the monitoring of the position of a servo control element as a sample material is positioned at various locations in a critical reactor configuration. In order to guarantee operational reliability and increase experimental flexibility for these measurements, the obsolete hardware-based control unit has been replaced with a microcomputer based equipment control and data acquisition system. This system is based on an S-100 bus, dual floppy disk computer with custom built cards to interface with the experimental system. To measure reactivity worths, the system accurately positions samples in the reactor core and acquires data on the position of the servo control element. The data are then analyzed to determine statistical adequacy. The paper covers both the hardware and software aspects of the design.

Nuclear reactor

International Nuclear Information System (INIS)

Hattori, Sadao; Sato, Morihiko.

1994-01-01

Liquid metals such as liquid metal sodium are filled in a reactor container as primary coolants. A plurality of reactor core containers are disposed in a row in the circumferential direction along with the inner circumferential wall of the reactor container. One or a plurality of intermediate coolers are disposed at the inside of an annular row of the reactor core containers. A reactor core constituted with fuel rods and control rods (module reactor core) is contained at the inside of each of the reactor core containers. Each of the intermediate coolers comprises a cylindrical intermediate cooling vessels. The intermediate cooling vessel comprises an intermediate heat exchanger for heat exchange of primary coolants and secondary coolants and recycling pumps for compulsorily recycling primary coolants at the inside thereof. Since a plurality of reactor core containers are thus assembled, a great reactor power can be attained. Further, the module reactor core contained in one reactor core vessel may be small sized, to facilitate the control for the reactor core operation. (I.N.)

Safety Design Criteria of Indian Sodium Cooled Fast Reactors

International Nuclear Information System (INIS)

Pillai, P.; Chellapandi, P.; Chetal, S.C.; Vasudeva Rao, P.R.

2013-01-01

• Important feedback has been gained through the design and safety review of PFBR. • The safety criteria document prepared by AERB and IGCAR would provide important input to prepare the dedicated document for the Sodium cooled Fast Reactors at the national and international level. • A common approach with regard to safety, among countries pursuing fast reactor program, is desirable. • Sharing knowledge and experimental facilities on collaborative basis. • Evolution of strong safety criteria – fundamental to assure safety

Position detector

International Nuclear Information System (INIS)

Hayakawa, Toshifumi.

1985-01-01

Purpose: To enable to detect the position of an moving object in a control rod position detector, stably in a digital manner at a high accuracy and free from the undesired effects of circumstantial conditions such as the reactor temperature. Constitution: Coils connected in parallel with each other are disposed along the passage of a moving object and variable resistors and relays are connected in series with each of the coils respectively. Light emitting diodes is connected in series with the contacts of the respective relays. The resistance value of the variable resistors are adjusted depending on the changes in the circumstantial conditions and temperature distribution upon carrying out the positional detection. When the object is inserted into a coils, the relevant relay is deenergized, by which the relay contacts are closed to light up the diode. In the same manner, as the object is successively inserted into the coils, the diodes are lighted-up successively thereby enabling highly accurate and stable positional detection in a digital manner, free from the undesired effects of the circumstantial conditions. (Horiuchi, T.)

Implementation of a management system for operating organizations of research reactors

International Nuclear Information System (INIS)

Kibrit, Eduardo; Aquino, Afonso Rodrigues de; Zouain, Desiree Moraes

2011-01-01

This paper presents the requirements established by an IAEA draft technical document for the implementation of a management system for operating organisations of research reactors. The following aspects will be discussed: structure of IAEA draft technical document, management system requirements, processes common to all research reactors, aspects for the implementation of the management system, and a formula for grading the management system requirements. (author)

Implementation of a management system for operating organizations of research reactors

Energy Technology Data Exchange (ETDEWEB)

Kibrit, Eduardo, E-mail: kibrit@ctmsp.mar.mil.b [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil); Aquino, Afonso Rodrigues de; Zouain, Desiree Moraes, E-mail: araquino@ipen.b, E-mail: dmzouain@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2011-07-01

This paper presents the requirements established by an IAEA draft technical document for the implementation of a management system for operating organisations of research reactors. The following aspects will be discussed: structure of IAEA draft technical document, management system requirements, processes common to all research reactors, aspects for the implementation of the management system, and a formula for grading the management system requirements. (author)

Nuclear reactors

International Nuclear Information System (INIS)

Prescott, R.F.; George, B.V.; Baglin, C.J.

1979-01-01

In a nuclear reactor (e.g. one having coolant down-flow through a core to a hearth below) thermal insulation (e.g. of a floor of the hearth) comprises a layer of bricks and a layer of tiles thereon, with smaller clearances between the tiles than between the bricks but with the bricks being of reduced cross-section immediately adjacent the tiles so as to be surrounded by interconnected passages, of relatively large dimensions, constituting a continuous chamber extending behind the layer of tiles. By this arrangement, lateral coolant flow in the inter-brick clearances is much reduced. The reactor core is preferably formed of hexagonal columns, supported on diamond-shaped plates each supported on a pillar resting on one of the hearth-floor tiles. Each plate has an internal duct, four upper channels connecting the duct with coolant ducts in four core columns supported by the plate, and lower channels connecting the duct to a downwardly-open recess common to three plates, grouped to form a hexagon, at their mutually-adjacent corners. This provides mixing, and temperature-averaging, of coolant from twelve columns

Rodded shutdown system for a nuclear reactor

International Nuclear Information System (INIS)

Golden, M.P.; Govi, A.R.

1978-01-01

A top mounted nuclear reactor diverse rodded shutdown system utilizing gas fed into a pressure bearing bellows region sealed at the upper extremity to an armature is described. The armature is attached to a neutron absorber assembly by a series of shafts and connecting means. The armature is held in an uppermost position by an electromagnet assembly or by pressurized gas in a second embodiment. Deenergizing the electromagnet assembly, or venting the pressurized gas, causes the armature to fall by the force of gravity, thereby lowering the attached absorber assembly into the reactor core

Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

Energy Technology Data Exchange (ETDEWEB)

D. E. Shropshire

2009-01-01

The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

International Nuclear Information System (INIS)

Shropshire, D.E.

2009-01-01

The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program's understanding of the cost drivers that will determine nuclear power's cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-irradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

Probabilistic analysis of ''common mode failures''

International Nuclear Information System (INIS)

Easterling, R.G.

1978-01-01

Common mode failure is a topic of considerable interest in reliability and safety analyses of nuclear reactors. Common mode failures are often discussed in terms of examples: two systems fail simultaneously due to an external event such as an earthquake; two components in redundant channels fail because of a common manufacturing defect; two systems fail because a component common to both fails; the failure of one system increases the stress on other systems and they fail. The common thread running through these is a dependence of some sort--statistical or physical--among multiple failure events. However, the nature of the dependence is not the same in all these examples. An attempt is made to model situations, such as the above examples, which have been termed ''common mode failures.'' In doing so, it is found that standard probability concepts and terms, such as statistically dependent and independent events, and conditional and unconditional probabilities, suffice. Thus, it is proposed that the term ''common mode failures'' be dropped, at least from technical discussions of these problems. A corollary is that the complementary term, ''random failures,'' should also be dropped. The mathematical model presented may not cover all situations which have been termed ''common mode failures,'' but provides insight into the difficulty of obtaining estimates of the probabilities of these events

Progress report on joint utilization of the Musashi reactor, 10(1985)

International Nuclear Information System (INIS)

1986-06-01

The results of the research by those who commonly utilized the reactor in Musashi Institute of Technology and others in fiscal year 1985 were compiled as a report. The number of application was one case of medical treatment, three cases related to living things, five experiments and 34 cases of general irradiation, and all were adopted. The number of actual utilization was all except one case related to living things. The actual result of the daily utilization from September, 1985 to March, 1986 is shown. The gists of 42 reports are collected. In fiscal year 1986, 44 researches are adopted for the common utilization of the reactor and others. The connection between the Musashi reactor and the large computer center in University of Tokyo was increased to 9600 bps, and 4800 bps are used for carrying out remote jobs as before, while 4800 bps are used for utilizing four TSS terminal equipments. (Kako, I.)

Brazilian multipurpose reactor

Energy Technology Data Exchange (ETDEWEB)

NONE

2014-07-01

The Brazilian Multipurpose Reactor (RMB) Project is an action of the Federal Government, through the Ministry of Science Technology and Innovation (MCTI) and has its execution under the responsibility of the Brazilian National Nuclear Energy Commission (CNEN). Within the CNEN, the project is coordinated by the Research and Development Directorate (DPD) and developed through research units of this board: Institute of Nuclear Energy Research (IPEN); Nuclear Engineering Institute (IEN); Centre for Development of Nuclear Technology (CDTN); Regional Center of Nuclear Sciences (CRCN-NE); and Institute of Radiation Protection and Dosimetry (IRD). The Navy Technological Center in Sao Paulo (CTMSP) and also the participation of other research centers, universities, laboratories and companies in the nuclear sector are important and strategic partnerships. The conceptual design and the safety analysis of the reactor and main facilities, related to nuclear and environmental licensing, are performed by technicians of the research units of DPD / CNEN. The basic design was contracted to engineering companies as INTERTHECNE from Brazil and INVAP from Argentine. The research units from DPD/CNEN are also responsible for the design verification on all engineering documents developed by the contracted companies. The construction and installation should be performed by specific national companies and international partnerships. The Nuclear Reactor RMB will be a open pool type reactor with maximum power of 30 MW and have the OPAL nuclear reactor of 20 MW, built in Australia and designed by INVAP, as reference. The RMB reactor core will have a 5x5 configuration, consisting of 23 elements fuels (EC) of U{sub 3}Si{sub 2} dispersion-type Al having a density of up to 3.5 gU/cm{sup 3} and enrichment of 19.75% by weight of {sup 23{sup 5}}U. Two positions will be available in the core for materials irradiation devices. The main objectives of the RMB Reactor and the other nuclear and radioactive

Brazilian multipurpose reactor

International Nuclear Information System (INIS)

2014-01-01

The Brazilian Multipurpose Reactor (RMB) Project is an action of the Federal Government, through the Ministry of Science Technology and Innovation (MCTI) and has its execution under the responsibility of the Brazilian National Nuclear Energy Commission (CNEN). Within the CNEN, the project is coordinated by the Research and Development Directorate (DPD) and developed through research units of this board: Institute of Nuclear Energy Research (IPEN); Nuclear Engineering Institute (IEN); Centre for Development of Nuclear Technology (CDTN); Regional Center of Nuclear Sciences (CRCN-NE); and Institute of Radiation Protection and Dosimetry (IRD). The Navy Technological Center in Sao Paulo (CTMSP) and also the participation of other research centers, universities, laboratories and companies in the nuclear sector are important and strategic partnerships. The conceptual design and the safety analysis of the reactor and main facilities, related to nuclear and environmental licensing, are performed by technicians of the research units of DPD / CNEN. The basic design was contracted to engineering companies as INTERTHECNE from Brazil and INVAP from Argentine. The research units from DPD/CNEN are also responsible for the design verification on all engineering documents developed by the contracted companies. The construction and installation should be performed by specific national companies and international partnerships. The Nuclear Reactor RMB will be a open pool type reactor with maximum power of 30 MW and have the OPAL nuclear reactor of 20 MW, built in Australia and designed by INVAP, as reference. The RMB reactor core will have a 5x5 configuration, consisting of 23 elements fuels (EC) of U 3 Si 2 dispersion-type Al having a density of up to 3.5 gU/cm 3 and enrichment of 19.75% by weight of 23 5 U. Two positions will be available in the core for materials irradiation devices. The main objectives of the RMB Reactor and the other nuclear and radioactive facilities are

Wet steam turbines for CANDU-Reactors

International Nuclear Information System (INIS)

Westmacott, C.H.L.

1977-01-01

The technical characteristics of 4 wet steam turbine aggregates used in the Pickering nuclear power station are reported on along with operational experience. So far, the general experience was positive. Furthermore, plans are mentioned to use this type of turbines in other CANDU reactors. (UA) [de

Operation of Packed-Bed Reactors Studied in Microgravity

Science.gov (United States)

Motil, Brian J.; Balakotaiah, Vemuri

2004-01-01

The operation of a packed bed reactor (PBR) involves gas and liquid flowing simultaneously through a fixed-bed of solid particles. Depending on the application, the particles can be various shapes and sizes but are generally designed to force the two fluid phases through a tortuous route of narrow channels connecting the interstitial space. The PBR is the most common type of reactor in industry because it provides for intimate contact and high rates of transport between the phases needed to sustain chemical or biological reactions. The packing may also serve as either a catalyst or as a support for growing biological material. Furthermore, this type of reactor is relatively compact and requires minimal power to operate. This makes it an excellent candidate for unit operations in support of long-duration human space activities.

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

International Nuclear Information System (INIS)

Anon.

1986-01-01

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

A complete fuel development facility utilizing a dual core TRIGA reactor system

Energy Technology Data Exchange (ETDEWEB)

Middleton, A; Law, G C [General Atomic Co., San Diego, CA (United States)

1974-07-01

A TRIGA Dual Core Reactor System has been chosen by the Romanian Government as the heart of a new fuel development facility which will be operated by the Romanian Institute for Nuclear Technologies. The Facility, which will be operational in 1976, is an integral part of the Romanian National Program for Power Reactor Development, with particular emphasis being placed on fuel development. The unique combination of a new 14 MW steady state TRIGA reactor, and the well-proven TRIGA Annular Core Pulsing Reactor (ACPR) in one below-ground reactor pool resulted in a substantial construction cost savings and gives the facility remarkable experimental flexibility. The inherent safety of the TRIGA fuel elements in both reactor cores means that a secondary containment building is not necessary, resulting in further construction cost savings. The 14 MW steady state reactor gives acceptably high neutron fluxes for long- term testing of various prototype fuel-cladding-coolant combinations; and the TRIGA ACPR high pulse capability allows transient testing of fuel specimens, which is so important for accurate prediction of the performance of power reactor fuel elements under postulated failure conditions. The 14 MW steady state reactor has one large and three small in-core irradiation loop positions, two large irradiation loop positions adjacent to the core face, and twenty small holes in the beryllium reflector for small capsule irradiation. The power level of 14 MW will yield peak unperturbed thermal neutron fluxes in the central experiment position approaching 3.0 x 10{sup 14} n/cm{sup 2}-sec. The ACPR has one large dry central experimental cavity which can be loaded at pool level through a shielded offset loading tube; a small diameter in-core flux trap; and an in-core pneumatically-operated capsule irradiation position. A peak pulse of 15,000 MW will yield a peak fast neutron flux in the central experimental cavity of about 1.5 x 10{sup 17} n/cm{sup 2}-sec. The pulse width at

Reactor science and technology: operation and control of reactors

International Nuclear Information System (INIS)

Qiu Junlong

1994-01-01

This article is a collection of short reports on reactor operation and research in China in 1991. The operation of and research activities linked with the Heavy Water Research Reactor, Swimming Pool Reactor and Miniature Neutron Source Reactor are briefly surveyed. A number of papers then follow on the developing strategies in Chinese fast breeder reactor technology including the conceptual design of an experimental fast reactor (FFR), theoretical studies of FFR thermo-hydraulics and a design for an immersed sodium flowmeter. Reactor physics studies cover a range of topics including several related to work on zero power reactors. The section on reactor safety analysis is concerned largely with the assessment of established, and the presentation of new, computer codes for use in PWR safety calculations. Experimental and theoretical studies of fuels and reactor materials for FBRs, PWRs, BWRs and fusion reactors are described. A final miscellaneous section covers Mo-Tc isotope production in the swimming pool reactor, convective heat transfer in tubes and diffusion of tritium through plastic/aluminium composite films and Li 2 SiO 3 . (UK)

Nuclear power reactors

International Nuclear Information System (INIS)

1982-11-01

After an introduction and general explanation of nuclear power the following reactor types are described: magnox thermal reactor; advanced gas-cooled reactor (AGR); pressurised water reactor (PWR); fast reactors (sodium cooled); boiling water reactor (BWR); CANDU thermal reactor; steam generating heavy water reactor (SGHWR); high temperature reactor (HTR); Leningrad (RMBK) type water-cooled graphite moderated reactor. (U.K.)

Utilization of Stop-flow Micro-tubing Reactors for the Development of Organic Transformations.

Science.gov (United States)

Toh, Ren Wei; Li, Jie Sheng; Wu, Jie

2018-01-04

A new reaction screening technology for organic synthesis was recently demonstrated by combining elements from both continuous micro-flow and conventional batch reactors, coined stop-flow micro-tubing (SFMT) reactors. In SFMT, chemical reactions that require high pressure can be screened in parallel through a safer and convenient way. Cross-contamination, which is a common problem in reaction screening for continuous flow reactors, is avoided in SFMT. Moreover, the commercially available light-permeable micro-tubing can be incorporated into SFMT, serving as an excellent choice for light-mediated reactions due to a more effective uniform light exposure, compared to batch reactors. Overall, the SFMT reactor system is similar to continuous flow reactors and more superior than batch reactors for reactions that incorporate gas reagents and/or require light-illumination, which enables a simple but highly efficient reaction screening system. Furthermore, any successfully developed reaction in the SFMT reactor system can be conveniently translated to continuous-flow synthesis for large scale production.

Review of the treat upgrade reactor scram system reliability analysis

International Nuclear Information System (INIS)

Montague, D.F.; Fussell, J.B.; Krois, P.A.; Morelock, T.C.; Knee, H.E.; Manning, J.J.; Haas, P.M.; West, K.W.

1984-10-01

In order to resolve some key LMFBR safety issues, ANL personnel are modifying the TREAT reactor to handle much larger experiments. As a result of these modifications, the upgraded Treat reactor will not always operate in a self-limited mode. During certain experiments in the upgraded TREAT reactor, it is possible that the fuel could be damaged by overheating if, once the computer systems fail, the reactor scram system (RSS) fails on demand. To help ensure that the upgraded TREAT reactor is shut down when required, ANL personnel have designed a triply redundant RSS for the facility. The RSS is designed to meet three reliability goals: (1) a loss of capability failure probability of 10 -9 /demand (independent failures only); (2) an inadvertent shutdown probability of 10 -3 /experiment; and (3) protection agaist any known potential common cause failures. According to ANL's reliability analysis of the RSS, this system substantially meets these goals

Fusion reactor design studies: standard accounts for cost estimates

International Nuclear Information System (INIS)