WorldWideScience

Sample records for candu power plants

  1. CANDU 9 nuclear power plant simulator

    International Nuclear Information System (INIS)

    Kattan, M.; MacBeth, M.J.; Lam, K.

    1995-01-01

    Simulators are playing, an important role in the design and operations of CANDU reactors. They are used to analyze operating procedures under standard and upset conditions. The CANDU 9 nuclear power plant simulator is a low fidelity, near full scope capability simulator. It is designed to play an integral part in the design and verification of the control centre mock-up located in the AECL design office. It will also provide CANDU plant process dynamic data to the plant display system (PDS), distributed control system (DCS) and to the mock-up panel devices. The simulator model employs dynamic mathematical models of the various process and control components that make up a nuclear power plant. It provides the flexibility to add, remove or update user supplied component models. A block oriented process input is provided with the simulator. Individual blocks which represent independent algorithms of the model are linked together to generate the required overall plant model. As a design tool the simulator will be used for control strategy development, human factors studies (information access, readability, graphical display design, operability), analysis of overall plant control performance, tuning estimates for major control loops and commissioning strategy development. As a design evaluation tool, the simulator will be used to perform routine and non-routine procedures, practice 'what if' scenarios for operational strategy development, practice malfunction recovery procedures and verify human factors activities. This paper will describe the CANDU 9 plant simulator and demonstrate its implementation and proposed utility as a tool in the control system and control centre design of a CANDU 9 nuclear power plant. (author). 2 figs

  2. Periodic inspection of CANDU nuclear power plant containment components

    International Nuclear Information System (INIS)

    1989-09-01

    This Standard is one in a series intended to provide uniform requirements for CANDU nuclear power plants. It provides requirements for the periodic inspection of containment components including the containment pressure suppression systems

  3. Distributed control system for CANDU 9 nuclear power plant

    International Nuclear Information System (INIS)

    Harber, J.E.; Kattan, M.K.; Macbeth, M.J.

    1996-01-01

    Canadian designed CANDU pressurized heavy water nuclear reactors have been world leaders in electrical power generation. The CANDU 9 project is AECL's next reactor design. The CANDU 9 plant monitoring, annunciation, and control functions are implemented in two evolutionary systems; the distributed control system (DCS) and the plant display system (PDS). The CDS implements most of the plant control functions in a single hardware platform. The DCS communicates with the PDS to provide the main operator interface and annunciation capabilities of the previous control computer designs along with human interface enhancements required in a modern control system. (author)

  4. Emergency core cooling systems in CANDU nuclear power plants

    International Nuclear Information System (INIS)

    1981-12-01

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

  5. Stochastic maintenance optimization at Candu power plants

    International Nuclear Information System (INIS)

    Doyle, E.K.; Duchesne, T.; Lee, C.G.; Cho, D.I.

    2004-01-01

    The use of various innovative maintenance optimization techniques at Bruce has lead to cost effective preventive maintenance applications for complex systems as previously reported at ICONE 6 in New Orleans (1996). Further refinement of the station maintenance strategy was evaluated via the applicability of statistical analysis of historical failure data. The viability of stochastic methods in Candu maintenance was illustrated at ICONE 10 in Washington DC (2002). The next phase consists of investigating the validity of using subjective elicitation techniques to obtain component lifetime distributions. This technique provides access to the elusive failure statistics, the lack of which is often referred to in the literature as the principal impediment preventing the use of stochastic methods in large industry. At the same time the technique allows very valuable information to be captured from the fast retiring 'baby boom generation'. Initial indications have been quite positive. The current reality of global competition necessitates the pursuit of all financial optimizers. The next construction phase in the power generation industry will soon begin on a worldwide basis. With the relatively high initial capital cost of new nuclear generation all possible avenues of financial optimization must be evaluated and implemented. (authors)

  6. Low Power Shutdown PSA for CANDU Type Plants

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Yeon Kyoung; Kim, Myung Su [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    KHNP also have concentrated on full power PSA. Some recently constructed OPR1000 type plants and APR1400 type plants have performed the low power and shutdown (LPSD) PSA. The purpose of LPSD PSA is to identify the main contributors on the accident sequences of core damage and to find the measure of safety improvement. After the Fukushima accident, Korean regulatory agency required the shutdown severe accident management guidelines (SSAMG) development for safety enhancement. For the reliability of SSAMG, KHNP should develop the LPSD PSA. Especially, the LPSD PSA for CANDU type plant had developed for the first time in Korea. This paper illustrates how the LPSD PSA for CANDU type developed and the core damage frequency (CDF) is different with that of full power PSA. KHNP performed LPSD PSA to develop the SSAMG after the Fukushima accidents. The results show that risk at the specific operation mode during outage is higher than that of full power operation. Also, the results indicated that recovery failure of class 4 power at the POS 5A, 5B contribute dominantly to the total CDF from importances analysis. LPSD PSA results such as CDF with initiating events and POSs, risk results with plant damage state, and containment failure probability and frequency with POSs can be used by inputs for developing the SSAMG.

  7. Strategic provisioning of replacement parts for CANDU power plants

    International Nuclear Information System (INIS)

    Mizuno, G.; Tume, P.; Prentice, J.

    2000-01-01

    Provisioning of replacement parts and management of critical spares are key factors in optimizing maintenance programs for CANDU power plants. With a view to supply assurance, Atomic Energy Canada Limited (AECL) has created a Spare Parts Branch (SPB) to provide a clear pipeline from the client to the delivered replacement part(s). SPB provides the client with assured access to a qualified supplier database, computer aided design, engineering and manufacturing services and material upgrades and design registration through the authorized inspection agency. The AECL spare parts strategic provisioning service plan that has four thrusts: 1) the efficient delivery of cost-effective replacement parts; 2) obsolete parts resolution; 3) a website that will provide our clients with real-time access to replacement part data; and 4) inventory recovery opportunities. Thrusts one and two are actively ensuring plant maintenance for on-shore and off-shore CANDU clients. Thrusts three and four are longer-term commitments. This paper will explore these thrusts in the context of our CANDU business practices. (author)

  8. Integrated control centre concepts for CANDU power plants

    International Nuclear Information System (INIS)

    Lupton, L.R.; Davey, E.C.; Lapointe, P.A.; Shah, R.R.

    1990-01-01

    The size and complexity of nuclear power plants has increased significantly in the last 20 years. There is general agreement that plant safety and power production can be enhanced if more operational support systems that are significantly different from the ones based on the more conventional technologies used in plant control rooms. In particular, artificial intelligence and related technologies will play a major role in the development of innovative methods for information processing and presentation. These technologies must be integrated into the overall management and control philosophy of the plant and not be treated as vehicles to implement point solutions. The underlying philosophy behind our approach is discussed in this paper. Operator support systems will integrate into the overall control philosophy by complementing the operator. Four support systems are described; each is a prototype of a system being considered for the CANDU 3 control centre

  9. Integrated control centre concepts for CANDU power plants

    Energy Technology Data Exchange (ETDEWEB)

    Lupton, L. R.; Davey, E. C.; Lapointe, P. A.; Shah, R. R.

    1990-01-15

    The size and complexity of nuclear power plants has increased significantly in the last 20 years. There is general agreement that plant safety and power production can be enhanced if more operational support systems that are significantly different from the ones based on the more conventional technologies used in plant control rooms. In particular, artificial intelligence and related technologies will play a major role in the development of innovative methods for information processing and presentation. These technologies must be integrated into the overall management and control philosophy of the plant and not be treated as vehicles to implement point solutions. The underlying philosophy behind our approach is discussed in this paper. Operator support systems will integrate into the overall control philosophy by complementing the operator. Four support systems are described; each is a prototype of a system being considered for the CANDU 3 control centre.

  10. Future CANDU nuclear power plant design requirements document executive summary

    International Nuclear Information System (INIS)

    Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young; S. A. Usmani

    1996-03-01

    The future CANDU Requirements Document (FCRED) describes a clear and complete statement of utility requirements for the next generation of CANDU nuclear power plants including those in Korea. The requirements are based on proven technology of PHWR experience and are intended to be consistent with those specified in the current international requirement documents. Furthermore, these integrated set of design requirements, incorporate utility input to the extent currently available and assure a simple, robust and more forgiving design that enhances the performance and safety. The FCRED addresses the entire plant, including the nuclear steam supply system and the balance of the plant, up to the interface with the utility grid at the distribution side of the circuit breakers which connect the switchyard to the transmission lines. Requirements for processing of low level radioactive waste at the plant site and spent fuel storage requirements are included in the FCRED. Off-site waste disposal is beyond the scope of the FCRED. 2 tabs., 1 fig. (Author) .new

  11. Design requirements, criteria and methods for seismic qualification of CANDU power plants

    International Nuclear Information System (INIS)

    Singh, N.; Duff, C.G.

    1979-10-01

    This report describes the requirements and criteria for the seismic design and qualification of systems and equipment in CANDU nuclear power plants. Acceptable methods and techniques for seismic qualification of CANDU nuclear power plants to mitigate the effects or the consequences of earthquakes are also described. (auth)

  12. Passive safety features for next generation CANDU power plants

    International Nuclear Information System (INIS)

    Natalizio, A.; Hart, R.S.; Lipsett, J.J.; Soedijono, P.; Dick, J.E.

    1989-01-01

    CANDU offers an evolutionary approach to simpler and safer reactors. The CANDU 3, an advanced CANDU, currently in the detailed design stage, offers significant improvements in the areas of safety, design simplicity, constructibility, operability, maintainability, schedule and cost. These are being accomplished by retaining all of the well known CANDU benefits, and by relying on the use of proven components and technologies. A major safety benefit of CANDU is the moderator system which is separate from the coolant. The presence of a cold moderator reduces the consequences arising from a LOCA or loss of heat sink event. In existing CANDU plants even the severe accident - LOCA with failure of the emergency core cooling system - is a design basis event. Further advances toward a simpler and more passively safe reactor will be made using the same evolutionary approach. Building on the strength of the moderator system to mitigate against severe accidents, a passive moderator cooling system, depending only on the law of gravity to perform its function, will be the next step of development. AECL is currently investigating a number of other features that could be incorporated in future evolutionary CANDU designs to enhance protection against accidents, and to limit off-site consequences to an acceptable level, for even the worst event. The additional features being investigated include passive decay heat removal from the heat transport system, a simpler emergency core cooling system and a containment pressure suppression/venting capability for beyond design basis events. Central to these passive decay heat removal schemes is the availability of a short-term heat sink to provide a decay heat removal capability of at least three days, without any station services. Preliminary results from these investigations confirm the feasibility of these schemes. (author)

  13. Preliminary evaluation of licensing issues associated with U. S. -sited CANDU-PHW nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    van Erp, J B

    1977-12-01

    The principal safety-related characteristics of current CANDU-PHW power plants are described, and a distinction between those characteristics which are intrinsic to the CANDU-PHW system and those that are not is presented. An outline is given of the main features of the Canadian safety and licensing approach. Differences between the U.S. and Canadian approach to safety and licensing are discussed. Some of the main results of the safety analyses, routinely performed for CANDU-PHW reactors, are presented. U.S.-NRC General Design Criteria are evaluated as regards their applicability to CANDU-PHW reactors; vice-versa the CANDU-PHW reactor is evaluated with respect to its conformance to the U.S.-NRC General Design Criteria. A number of design modifications are proposed to be incorporated into the CANDU-PHW reactor in order to facilitate its introduction into the U.S.

  14. General requirements for pressure-retaining systems and components in CANDU nuclear power plants

    International Nuclear Information System (INIS)

    1991-11-01

    This standard specifies the general requirements for the design, fabrication and installation of pressure-retaining systems, components, and their supports in CANDU nuclear power plants. (16 figs., 2 tabs., 25 refs.)

  15. Experience with digital instrumentation and control systems for CANDU power plant modifications

    International Nuclear Information System (INIS)

    Basu, S.

    1997-01-01

    Over the last years, Ontario Hydro CANDU power plants have gone through many modifications. This includes modification from analog hardwired controls to digital and solid state controls and replacement of the existing digital controls with the latest hardware and software technology. Examples of digital modifications at Bruce A and other CANDU power plants are briefly described and categorized. Most of the I and C technology development has been supported by the CANDU Owners Group (COG) a consortium of Canadian nuclear utilities and the Atomic Energy Canada Limited (AECL). (author)

  16. Experience with digital instrumentation and control systems for CANDU power plant modifications

    Energy Technology Data Exchange (ETDEWEB)

    Basu, S [Ontario Hydro, Toronto, ON (Canada)

    1997-07-01

    Over the last years, Ontario Hydro CANDU power plants have gone through many modifications. This includes modification from analog hardwired controls to digital and solid state controls and replacement of the existing digital controls with the latest hardware and software technology. Examples of digital modifications at Bruce A and other CANDU power plants are briefly described and categorized. Most of the I and C technology development has been supported by the CANDU Owners Group (COG) a consortium of Canadian nuclear utilities and the Atomic Energy Canada Limited (AECL). (author).

  17. Requirements for the support power systems of CANDU nuclear power plants

    International Nuclear Information System (INIS)

    1990-08-01

    This Standard covers principal criteria and requirements for design, fabrication, installation, qualification, inspection, and documentation for assurance that support power will be available as required. The minimum requirements for support power are determined by the special safety systems and other safety-related systems that must function to ensure that the public health risk is acceptably low. Support power systems of a CANDU nuclear power plant include those parts of the electrical systems and instrument air systems that are necessary for the operation of safety-related systems

  18. Licensing evaluation of CANDU-PHW nuclear power plants relative to U.S. regulatory requirements

    International Nuclear Information System (INIS)

    Erp, J.B. van

    1978-01-01

    Differences between the U.S. and Canadian approach to safety and licensing are discussed. U.S. regulatory requirements are evaluated as regards their applicability to CANDU-PHW reactors; vice-versa the CANDU-PHW reactor is evaluated with respect to current Regulatory Requirements and Guides. A number of design modifications are proposed to be incorporated into the CANDU-PHW reactor in order to facilitate its introduction into the U.S. These modifications are proposed solely for the purpose of maintaining consistency within the current U.S. regulatory system and not out of a need to improve the safety of current-design CANDU-PHW nuclear power plants. A number of issues are identified which still require resolution. Most of these issues are concerned with design areas not (yet) covered by the ASME code. (author)

  19. Instrumentation and control systems for CANDU-PHW nuclear power plants

    International Nuclear Information System (INIS)

    Lepp, R.M.; Watkins, L.M.

    1982-02-01

    The instrumentation and control of CANDU nuclear power plants takes advantage of modern electronics technology in the extensive computerization of important control and man-machine functions. A description of these functions as well as those of the four Special Safety Systems is provided

  20. General requirements for concrete containment structures for CANDU nuclear power plants

    International Nuclear Information System (INIS)

    1993-07-01

    This standard provides the general requirements used in the design, construction, testing, and commissioning of concrete containment structures for CANDU nuclear power plants designated as class containment and is directed to the owners, designers, manufacturers, fabricators, and constructors of the concrete components and parts

  1. Study of candu fuel elements irradiated in a nuclear power plant

    International Nuclear Information System (INIS)

    Ionescu, S.; Uta, O.; Mincu, M.; Anghel, D.; Prisecaru, I.

    2015-01-01

    The object of this work is the behaviour of CANDU fuel elements after service in nuclear power plant. The results are analysed and compared with previous result obtained on unirradiated samples and with the results obtained on samples irradiated in the TRIGA reactor of INR Pitesti. Zircaloy-4 is the material used for CANDU fuel sheath. The importance of studying its behaviour results from the fact that the mechanical properties of the CANDU fuel sheath suffer modifications during normal and abnormal operation. In the nuclear reactor, the fuel elements endure dimensional and structural changes as well as cladding oxidation, hydriding and corrosion. These changes can lead to defects and even to the loss of integrity of the cladding. This paper presents the results of examinations performed in the Post Irradiation Examination Laboratory (PIEL) of INR Pitesti on samples from fuel elements after they were removed out of the nuclear power plant: - dimensional and macrostructural characterization; - microstructural characterization by metallographic analyses; - determination of mechanical properties; - fracture surface analysis by scanning electron microscopy (SEM). A full set of non-destructive and destructive examinations concerning the integrity, dimensional changes, oxidation, hydriding and mechanical properties of the cladding was performed. The obtained results are typical for CANDU 6-type fuel. The obtained data could be used to evaluate the security, reliability and nuclear fuel performance, and for the improvement of the CANDU fuel. (authors)

  2. Tricon hardware controller implementation of CANDU nuclear power plant shutdown system

    International Nuclear Information System (INIS)

    Zahedi, P.

    2007-01-01

    This paper introduces the implementation of logic functions associated with the shutdown systems of CANDU nuclear power plants. The experimental aspects of this work include development of control program embedded in shutdown systems of CANDU based NPPs. A physical test environment is designed to simulate the measurements of in-core flux detector (ICFD) and ion chamber (I/C) signals. The programmable logic used in this experimentation provides Triple Modular Redundant (TMR) architecture as well as a voting mechanism used upon execution of control program on each independent channel. (author)

  3. Preliminary evaluation of licensing issues associated with U.S.-sited CANDU-PHW nuclear power plants

    International Nuclear Information System (INIS)

    van Erp, J.B.

    1977-12-01

    The principal safety-related characteristics of current CANDU-PHW power plants are described, and a distinction between those characteristics which are intrinsic to the CANDU-PHW system and those that are not is presented. An outline is given of the main features of the Canadian safety and licensing approach. Differences between the U.S. and Canadian approach to safety and licensing are discussed. Some of the main results of the safety analyses, routinely performed for CANDU-PHW reactors, are presented. U.S.-NRC General Design Criteria are evaluated as regards their applicability to CANDU-PHW reactors; vice-versa the CANDU-PHW reactor is evaluated with respect to its conformance to the U.S.-NRC General Design Criteria. A number of design modifications are proposed to be incorporated into the CANDU-PHW reactor in order to facilitate its introduction into the U.S

  4. Advancement of safeguards inspection technology for CANDU nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Sung; Park, W S; Cha, H R; Ham, Y S; Lee, Y G; Kim, K P; Hong, Y D

    1999-04-01

    The objectives of this project are to develop both inspection technology and safeguards instruments, related to CANDU safeguards inspection, through international cooperation, so that those outcomes are to be applied in field inspections of national safeguards. Furthermore, those could contribute to the improvement of verification correctness of IAEA inspections. Considering the level of national inspection technology, it looked not possible to perform national inspections without the joint use of containment and surveillance equipment conjunction with the IAEA. In this connection, basic studies for the successful implementation of national inspections was performed, optimal structure of safeguards inspection was attained, and advancement of safeguards inspection technology was forwarded. The successful implementation of this project contributed to both the improvement of inspection technology on CANDU reactors and the implementation of national inspection to be performed according to the legal framework. In addition, it would be an opportunity to improve the ability of negotiating in equal shares in relation to the IAEA on the occasion of discussing or negotiating the safeguards issues concerned. Now that the national safeguards technology for CANDU reactors was developed, the safeguards criteria, procedure and instruments as to the other item facilities and fabrication facilities should be developed for the perfection of national inspections. It would be desirable that the recommendations proposed and concreted in this study, so as to both cope with the strengthened international safeguards and detect the undeclared nuclear activities, could be applied to national safeguards scheme. (author)

  5. Cernavoda nuclear power plant: Modifications in the fire protection measures of the CANDU 6 standard design

    International Nuclear Information System (INIS)

    Covalschi, V.

    1998-01-01

    Having as purpose the improvement of fire safety at the Cernavoda NPP - both in the prevention and the protection aspects in the case of fire - we implemented some modifications in the CANDU 6 standard design. These improvements are inspired, mainly, from two sources: the world-wide achievements in the field of fire protection techniques, introduced in nuclear power plants since the middle of 70's, when the CANDU 6 design was completed; the national practice and experience in fire protection, usually applied in industrial objectives (conventional power plants, in particular). The absence of any incident may be considered as a proof of the efficiency of the implemented fire preventing and protection measures. (author)

  6. Evolutionary CANDU 9 plant improvements

    International Nuclear Information System (INIS)

    Yu, S.K.W.

    1999-01-01

    The CANDU 9 is a 935 MW(e) nuclear power plant (NPP) based on the multi-unit Darlington and Bruce B designs with additional enhancements from our ongoing engineering and research programs. Added to the advantages of using proven systems and components, CANDU 9 offers improvement features with enhanced safety, improved operability and maintenance including a control centre with advanced man-machine interface, and improved project delivery in both engineering and construction. The CANDU 9 NPP design incorporated safety enhancements through careful attention to emerging licensing and safety issues. The designers assessed, revised and evolved such systems as the moderator, end shield, containment and emergency core cooling (ECC) systems while providing an integrated final design that is more passive and severe-accident-immune. AECL uses a feedback process to incorporate lessons learned from operating plants, from current projects experiences and from the implementation or construction phase of previous projects. Most of the requirements for design improvements are based on a systematic review of current operating CANDU stations in the areas of design and reliability, operability, and maintainability. The CANDU 9 Control Centre provides plant staff with improved operability and maintainability capabilities due to the combination of systematic design with human factors engineering and enhanced operating and diagnostics features. The use of advanced engineering tools and modem construction methods will reduce project implementation risk on project costs and schedules. (author)

  7. Initiating stochastic maintenance optimization at Candu Power Plants

    International Nuclear Information System (INIS)

    Doyle, E.K.

    2003-01-01

    As previously reported at ICONE 6 in New Orleans (1996), the use of various innovative maintenance optimization techniques at Bruce has lead to cost effective preventive maintenance applications for complex systems. Further cost refinement of the station maintenance strategy is being evaluated via the applicability of statistical analysis of historical failure data. Since the statistical evaluation was initiated in 1999 significant progress has been made in demonstrating the viability of stochastic methods in Candu maintenance. Some of the relevant results were presented at ICONE 10 in Washington DC (2002). Success with the graphical displays and the relatively easy to implement stochastic computer programs was sufficient to move the program along to the next significant phase. This next phase consists of investigating the validity of using subjective elicitation techniques to obtain component lifetime distributions. This technique provides access to the elusive failure statistics, the lack of which is often referred to in the literature as the principle impediment preventing the use of stochastic methods in large industry. At the same time the technique allows very valuable information to be captured from the fast retiring 'baby boom generation'. Initial indications have been quite positive. (author)

  8. Experience teaching CD-ROM-based course on CANDU nuclear-power-plant systems and operation

    International Nuclear Information System (INIS)

    Rouben, Benjamin

    2008-01-01

    This paper presents personal experience garnered from teaching a CD-ROM-based course on CANDU Power-Plant Systems and Operation. This course was originally developed by Prof. G.T. Bereznai as research in distance-learning techniques when he was directing the Thai-Canadian Human Resources Development Project at Chulalongkorn University in Bangkok. The course has been offered in a number of universities, including McMaster University and the University of Ontario Institute of Technology. All the course material, including lectures, assignments, and a simulator, is provided on a CD-ROM. Lectures include a spoken soundtrack covering the material. The class often includes both undergraduate and graduate students. I found that most students appreciate having the material on electronic format, which they can view and review at will and on their own time. Students find this course quite intensive - it covers all major systems in the CANDU reactor and power plant in detail. A very important component of the course is the simulator, which teaches students how systems operate in normal operation, in power manoeuvres, and during process-system malfunctions. Effort in absorbing the material and performing assignments can often exceed 10 hours per week. Some of the simulator assignments involve tricky manoeuvres, requiring several tries to achieve the expected result. Some assignments may take several hours, especially if the manoeuvres requiring repetition take 30 minutes or more in real time. I found that some instruction in the basic theory of reactor physics and systems is appreciated by students. A few possible enhancements to the simulator model were identified. Graduate students taking the course are required to do an additional project; I assigned an investigation of the effects of xenon-concentration changes during 1 week of load cycling. In summary, this course provides to students the opportunity to learn a great deal about the workings of CANDU-plant systems. (author)

  9. Multivariable controller for a 600 MWe CANDU nuclear power plant

    International Nuclear Information System (INIS)

    Mensah, S.

    1982-11-01

    The problems of designing a multivariable regulator for a nuclear power station of the Gentilly-2 type are studied. A reduced model, G2LDM, linearized around steady state operating conditions, is derived from the non-linear model G2SIM. The resulting linear model is described by state-space equations. Good agreement is demonstrated between the transient responses of both models. Properties of G2LDM are assessed by performing controllability and observability tests, cyclicity and rank tests, and eigenanalysis. A comprehensive set of application-orinented algorithms which allow multivariable controller design with closed-loop pole-assignment techniques are implemented in a computer-aided design package via several modules. A general scheme for the implementation of a multivariable controller in G2SIM is designed, and simulation tests show satisfactory performance of the controller [fr

  10. A study for good regulatin of the CANDU's in Korea. Development of safety regulatory requirement for CANDU nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Se Ki; Shin, Y. K.; Kim, J. S.; Yu, Y. J.; Lee, Y. J. [Ajou Univ., Suwon (Korea, Republic of)

    2001-03-15

    The objective of project is to derive the policy recommendations to improve the efficiency of CANDU plants regulation. These policy recommendations will eventually contribute to the upgrading of Korean nuclear regulatory system and safety enhancement. During the first phase of this 2 years study, following research activities were done. On-site survey and analysis on CANDU plants regulation. Review on CANDU plants regulating experiences and current constraints. Review and analysis on the new Canadian regulatory approach.

  11. The development of emergency core cooling systems in the PWR, BWR, and HWR Candu type of nuclear power plants

    International Nuclear Information System (INIS)

    Mursid Djokolelono.

    1976-01-01

    Emergency core cooling systems in the PWR, BWR, and HWR-Candu type of nuclear power plant are reviewed. In PWR and BWR the emergency cooling can be catagorized as active high pressure, active low pressure, and a passive one. The PWR uses components of the shutdown cooling system: whereas the BWR uses components of pressure suppression contaiment. HWR Candu also uses the shutdown cooling system similar to the PWR except some details coming out from moderator coolant separation and expensive cost of heavy water. (author)

  12. CANDU nuclear power system

    International Nuclear Information System (INIS)

    1981-01-01

    This report provides a summary of the components that make up a CANDU reactor. Major emphasis is placed on the CANDU 600 MW(e) design. The reasons for CANDU's performance and the inherent safety of the system are also discussed

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-12-31

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

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  15. Next generation CANDU plants

    International Nuclear Information System (INIS)

    Hedges, K.R.; Yu, S.K.W.

    1998-01-01

    Future CANDU designs will continue to meet the emerging design and performance requirements expected by the operating utilities. The next generation CANDU products will integrate new technologies into both the product features as well as into the engineering and construction work processes associated with delivering the products. The timely incorporation of advanced design features is the approach adopted for the development of the next generation of CANDU. AECL's current products consist of 700MW Class CANDU 6 and 900 MW Class CANDU 9. Evolutionary improvements are continuing with our CANDU products to enhance their adaptability to meet customers ever increasing need for higher output. Our key product drivers are for improved safety, environmental protection and improved cost effectiveness. Towards these goals we have made excellent progress in Research and Development and our investments are continuing in areas such as fuel channels and passive safety. Our long term focus is utilizing the fuel cycle flexibility of CANDU reactors as part of the long term energy mix

  16. Sensitivity Study for Feed and Bleed Operation for Domestic CANDU Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Kim, K. S; Kim, B. S.; Yoo, H. K.; Kim, H. J. [Atomic Creative Technology Co., Daejeon (Korea, Republic of); Whang, S. W. [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

    2016-05-15

    The purpose of this paper is to evaluate the effects of the feed and bleed operation using DCT(Degassing Condenser Tank) over-pressure protection valves when LOCL4 with LOCL3 occurs during LPSD mid-loop operation. The CDF(Core Damage Frequency) of Level-1 Internal Event for Wolsong NPP Unit 1 during LPSD POS5A/5B accounts for about 89.08%(POS5A: 30.04%, POS5B: 59.04%) of the total CDF. For Wolsong NPP Unit 1 LPSD External Event, seismic analysis is excluded from this study because it is PSA-based SMA(Seismic Margin Assessment based Probabilistic Safety Assessment). For the domestic CANDU NPP, the feed and bleed operation using DCT over-pressure protection valves has been incorporated as an additional measure to mitigate the consequences during LPSD mid-loop operation. Since LOCL4 with LOCL3 is considered to be the event with highest frequency among all initial events, the effect of the feed and bleed operation on the safety of Nuclear Power Plant has been evaluated using PSA methodology.

  17. Life extension, power upgrade, and return to service work for Pickering NGS and other PWR and CANDU plants

    International Nuclear Information System (INIS)

    Millman, J.; Idvorian, N.; Schneider, W.

    2002-01-01

    Work on life extension, power upgrade and return to service has been performed and is in progress for a number of PWR and CANDU plants. For PWR plants, power upgrade work has been done for the new replacement steam generators in several cases. This work consists of redoing the formal equipment qualification analysis and reports for the uprated operating conditions to support the application for license adjustment. Life extension assessments have been performed for several CANDU plants. These are highly detailed assessments in which the particular steam generator is reassessed part by part as to the ability of each to sustain full life operation and also extended life operation. Return to service work for Pickering NGSA specifically has included this type of assessment and also specific repair, cleaning and retrofit activities including secondary side inspection, waterlancing, divider plate repair, eddy current inspection, etc. Steam generator modifications and retrofit work have been performed in a number of cases. The paper discusses various life extension, power upgrade, equipment modification and return to service activities all of which are part of the renewed drive in the industry to realise the full potential of nuclear plants by getting more and better performance from the extended service of existing plants. (author)

  18. On the speed of response of an FPGA-based shutdown system in CANDU nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    She Jingke, E-mail: jshe2@uwo.ca [Department of Electrical and Computer Engineering, The University of Western Ontario, London, Ontario, N6A 5B9 (Canada); Jiang Jin, E-mail: jjiang@eng.uwo.ca [Department of Electrical and Computer Engineering, The University of Western Ontario, London, Ontario, N6A 5B9 (Canada)

    2011-06-15

    Highlights: > Design and implementation of an FPGA-based CANDU SDS1. > Hardware-in-the-loop simulation for performance evaluation involved with an NPP simulator. > Comparison of the response time between FPGA-based trip channel and software-based PLC. - Abstract: Several issues in an FPGA based implementation of shutdown systems in CANDU nuclear power plants have been investigated in this paper. A particular attention is on the response time of an FPGA implementation of safety shutdown systems in comparison with operating system based software solutions as in existing CANDU plants. The trip decision logic under 'steam generator (SG) level low' condition has been examined in detail. The design and implementation of this logic on an FPGA platform have been carried out. The functionality tests are performed in a hardware-in-the-loop (HIL) environment by connecting the FPGA based system to an NPP simulator, and replacing one channel of Shutdown System Number 1 (SDS1) in the simulator by the FPGA implementation. The response time of the designed system is also measured through multiple tests under different conditions, and statistical data analysis has been performed. The results of the response time tests are compared against those of a software-based implementation of the same trip logic.

  19. Development situation about the Canadian CANDU Nuclear Power Generating Stations

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Yu Mi; Kim, Yong Hee; Park, Joo Hwan

    2009-07-15

    The CANDU reactor is the most versatile commercial power reactor in the world. The acronym 'CANDU', a registered trademark of Atomic Energy of Canada Limited, stands for 'CANada Deuterium Uranium'. CANDU uses heavy water as moderator and uranium (originally, natural uranium) as fuel. All current power reactors in Canada are of the CANDU type. Canada exports CANDU type reactor in abroad. CANDU type is used as the nuclear power plants to produce electrical. Today, there are 41 CANDU reactors in use around the world, and the design has continuously evolved to maintain into unique technology and performance. The CANDU-6 power reactor offers a combination of proven, superior and state-of-the-art technology. CANDU-6 was designed specifically for electricity production, unlike other major reactor types. One of its characteristics is a very high operating and fuel efficiency. Canada Nuclear Power Generating Stations were succeeded in a commercial reactor of which the successful application of heavy water reactor, natural uranium method and that on-power fuelling could be achieved. It was achieved through the joint development of a major project by strong support of the federal government, public utilities and private enterprises. The potential for customization to any country's needs, with competitive development and within any level of domestic industrial infrastructure, gives CANDU technology strategic importance in the 21st century.

  20. Development situation about the Canadian CANDU Nuclear Power Generating Stations

    International Nuclear Information System (INIS)

    Jeon, Yu Mi; Kim, Yong Hee; Park, Joo Hwan

    2009-07-01

    The CANDU reactor is the most versatile commercial power reactor in the world. The acronym 'CANDU', a registered trademark of Atomic Energy of Canada Limited, stands for 'CANada Deuterium Uranium'. CANDU uses heavy water as moderator and uranium (originally, natural uranium) as fuel. All current power reactors in Canada are of the CANDU type. Canada exports CANDU type reactor in abroad. CANDU type is used as the nuclear power plants to produce electrical. Today, there are 41 CANDU reactors in use around the world, and the design has continuously evolved to maintain into unique technology and performance. The CANDU-6 power reactor offers a combination of proven, superior and state-of-the-art technology. CANDU-6 was designed specifically for electricity production, unlike other major reactor types. One of its characteristics is a very high operating and fuel efficiency. Canada Nuclear Power Generating Stations were succeeded in a commercial reactor of which the successful application of heavy water reactor, natural uranium method and that on-power fuelling could be achieved. It was achieved through the joint development of a major project by strong support of the federal government, public utilities and private enterprises. The potential for customization to any country's needs, with competitive development and within any level of domestic industrial infrastructure, gives CANDU technology strategic importance in the 21st century

  1. A foundation for allocating control functions to humans and machines in future CANDU nuclear power plants

    International Nuclear Information System (INIS)

    Lupton, L.R.; Lipsett, J.J.; Davey, E.C.; Olmstead, R.A.

    1990-06-01

    Since the control room for the Atomic Energy of Canada Limited CANDU 6 plant was designed in the 1970s, requirements for control rooms have changed dramatically as a result of new licensing requirements, evolution of major new standards for control centre design and technological advances. The role of the human operator has become prominent in the design and operation of industrial and, in particular, nuclear plants. Major industrial accidents in the last decade have highlighted the need for paying significantly more attention to the requirements of the human as an integral part of the plant control system. A Functional Design Methodology has been defined that addresses the issues related to maximizing the strengths of the human and the machine in the next generation of CANDU plants. This method is based, in part, on the recently issued international standard IEC 964. The application of this method will lead to the definition of the requirements for detailed design of the control room, including man-machine interfaces, preliminary operating procedures, staffing and training. Further, it provides a basis for the verification and validation of the allocation of functions to the operator and the machine

  2. Requirements for class 1, 2, and 3 pressure-retaining systems and components in CANDU nuclear power plants

    International Nuclear Information System (INIS)

    1991-09-01

    This third edition of CAN/CSA-N285.1 supersedes the 1981 and 1975 editions. It provides the specific requirements for design, fabrication, and installation of Class 1, 2 and 3 pressure-retaining systems and components in CANDU nuclear power plants, and over pressure protection of the heat transport system. The general requirements for pressure-retaining systems and components are given in CSA Standard CAN/CSA-N285.0, with which Class 1, 2 and 3 systems and components must also comply

  3. The transition criteria of circulating flow pattern of moderator in the calandria tank of CANDU nuclear power plant

    International Nuclear Information System (INIS)

    Jung, Yun Sik; Lee, Jae Young; Kim, Man Woong

    2004-01-01

    The moderator cooling system to the Calandria tank of CANDU nuclear power plant provides an alternative pass of heat sink during the hypothetical loss of coolant accident. Also, the neutron population in the CANDU plant can be affected by the moderator temperature change which strongly depends on the circulating flow pattern in the Calandria tank. It has been known that there are three distinguished flow patterns: the buoyancy dominated flow, the momentum dominated flow, and the mixed type flow. The Canadian Nuclear Safety Commission (CNSC) recommended that a series of experimental works should be performed to verify the three dimensional codes. Two existing facilities, SPEL (1982) and STERN (1990), have produced experimental data for these purposes. The present work is also motivated to build up a new scaled experimental facility named HGU for the same purposes. CANDU-6 was selected as the target plant to be scaled down. In the design for the scaled facility, the knowledge on the flow regime transitions in the circulating flow was imperative. In the present study, to pave the way for the scaling, the flow pattern maps of circulating flow were constructed based on the Reynolds number and Archimedes number. The CFX code was employed with real meshes to represent all calandria tubes in the tank. The flow pattern maps were constructed for SPEL, STERN, HGU, and CANDU6. As the key transition criterion useful for scaling law, a new Archimedes number considering the jet impingement of the feed water in the Calandria tank was found. The transition of flow patterns was made with the same Archimedes number for CANDU6, STERN and HGU. However, SPEL which has third of the modified Archimedes number showed different maps in the wider region of mixed flow pattern was observed. It was found that the Archimedes number considering the inlet nozzle velocity plays the key role in patterns classification. Also, it can be suggested that the moderator cooling system needs to be designed

  4. Thermal stability of chloroform in the steam condensate cycle of CANDU-PHW nuclear power plant

    International Nuclear Information System (INIS)

    Lepine, Louis; Gilbert, Roland; Ouellet, Lorenzo

    1992-01-01

    Analysis of samples taken at the Gentilly 2 (Quebec) CANDU-PHW (CANadian Deuterium Uranium - Pressurized Heavy Water) plant after chlorination and demineralization revealed the presence of all four trihalomethanes (THMs) (CHCl 3 , CHBrCl 2 , CHBr 2 Cl and CHBr 3 ) and other unidentified halogenated volatile compounds. Among the THMs, chloroform was the major contaminant. A study of its thermal stability in water at different temperatures confirmed the degradation of the CHCl 3 molecule according to the equation CHCl 3 + H2O → CO + 3 HCl. The reaction follows first order kinetics and has an activation energy of 100 kJ/mol. The estimated half-life is six seconds at 260 deg C, the maximum temperature of the steam condensate cycle

  5. CANDU plant maintenance: Recent developments

    International Nuclear Information System (INIS)

    Charlebois, P.

    2000-01-01

    CANDU units have long been recognized for their exceptional safety and reliability. Continuing development in the maintenance area has played a key role in achieving this performance level. For over two decades, safety system availability has been monitored closely and system maintenance programs adjusted accordingly to maintain high levels of performance. But as the plants approach mid life in a more competitive environment and component aging becomes a concern, new methods and techniques are necessary. As a result, recent developments are moving the maintenance program largely from a corrective and preventive approach to predictive and condition based maintenance. The application of these techniques is also being extended to safety related systems. These recent developments include use of reliability centred methods to define system maintenance requirements and strategies. This approach has been implemented on a number of systems at Canadian CANDU plants with positive results. The pilot projects demonstrated that the overall maintenance effort remained relatively constant while the system performance improved. It was also possible to schedule some of the redundant component maintenance during plant operation without adverse impact on system availability. The probabilistic safety assessment was found to be useful in determining the safety implications of component outages. These new maintenance strategies are now making use of predictive and condition based maintenance techniques to anticipate equipment breakdown and schedule preventive maintenance as the need arises rather than time based. Some of these techniques include valve diagnostics, vibration monitoring, oil analysis, thermography. Of course, these tools and techniques must form part of an overall maintenance management system to ensure that maintenance becomes a living program. To facilitate this process and contain costs, new information technology tools are being introduced to provide system engineers

  6. Seismic sensitivity study of a generic CANDU nuclear power plant: Soil-structure interaction

    International Nuclear Information System (INIS)

    Lee, L.S.S.; Duff, C.G.

    1983-01-01

    The seismic sensitivity and capability study for a generic CANDU Plant is part of an overall development program of design standardization. The purpose of this paper is to investigate the sensitivities of structural responses and floor response spectra (FRS) to variations of structural and soil parameters. In the seismic design standardization, a wide range of soil conditions is considered and the envelopes of the resulting site spectra (soil-structure interaction effect) are then used for the design of the generic plant. The nuclear island structures considered herein have different relative stiffness and one of them has two layout/structure schemes: one is relatively flexible and the other is moderately stiff. In the preliminary phase of the seismic sensitivity study presented hereby, the soil-structure interaction seismic analysis is based on the half-space modelling (soil-spring lumped-mass) method and the response spectrum method for the seismic responses. Distinct patterns and sensitivity of the site spectrum analysis for structure schemes of different relative stiffness and for different structural elevations are observed and discussed. (orig.)

  7. The seismic fragility analysis for multi-story steel structure in CANDU nuclear power plant

    International Nuclear Information System (INIS)

    Hwang, K.H.; Lee, B.S.; Kang, S-K.

    1996-01-01

    The Wolsong Unit 2 is a CANDU-6 type plant and is being constructed in the Wolsong site, where Design Basis Earthquake (DBE) was determined to be 0.2g. A seismic PSA for Wolsong Unit 2 is being performed as one of the conditions for the Construction Permit. One of the issues in the seismic PSA is the availability of the seismically non-qualified systems, which are located in the Turbine Building(T/B). Thus, the seismic fragility analysis for the T/B was performed to estimate the operability of the systems. The design seismic loads for the building were based on a ground response spectrum scaled down from the DBE to horizontal peak ground acceleration (pga) of 0.05g. The seismic fragility analysis for the building was performed using a factor of the safety method. It is estimated that the most critical failure is that of masonry walls and its High Confidence and Low Probability of Failure (HCLPF) capacity is 0.13g. The critical failure mode of the structure is identified to be tensile yielding failure of grip angle, and its HCLPF capacity is 0.34g. (author)

  8. Assessment and management of ageing of major nuclear power plant components important to safety: CANDU reactor assemblies

    International Nuclear Information System (INIS)

    2001-02-01

    At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of the major NPP components (e.g. caused by unanticipated phenomena and by operating, maintenance, design or manufacturing errors) can jeopardize plant safety and also plant life. Ageing in these NPPs must therefore be effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wearout of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This TECDOC is one in a series of reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers, and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety which was issued by the IAEA in 1992. The current practices for the assessment of safety margins (fitness for service) and the inspection, monitoring, and mitigation of ageing degradation of selected components of Canada deuterium-uranium (CANDU) reactors, boiling water reactors (BWRs), pressurized water reactors (PWRs) including the Soviet designed water moderated and water cooled energy reactors (WWERs), are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs and also to provide a common technical basis for dialogue between plant operators and regulators when dealing with age-related licensing issues. Since the reports are written from a safety perspective, they do not address life or life-cycle management of the plant components, which

  9. Tritium inventory prediction in a CANDU plant

    International Nuclear Information System (INIS)

    Song, M.J.; Son, S.H.; Jang, C.H.

    1995-01-01

    The flow of tritium in a CANDU nuclear power plant was modeled to predict tritium activity build-up. Predictions were generally in good agreement with field measurements for the period 1983--1994. Fractional contributions of coolant and moderator systems to the environmental tritium release were calculated by least square analysis using field data from the Wolsong plant. From the analysis, it was found that: (1) about 94% of tritiated heavy water loss came from the coolant system; (2) however, about 64% of environmental tritium release came from the moderator system. Predictions of environmental tritium release were also in good agreement with field data from a few other CANDU plants. The model was used to calculate future tritium build-up and environmental tritium release at Wolsong site, Korea, where one unit is operating and three more units are under construction. The model predicts the tritium inventory at Wolsong site to increase steadily until it reaches the maximum of 66.3 MCi in the year 2026. The model also predicts the tritium release rate to reach a maximum of 79 KCi/yr in the year 2012. To reduce the tritium inventory at Wolsong site, construction of a tritium removal facility (TRF) is under consideration. The maximum needed TRF capacity of 8.7 MCi/yr was calculated to maintain tritium concentration effectively in CANDU reactors

  10. CANDU plant life management - An integrated approach

    International Nuclear Information System (INIS)

    Hopkins, J.R.

    1998-01-01

    An integrated approach to plant life management has been developed for CANDU reactors. Strategies, methods, and procedures have been developed for assessment of critical systems structures and components and for implementing a reliability centred maintenance program. A Technology Watch program is being implemented to eliminate 'surprises'. Specific work has been identified for 1998. AECL is working on the integrated program with CANDU owners and seeks participation from other CANDU owners

  11. Aspects regarding the lifetime of a fuel channel in a CANDU nuclear power plant

    International Nuclear Information System (INIS)

    Calinescu, A.

    1998-01-01

    The paper presents the analysis of factors influencing upon the time life of a fuel channel of CANDU reactors built in Romania. Fuel channels are made of Zr-2.5%Nb alloy. Means and methodology to detect cracking of fuel channels are described, as well as improvements to increase life time of Cernavoda NPP fuel channels and national programme in this area. (author)

  12. CANDU fuel deposits and chemistry optimizations. Recent regulatory experience in Canadian Nuclear Power Plants

    International Nuclear Information System (INIS)

    Kameswaran, Ram

    2014-01-01

    Water chemistry of the Primary Heat Transport System (PHT) of CANDU – Pressurised Heavy Water Reactors profoundly influences the transport of corrosion products around the Heat Transport System (HTS), where they can be deposited as crud on steam generators, feeder pipes and on the fuel. Fuel cladding can be covered with deposits which have precipitated from the coolant as a result of temperature changes or non-optimal coolant pH. Precipitation of deposits in-core must be avoided as far as possible, as it leads to fouling of the fuel, loss of heat transfer efficiency, and increased radiation fields. In the recent years a Canadian NPP experienced increased instances of black deposits being observed on fuel bundles discharged from one of the units. The black deposits were initially observed in 2008 during in-bay fuel inspections. Since then it has been determined that all the discharged fuel bundles have black deposits on them and that observed deposits have been increasing in size (thickness and surface area). This negative trend has persisted through to 2012, when one of fuel bundles was observed with significantly larger deposit than previously seen. Initial analysis of the deposit indicated it to be iron oxide (magnetite). Flow Accelerated Corrosion (FAC) of carbon steel feeder pipes is the primary source of iron, which deposits as magnetite on HTS surfaces. The black deposits have predominantly been located immediately downstream of the bearing pads of the fuel bundle. Deposits have also tended to form on the bottom-downstream quadrant of the fuel bundles. The deposits were most prevalent in low power channels, but some deposits have been observed on high power channels. It was reported by the utility that the PHT system chemistry has been maintained in specification for most of the time during normal operation but the chemistry control during outages was inadequate. Due to design constraints, purification circuit was not available during outages and ion

  13. Probabilistic seismic safety assessment of a CANDU 6 nuclear power plant including ambient vibration tests: Case study

    Energy Technology Data Exchange (ETDEWEB)

    Nour, Ali [Hydro Québec, Montréal, Québec H2L4P5 (Canada); École Polytechnique de Montréal, Montréal, Québec H3C3A7 (Canada); Cherfaoui, Abdelhalim; Gocevski, Vladimir [Hydro Québec, Montréal, Québec H2L4P5 (Canada); Léger, Pierre [École Polytechnique de Montréal, Montréal, Québec H3C3A7 (Canada)

    2016-08-01

    Highlights: • In this case study, the seismic PSA methodology adopted for a CANDU 6 is presented. • Ambient vibrations testing to calibrate a 3D FEM and to reduce uncertainties is performed. • Procedure for the development of FRS for the RB considering wave incoherency effect is proposed. • Seismic fragility analysis for the RB is presented. - Abstract: Following the 2011 Fukushima Daiichi nuclear accident in Japan there is a worldwide interest in reducing uncertainties in seismic safety assessment of existing nuclear power plant (NPP). Within the scope of a Canadian refurbishment project of a CANDU 6 (NPP) put in service in 1983, structures and equipment must sustain a new seismic demand characterised by the uniform hazard spectrum (UHS) obtained from a site specific study defined for a return period of 1/10,000 years. This UHS exhibits larger spectral ordinates in the high-frequency range than those used in design. To reduce modeling uncertainties as part of a seismic probabilistic safety assessment (PSA), Hydro-Québec developed a procedure using ambient vibrations testing to calibrate a detailed 3D finite element model (FEM) of the containment and reactor building (RB). This calibrated FE model is then used for generating floor response spectra (FRS) based on ground motion time histories compatible with the UHS. Seismic fragility analyses of the reactor building (RB) and structural components are also performed in the context of a case study. Because the RB is founded on a large circular raft, it is possible to consider the effect of the seismic wave incoherency to filter out the high-frequency content, mainly above 10 Hz, using the incoherency transfer function (ITF) method. This allows reducing significantly the non-necessary conservatism in resulting FRS, an important issue for an existing NPP. The proposed case study, and related methodology using ambient vibration testing, is particularly useful to engineers involved in seismic re-evaluation of

  14. Safety systems and safety analysis of the Qinshan phase III CANDU nuclear power plant

    International Nuclear Information System (INIS)

    Cai Jianping; Shen Sen; Barkman, N.

    1999-01-01

    The author introduces the Canadian nuclear reactor safety philosophy and the Qinshan Phase III CANDU NPP safety systems and safety analysis, which are designed and performed according to this philosophy. The concept of 'defence-in-depth' is a key element of the Canadian nuclear reactor safety philosophy. The design concepts of redundancy, diversity, separation, equipment qualification, quality assurance, and use of appropriate design codes and standards are adopted in the design. Four special safety systems as well as a set of reliable safety support systems are incorporated in the design of Qinshan phase III CANDU for accident mitigation. The assessment results for safety systems performance show that the fundamental safety criteria for public dose, and integrity of fuel, channels and the reactor building, are satisfied

  15. Design of a multivariable controller for a CANDU 600 MWe nuclear power plant using the INA method

    International Nuclear Information System (INIS)

    Roy, N.; Boisvert, J.; Mensah, S.

    1984-04-01

    The development of large and complex nuclear and process plants requires high-performance control systems, designed with rigorous multivariable techniques. This work is part of an analytical study demonstrating the real potential of multivariable methods. It covers every step in the design of a multi-variable controller for a Gentilly-2 type CANDU 600 MWe nuclear power plant using the Inverse Nyquist Array (INA) method. First the linear design model and its preliminary modifications are described. The design tools are reviewed and the operations required to achieve open-loop diagonal dominance are thoroughly described. Analysis of the closed-loop system is then performed and a feedback matrix is selected to meet the design specifications. The performance of the controller on the linear model is verified by simulation. Finally, the controller is implemented on the reference non-linear model to assess its overall performance. The results show that the INA method can be used successfully to design controllers for large and complex systems

  16. CANDU plant life management - An integrated approach

    International Nuclear Information System (INIS)

    Charlebois, P.; Hart, R.S.; Hopkins, J.R.

    1998-01-01

    Commercial versions of CANDU reactors were put into service starting more than 25 years ago. The first unit of Ontario Hydro's Pickering A station was put into service in 1971, and Bruce A in 1977. Most CANDU reactors, however, are only now approaching their mid-life of 15 to 20 years of operation. In particular, the first series of CANDU 6 plants which entered service in the early 1980's were designed for a 30 year life and are now approaching mid life. The current CANDU 6 design is based on a 40 year life as a result of advancement in design and materials through research and development. In order to assure safe and economic operation of these reactors, a comprehensive CANDU Plant Life Management (PLIM) program is being developed from the knowledge gained during the operation of Ontario Hydro's Pickering, Bruce, and Darlington stations, worldwide information from CANDU 6 stations, CANDU research and development programs, and other national and international sources. This integration began its first phase in 1994, with the identification of most of the critical systems structures and components in these stations, and a preliminary assessment of degradation and mechanisms that could affect their fitness for service for their planned life. Most of these preliminary assessments are now complete, together with the production of the first iteration of Life Management Plans for several of the systems and components. The Generic CANDU 6 PLIM program is now reaching its maturity, with formal processes to systematically identify and evaluate the major CSSCs in the station, and a plan to ensure that the plant surveillance, operation, and maintenance programs monitor and control component degradation well within the original design specifications essential for the plant life attainment. A Technology Watch program is being established to ensure that degradation mechanisms which could impact on plant life are promptly investigated and mitigating programs established. The

  17. Methodology for identifying boundaries of systems important to safety in CANDU nuclear power plants

    International Nuclear Information System (INIS)

    Therrien, S.; Komljenovic, D.; Therrien, P.; Ruest, C.; Prevost, P.; Vaillancourt, R.

    2007-01-01

    This paper presents a methodology developed to identify the boundaries of the systems important to safety (SIS) at the Gentilly-2 Nuclear Power Plant (NPP), Hydro-Quebec. The SIS boundaries identification considers nuclear safety only. Components that are not identified as important to safety are systematically identified as related to safety. A global assessment process such as WANO/INPO AP-913 'Equipment Reliability Process' will be needed to implement adequate changes in the management rules of those components. The paper depicts results in applying the methodology to the Shutdown Systems 1 and 2 (SDS 1, 2), and to the Emergency Core Cooling System (ECCS). This validation process enabled fine tuning the methodology, performing a better estimate of the effort required to evaluate a system, and identifying components important to safety of these systems. (author)

  18. Support analysis for safety analysis development for CANDU nuclear power plant

    International Nuclear Information System (INIS)

    Bedreaga, L.; Florescu, Gh.; Apostol, M.; Nitoi, M.

    2004-01-01

    Probabilistic Safety Assessment analysis (PSA) is a technique used to assess the safety of a nuclear power plant. Assessments of the nuclear plant systems/components from safety point of view consist in accomplishment of a lot of support analyses that are the base for the main analysis, in order to evaluate the impact of occurrences of abnormal states for these systems. Evaluation of initiating events frequency and components failure rate is based on underlying probabilistic theory and mathematic statistics. Some of these analyses are detailed analyses and are known very well in PSA. There are also some analyses, named support analyses for PSA, which are very important but less applicable because they involve a huge human effort and hardware facilities to accomplish. The usual methods applicable in PSA such as input data extracted from the specific documentation (operation procedures, testing procedures, maintenance procedures and so on) or conservative evaluation provide a high level of uncertainty for both input and output data. The paper describes support analysis required to improve the certainty level in evaluation of reliability parameters and also in the final results (either risk, reliability or safety assessment). (author)

  19. Approach for seismic risk analysis for CANDU plants in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Lee, B-S; Kim, T; Kang, S-K [Korea Power Engineering Co., Seoul (Korea, Republic of); Hong, S-Y; Roh, S-R [Korea Electric Power Corp., Taejon (Korea, Republic of). Research Centre

    1996-12-31

    A seismic risk analysis for CANDU type plants has never been performed. The study presented here suggested that the approach generally applied to LWR type plants could lead to unacceptable result, if directly applied to CANDU plants. This paper presents a modified approach for the seismic risk analysis of CANDU plants. (author). 5 refs., 2 tabs., 2 figs.

  20. Corrosion control in CANDU nuclear power reactors

    International Nuclear Information System (INIS)

    Lesurf, J.E.

    1974-01-01

    Corrosion control in CANDU reactors which use pressurized heavy water (PHW) and boiling light water (BLW) coolants is discussed. Discussions are included on pressure tubes, primary water chemistry, fuel sheath oxidation and hydriding, and crud transport. It is noted that corrosion has not been a significant problem in CANDU nuclear power reactors which is a tribute to design, material selection, and chemistry control. This is particularly notable at the Pickering Nuclear Generating Station which will have four CANDU-PHW reactors of 540 MWe each. The net capacity factor for Pickering-I from first full power (May 1971) to March 1972 was 79.5 percent, and for Pickering II (first full power November 1971) to March 1972 was 83.5 percent. Pickering III has just reached full power operation (May 1972) and Pickering IV is still under construction. Gentilly CANDU-BLW reached full power operation in May 1972 after extensive commissioning tests at lower power levels with no major corrosion or chemistry problems appearing. Experience and operating data confirm that the value of careful attention to all aspects of corrosion control and augur well for future CANDU reactors. (U.S.)

  1. Value added services to CANDU plants

    International Nuclear Information System (INIS)

    Kakaria, B.K.

    2003-01-01

    Over the last decade or so, nuclear power plants, just like other types of electricity generating plants, have been facing a number of challenges. Depending on the operating environment of the utility, these challenges are forcing plant owners to examine all facets of the operating costs. Privatization, deregulation and economics of alternative electricity generation methods are exerting enormous pressure on nuclear power plants to streamline costs and improve their operational performance. CANDU reactors are no exception to these forces and face similar pressures. In particular, operating plants that are contemplating plant life extensions are being required to clearly demonstrate the economics of continued operation over other forms of power generation available to the utility. Improvement of capacity factors has the effect of increasing the revenues from the plant and as these revenues increase, the fixed portion of the plant costs including OM and A costs become a smaller percentage of the total revenues. Similar results can be achieved by aiming to reduce the plant OM and A costs. In reality, most well-planned intervention schemes directed at reducing OM and A costs tend to also increase the plant availability. Following plant turnover after commissioning, AECL has been supporting the CANDU owners and utilities with an assortment of products and services dealing with plant operations and outage management issues. AECL has taken the lead in arranging specialized resources, products and services by teaming with other complementary organizations to provide a complete suite of services. Recent examples of such support to operating CANDU plants will be described in the paper. AECL is responding to this changing business environment in two important ways. First, AECL is changing from simply providing a service to its clients towards providing value, something much more important. To this end, AECL is looking to other organizations to form alliances, partnerships and

  2. Used fuel packing plant for CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Menzies, I.; Thayer, B.; Bains, N., E-mail: imenzies@atsautomation.com [ATS Automation, Cambridge, ON (Canada); Murchison, A., E-mail: amurchison@nwmo.ca [NWMO, Toronto, ON (Canada)

    2015-07-01

    Large forgings have been selected to containerize Light Water Reactor used nuclear fuel. CANDU fuel, which is significantly smaller in size, allows novel approaches for containerization. For example, by utilizing commercially available extruded ASME pipe a conceptual design of a Used Fuel Packing Plant for containerization of used CANDU fuel in a long lived metallic container has been developed. The design adopts a modular approach with multiple independent work cells to transfer and containerize the used fuel. Based on current technologies and concepts from proven industrial systems, the Used Fuel Packing Plant can assemble twelve used fuel containers per day considering conservative levels of process availability. (author)

  3. Radionuclide Release after LBLOCA with Loss of Class IV Power Accident in CANDU-6 Plant

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hoon [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2011-10-15

    A large break in a pipe train of a primary heat transport system discharges coolant, which has high energy and large mass, into the containment building. Reactor shutdown and emergency core cooling water will limit the fuel cladding failure, but cannot prevent it entirely. The containment building is the last barrier of radionuclide release to the environment. Containment isolation and pressure suppression by dousing and local air cooler reduce the amount of radionuclide release to the environment. The objective of containment behavior analysis for large break loss of coolant with loss of class IV power accident is to assess the amount of radionuclide release to the ambient atmosphere. Radionuclide release rates in this event, with all safety system available, that is, the containment building is intact, as well as with containment system impairment, are analyzed with GOTHIC and SMART code

  4. Design of multivariable controller for a 600 MWe CANDU nuclear power plant

    International Nuclear Information System (INIS)

    Mensah, S.; McMorran, P.D.

    1982-04-01

    This paper reports the results of a case study on the design of a multivariable regulator for a nuclear power station of the Gentilly-2 type. In this study, a design model was derived by simplifying and linearizing equations in the G2SIM non-linear model. Open-loop simulation showed good agreement between transient responses of both models. After a critical review of multivariable design techniques, the authors explored pole shifting with output feedback. A comprehensive set of application-oriented algorithms for closed-loop pole shifting, implemented via modules in the MVPACK computer-aided design package were derived. A controller was designed for the linear model, then implemented on the non-linear simulation. After adjustment of controller gains, mainly in the dynanamic section of the feedback, simulation results showed that the performance of the multivariable controller on G2SIM is satisfactory. The results demonstrate the relative superiority of the multi-variable controller over the existing conventional controller

  5. CANDU 9 operator plant display system

    International Nuclear Information System (INIS)

    Trueman, R.; Webster, A.; MacBeth, M.J.

    1997-01-01

    To meet evolving client and regulatory needs, AECL has adopted an evolutionary approach to the design of the CANDU 9 control centre. That is, the design incorporates feedback from existing stations, reflects the growing diversity in the roles and responsibilities of the operating staff, and reduces costs associated with plant capital and operations, maintenance and administration (OM and A), through the appropriate introduction of new technologies. Underlying this approach is a refined engineering design process that cost-effectively integrates operational feedback and human factors engineering to define the operating staff information and information presentation requirements. Based on this approach, the CANDU 9 control centre will provide utility operating staff with the means to achieve improved operations and reduced OM and A costs. One of the design features that will contribute to the improved operational capabilities of the control centre is a new Plant Display System (PDS) that is separate from the digital control system. The PDS will be used to implement non-safety panel, and console video display systems within the CANDU 9 main control room (MCR). This paper presents a detailed description of the CANDU 9 Plant Display System and features that provide increased operational capabilities. (author)

  6. Plant condition assessments as a requirement before major investment in life extension for a CANDU nuclear power plant

    International Nuclear Information System (INIS)

    Aubray, Marc

    2002-01-01

    Full text: Since, to extend the life of a CANDU-6 reactor beyond its original design life requires the replacement of reactor components (380 pressure and calandria tubes), a major investment will have to be done. After a preliminary technical and economical feasibility study, Hydro- Quebec, owner of the Gentilly-2 NPP, has decided to perform a more detailed assessment to: 1. Get assurance that it is technically and economically viable to extend Gentilly-2 for another 20 years beyond the original design life; 2. Identify the detailed work to be done during the refurbishment period planned in 2008-2009; 3. Define the overall cost and the general schedule of the refurbishment phase; 4. Ensure an adequate licensing strategy to restart after refurbishment; 5. Complete all the Environmental Impact Studies required to obtain the government authorizations. The business case to support the refurbishment of Gentilly-2 has to take in consideration the reactor core components, which will be the major work to be completed during refurbishment. In summary the following main component will have to be changed or refreshed: The pressure and calandria tubes and the feeders (partial replacement only) (ageing mechanisms); The control computers (obsolescence); The condenser tubes (tubes plugging); The turbine control and electric-governor (obsolescence). An extensive campaign is under way to assess the 'health' of the station systems, structures and components (SSC). Two processes have been used for this assessment: Plant Life Management Studies (PLIM) for approximately 10 critical SSC or families of SSC (PLIM Studies); Condition Assessment Studies for other SSC with a lower impact on the Plant production or safety). The PLIM Studies are done on SSC's, which were judged critical because they are not replaceable (Reactor Building, Calandria), or that their failure could have a significant impact on safety or production (electrical motors, majors pumps, heat exchangers and pressure

  7. Requirements for class 1C, 2C, and 3C pressure-retaining components and supports in CANDU nuclear power plants

    International Nuclear Information System (INIS)

    1989-01-01

    This Standard applies to pressure-retaining components of CANDU nuclear power plants that have a code classification of Class 1C, 2C or 3C. These are pressure-retaining components where, because of the design concept, the rules of the ASME Boiler and Pressure Vessel Code do not exist, are not applicable, or are not sufficient. The Standard provides rules for the design, fabrication, installation, examination and inspection of these components and supports. It provides rules intended to ensure the pressure-retaining integrity of components, not the operability. It also provides rules for the support of fueling machines. The Standard applies only to new construction prior to the plant being declared in service

  8. Cobalt-60 production in CANDU power reactors

    International Nuclear Information System (INIS)

    Malkoske, G.R.; Norton, J.L.; Slack, J.

    2002-01-01

    MDS Nordion has been supplying cobalt-60 sources to industry for industrial and medical purposes since 1946. These cobalt-60 sources are used in many market and product segments, but are primarily used to sterilize single-use medical products including; surgical kits, gloves, gowns, drapes, and cotton swabs. Other applications include sanitization of cosmetics, microbial reduction of pharmaceutical raw materials, and food irradiation. The technology for producing the cobalt-60 isotope was developed by MDS Nordion and Atomic Energy of Canada Limited (AECL) almost 55 years ago using research reactors at the AECL Chalk River Laboratories in Ontario, Canada. The first cobalt-60 source produced for medical applications was manufactured by MDS Nordion and used in cancer therapy. The benefits of cobalt-60 as applied to medical product manufacturing, were quickly realized and the demand for this radioisotope quickly grew. The same technology for producing cobalt-60 in research reactors was then designed and packaged such that it could be conveniently transferred to a utility/power reactor. In the early 1970's, in co-operation with Ontario Power Generation (formerly Ontario Hydro), bulk cobalt-60 production for industrial irradiation applications was initiated in the four Pickering A CANDU reactors. As the demand and acceptance of sterilization of medical products grew, MDS Nordion expanded its bulk supply by installing the proprietary Canadian technology for producing cobalt-60 in additional CANDU reactors. CANDU is unique among the power reactors of the world, being heavy water moderated and fuelled with natural uranium. They are also designed and supplied with stainless steel adjusters, the primary function of which is to shape the neutron flux to optimize reactor power and fuel bum-up, and to provide excess reactivity needed to overcome xenon-135 poisoning following a reduction of power. The reactor is designed to develop full power output with all of the adjuster

  9. MAAP4 CANDU analysis of a generic CANDU-6 plant: preliminary results

    Energy Technology Data Exchange (ETDEWEB)

    Petoukhov, S.M.; Mathew, P.M

    2001-10-01

    To support the generic probabilistic safety analysis (PSA) program at AECL, in particular to conduct Level 2 PSA analysis of a CANDU 6 plant undergoing a postulated severe accident, the capability to conduct severe accident consequence analysis for a CANDU plant is required. For this purpose, AECL selected MAAP4 CANDU from a number of other severe accident codes. The necessary models for a generic CANDU 6 station have been implemented in the code, and the code version 0.2 beta was tested using station data, which were assembled for a generic CANDU 6 station. This paper describes the preliminary results of the consequence analysis using MAAP4 CANDU for a generic CANDU 6 station, when it undergoes a station blackout and a large loss-of-coolant accident scenario. The analysis results show that the plant response is consistent with the physical phenomena modeled and the failure criteria used. The results also confirm that the CANDU design is robust with respect to severe accidents, which is reflected in the calculated long times that are available for administering accident management measures to arrest the accident progression before the calandria vessel or containment become at risk. (author)

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  11. Cobalt-60 production in CANDU power reactors

    International Nuclear Information System (INIS)

    Slack, J.; Norton, J.L.; Malkoske, G.R.

    2003-01-01

    therapy machines. Today the majority of the cancer therapy cobalt-60 sources used in the world are manufactured using material from the NRU reactor in Chalk River. The same technology that was used for producing cobalt-60 in a research reactor was then adapted and transferred for use in a CANDU power reactor. In the early 1970s, in co-operation with Ontario Power Generation (formerly Ontario Hydro), bulk cobalt-60 production was initiated in the four Pickering A CANDU reactors located east of Toronto. This was the first full scale production of millions of curies of cobalt-60 per year. As the demand and acceptance of sterilization of medical products grew, MDS Nordion expanded its bulk supply by installing the proprietary Canadian technology in additional CANDUs. Over the years MDS Nordion has partnered with CANDU reactor owners to produce cobalt-60 at various sites. CANDU reactors that have, or are still producing cobalt-60, include Pickering A, Pickering B, Gentilly 2, Embalse in Argentina, and Bruce B. In conclusion, the technology for cobalt-60 production in CANDU reactors, designed and developed by MDS Nordion and Atomic Energy of Canada, has been safely, economically and successfully employed in CANDU reactors with over 195 reactor years of production. Today over forty percent of the world's disposable medical supplies are made safer through sterilization using cobalt-60 sources from MDS Nordion. Over the past 40 years, MDS Nordion with its CANDU reactor owner partners, has safely and reliably shipped more than 500 million curies of cobalt-60 sources to customers around the world. MDS Nordion is presently adding three more CANDU power reactors to its supply chain. These three additional cobalt producing CANDU's will help supplement the ability of the health care industry to provide safe, sterile, medical disposable products to people around the world. As new applications for cobalt-60 are identified, and the demand for bulk cobalt-60 increases, MDS Nordion and AECL

  12. Assessment and management of ageing of major nuclear power plant components important to safety: CANDU pressure tubes

    International Nuclear Information System (INIS)

    1998-08-01

    The report documents the current practices for assessment and management of the ageing of the pressure tubes in CANDU reactors and Indian PHWTRs. Chapter headings are: fuel channel and pressure tube description, design basis for the fuel channel and pressure tube, degradation mechanisms and ageing concerns for pressure tubes, inspection and monitoring methods for pressure tubes,assessment methods and fitness-for-service guidelines for pressure tubes, mitigation methods for pressure tubes, and pressure tube ageing management programme

  13. The licensing process of the design modifications of Cernavoda 2 NPP resulting from the operating experience of CANDU plants

    International Nuclear Information System (INIS)

    Goicea, L.

    2005-01-01

    The CANDU 6 plant now under construction in Cernavoda include over two hundred significant improvements made in order to comply with current codes and standards and licensing requirements relative to the operating CANDU 6 in Romania. These evolutionary improvements are incorporated in CANDU 6 design taking advance of CANDU operating experience, of the designer company research and development and technical advances worldwide in order to further enhance safety, reliability and economics. This paper gives a general idea of the evaluation of the modifications of the Cernavoda 2 nuclear power plant against the design of Cernavoda 1 and states the safety principles and requirements which are the basis for this evaluation. (author)

  14. Qinshan Phase III (CANDU) nuclear power project quality assurance

    International Nuclear Information System (INIS)

    Wang Lingen; Du Jinxiang

    2001-01-01

    The completion and implementation of quality assurance system of Qinshan Phase III (CANDU) nuclear power project are presented. Some comments and understanding with consideration of the project characteristics are put forward

  15. Advanced CANDU Design With Negative Power Feedback

    International Nuclear Information System (INIS)

    Andang-Widi-Harto; Muslim

    2004-01-01

    The problem of positive power feedback in the recent PHWR-CANDU design, especially related to coolant void increase, will be overcame by the use of dual moderator concept, in which two moderator systems are used, i.e. a main moderator outside the calandria tube and an annular moderator inside the annular space. Annular moderator is allowed to boil in the case of overheating. The numerical calculations have been performed for two core design namely HWR-DM-ST and HWR-DM-XI which can reach burn up of 16,000 and 17,500 MWd/ ton U respectively. The results for the two designs is that the values of k at fully annular moderator filling condition are 1.0054 (HWR-DM-ST) and 1.0019 (HWR-DM-XI), while at completely empty annular moderator condition are 0.9634 (HWR-DM-ST) and 0.9143 (HWR-DM-XI). The decrease of coolant flow rate from 3,043 kg/s to 853 kg/s decrease k values of 0.0109 (HWR-DM-ST) and 0.0232 (HWR-DM-XI). While increasing inlet coolant enthalpy from 2,950 kJ/kg to 3,175 kJ/kg decreases of k values of 0.0074 (HWR-DM-ST) and 0.0239 (HWR-DM-XI). Thus, it can be summarized that the HWR-DM design has negative power reactivity feedback.(author)

  16. Economics of CANDU

    International Nuclear Information System (INIS)

    McConnell, L.G.; Woodhead, L.W.

    1981-02-01

    The cost of producing electricity from CANDU reactors is discussed. The total unit energy cost of base-load electricity from CANDU reactors is compared with that of coal-fired plants in Ontario. In 1980 nuclear power was 8.41 m$/kW.h less costly for plants of similar size and vintage. Comparison of CANDU with pressurized water reactors indicated that the latter would be about 26 percent more costly in Ontario

  17. Machine learning techniques for the verification of refueling activities in CANDU-type nuclear power plants (NPPs) with direct applications in nuclear safeguards

    International Nuclear Information System (INIS)

    Budzinski, J.

    2006-06-01

    This dissertation deals with the problem of automated classification of the signals obtained from certain radiation monitoring systems, specifically from the Core Discharge Monitor (CDM) systems, that are successfully operated by the International Atomic Energy Agency (IAEA) at various CANDU-type nuclear power plants around the world. In order to significantly reduce the costly and error-prone manual evaluation of the large amounts of the collected CDM signals, a reliable and efficient algorithm for the automated data evaluation is necessary, which might ensure real-time performance with maximum of 0.01 % misclassification ratio. This thesis describes the research behind finding a successful prototype implementation of such automated analysis software. The finally adopted methodology assumes a nonstationary data-generating process that has a finite number of states or basic fueling activities, each of which can emit observable data patterns having particular stationary characteristics. To find out the underlying state sequences, a unified probabilistic approach known as the hidden Markov model (HMM) is used. Each possible fueling sequence is modeled by a distinct HMM having a left-right profile topology with explicit insert and delete states. Given an unknown fueling sequence, a dynamic programming algorithm akin to the Viterbi search is used to find the maximum likelihood state path through each model and eventually the overall best-scoring path is picked up as the recognition hypothesis. Machine learning techniques are applied to estimate the observation densities of the states, because the densities are not simply parameterizable. Unlike most present applications of continuous monitoring systems that rely on heuristic approaches to the recognition of possibly risky events, this research focuses on finding techniques that make optimal use of prior knowledge and computer simulation in the recognition task. Thus, a suitably modified, approximate n-best variant of

  18. Canadian CANDU plant data systems for technical surveillance and analysis

    International Nuclear Information System (INIS)

    Deverno, M.; Pothier, H.; Xian, C.; Grosbois, J. De; Bosnich, M.

    1996-01-01

    Plant data systems are emerging as a critical plant support system technology. In particular, plant-wide Historical Data Systems (HDS) are pivotal to the successful implementation of technical surveillance and analysis programs supporting plant operations, maintenance, safety, and licensing activities. In partnership with Canadian CANDU utility and design organizations, AECL has conducted a review of current Canadian CANDU HDS approaches with emphasis on understanding the existing functionality and uses of plant historical data systems, their future needs and benefits. The results is a vision of a plant-wide HDS providing seamless access to both near real-time and historical data, user tool-kits for data visualization and analysis, and data management of the large volume of data acquired during the life of a plant. The successful implementation of the HDS vision will lead to higher capability and capacity factors while minimizing Operations, Maintenance, and Administration (OM and A) costs. (author). 5 refs, 3 figs

  19. Canadian CANDU plant data systems for technical surveillance and analysis

    Energy Technology Data Exchange (ETDEWEB)

    Deverno, M; Pothier, H; Xian, C [Atomic Energy of Canada Ltd., Control Centre Technology Branch, Fredericton, NB (Canada); Grosbois, J De; Bosnich, M [Atomic Energy of Canada Ltd., Control Centre Technology Branch, Chalk River, ON (Canada). Chalk River Labs.

    1997-12-31

    Plant data systems are emerging as a critical plant support system technology. In particular, plant-wide Historical Data Systems (HDS) are pivotal to the successful implementation of technical surveillance and analysis programs supporting plant operations, maintenance, safety, and licensing activities. In partnership with Canadian CANDU utility and design organizations, AECL has conducted a review of current Canadian CANDU HDS approaches with emphasis on understanding the existing functionality and uses of plant historical data systems, their future needs and benefits. The results is a vision of a plant-wide HDS providing seamless access to both near real-time and historical data, user tool-kits for data visualization and analysis, and data management of the large volume of data acquired during the life of a plant. The successful implementation of the HDS vision will lead to higher capability and capacity factors while minimizing Operations, Maintenance, and Administration (OM and A) costs. (author). 5 refs, 3 figs.

  20. CANDU steam generator life management: laboratory data and plant experience

    International Nuclear Information System (INIS)

    Tapping, R.L.; Nickerson, J.H.; Subash, N.; Wright, M.D.

    2001-10-01

    As CANDU reactors enter middle age, and the potential value of the plants in a deregulated market is realized, life management and life extension issues become increasingly important. An accurate assessment of critical components, such as the CANDU 6 steam generators (SGs), is crucial for successful life extension, and in this context, material issues are a key factor. For example, service experience with Alloy 900 tubing indicates very low levels of degradation within CANDU SGs; the same is also noted worldwide. With little field data for extrapolation, life management and life extension decisions for the tube bundles rely heavily on laboratory data. Similarly, other components of the SGs, in particular the secondary side internals, have only limited inspection data upon which to base a condition assessment. However, in this case there are also relatively little laboratory data. Decisions on life management and life extension are further complicated--not only is inspection access often restricted, but repair or replacement options for internal components are, by definition, also limited. The application of CANDU SG life management and life extension requires a judicious blend of in-service data, laboratory research and development (R and D) and materials and engineering judgment. For instance, the available laboratory corrosion and fretting wear data for Alloy 800 SG tubing have been compared with plant experience (with all types of tubing), and with crevice chemistry simulations, in order to provide an appropriate inspection guide for a 50-year SG life. A similar approach has been taken with other SG components, where the emphasis has been on known degradation mechanisms worldwide. This paper provides an outline of the CANDU SG life management program, including the results to date, a summary of the supporting R and D program showing the integration with condition assessment and life management activities, and the approach taken to life extension for a typical

  1. Design, construction and operation of Ontario Hydro's CANDU plants

    International Nuclear Information System (INIS)

    Campbell, P.G.

    1981-06-01

    Ontario Hydro has been producing electricity commercially from nuclear power since 1968, using CANDU reactors which have proved enormously successful. The 206-MW Douglas Point station, nearly 10 times larger than the first Canadian power reactor, NPD-2, resulted from a cooperative effort between Atomic Energy of Canada Ltd., the provincial government of Ontario, and Ontario Hydro. This approach led to a basic working relationship between the parties, with Ontario Hydro acting as project manager and builder, and AECL acting as consultant with respect to the nuclear components. Before Douglas Point was fully commissioned Ontario Hydro was ready to commit itself to more nuclear stations, and work was started on the four-unit Pickering nuclear generating station. Multi-unit stations were adopted to achieve economies of scale, and the concept has been retained for all subsequent nuclear power plants constructed in the province. The organization of Ontario Hydro's project management, construction, and operation of nuclear generating stations is described. Performance of the existing stations and cost of the power they produce have been entirely acceptable

  2. CANDU fuel - fifteen years of power reactor experience

    International Nuclear Information System (INIS)

    Fanjoy, G.R.; Bain, A.S.

    1977-01-01

    CANDU (Canada Deuterium Uranium) fuel has operated in power reactors since 1962. Analyses of performance statistics, supplemented by examinations of fuel from power reactors and experimental loops have yielded: (a) A thorough understanding of the fundamental behaviour of CANDU fuel. (b) Data showing that the predicted high utilization of uranium has been achieved. Actual fuelling costs in 1976 at the Pickering Generating Station are 1.2 m$/kWh (1976 Canadian dollars) with the simple oncethrough natural-UO 2 fuel cycle. (c) Criteria for operation, which have led to the current very low defect rate of 0.03% of all assemblies and to ''CANLUB'' fuel, which has a graphite interlayer between the fuel and sheath to reduce defects on power increases. (d) Proof that the short length (500 mm), collapsible cladding features of the CANDU bundle are successful and that the fuel can operate at high-power output (current peak outer-element linear power is 58 +- 15% kW/m). Involvement by the utility in all stages of fuel development has resulted in efficient application of this fundamental knowledge to ensure proper fuel specifications, procurement, scheduling into the reactor and feedback to developers, designers and manufacturers. As of mid-1976 over 3 x 10 6 individual elements have been built in a well-estabilished commercially competitive fuel fabrication industry and over 2 x 10 6 elements have been irradiated. Only six defects have been attributed to faulty materials or fabrication, and the use of high-density UO 2 with low-moisture content precluded defects from hydrogen contamination and densification. Development work on UO 2 and other fuel cycles (plutonium and thorium) is continuing, and, because CANDU reactors use on-power fuelling, bundles can be inserted into power reactors for testing. Thus new fuel designs can be quickly adopted to ensure that the CANDU system continues to provide low-cost energy with high reliability

  3. Impact of aging and material structure on CANDU plant performance

    International Nuclear Information System (INIS)

    Nadeau, E.; Ballyk, J.; Ghalavand, N.

    2011-01-01

    In-service behaviour of pressure tubes is a key factor in the assessment of safety margins during plant operation. Pressure tube deformation (diametral expansion) affects fuel bundle dry out characteristics resulting in reduced margin to trip for some events. Pressure tube aging mechanisms also erode design margins on fuel channels or interfacing reactor components. The degradation mechanisms of interest are primarily deformation, loss of fracture resistance and hydrogen ingress. CANDU (CANada Deuterium Uranium, a registered trademark of the Atomic Energy of Canada Limited used under exclusive licence by Candu Energy Inc.) owners and operators need to maximize plant capacity factor and meet or exceed the reactor design life targets while maintaining safety margins. The degradation of pressure tube material and geometry are characterized through a program of inspection, material surveillance and assessment and need to be managed to optimize plant performance. Candu is improving pressure tubes installed in new build and life extension projects. Improvements include changes designed to reduce or mitigate the impact of pressure tube elongation and diametral expansion rates, improvement of pressure tube fracture properties, and reduction of the implications of hydrogen ingress. In addition, Candu provides an extensive array of engineering services designed to assess the condition of pressure tubes and address the impact of pressure tube degradation on safety margins and plant performance. These services include periodic and in-service inspection and material surveillance of pressure tubes and deterministic and probabilistic assessment of pressure tube fitness for service to applicable standards. Activities designed to mitigate the impact of pressure tube deformation on safety margins include steam generator cleaning, which improves trip margins, and trip design assessment to optimize reactor trip set points restoring safety and operating margins. This paper provides an

  4. Impact of aging and material structure on CANDU plant performance

    Energy Technology Data Exchange (ETDEWEB)

    Nadeau, E.; Ballyk, J.; Ghalavand, N. [Candu Energy Inc., Mississauga, Ontario (Canada)

    2011-07-01

    In-service behaviour of pressure tubes is a key factor in the assessment of safety margins during plant operation. Pressure tube deformation (diametral expansion) affects fuel bundle dry out characteristics resulting in reduced margin to trip for some events. Pressure tube aging mechanisms also erode design margins on fuel channels or interfacing reactor components. The degradation mechanisms of interest are primarily deformation, loss of fracture resistance and hydrogen ingress. CANDU (CANada Deuterium Uranium, a registered trademark of the Atomic Energy of Canada Limited used under exclusive licence by Candu Energy Inc.) owners and operators need to maximize plant capacity factor and meet or exceed the reactor design life targets while maintaining safety margins. The degradation of pressure tube material and geometry are characterized through a program of inspection, material surveillance and assessment and need to be managed to optimize plant performance. Candu is improving pressure tubes installed in new build and life extension projects. Improvements include changes designed to reduce or mitigate the impact of pressure tube elongation and diametral expansion rates, improvement of pressure tube fracture properties, and reduction of the implications of hydrogen ingress. In addition, Candu provides an extensive array of engineering services designed to assess the condition of pressure tubes and address the impact of pressure tube degradation on safety margins and plant performance. These services include periodic and in-service inspection and material surveillance of pressure tubes and deterministic and probabilistic assessment of pressure tube fitness for service to applicable standards. Activities designed to mitigate the impact of pressure tube deformation on safety margins include steam generator cleaning, which improves trip margins, and trip design assessment to optimize reactor trip set points restoring safety and operating margins. This paper provides an

  5. A short history of the CANDU nuclear power system

    Energy Technology Data Exchange (ETDEWEB)

    Brooks, G L

    1993-04-01

    This paper provides a short historical summary of the evolution of the CANDU nuclear power system with emphasis on the roles played by Ontario Hydro and private sector companies in Ontario in collaboration with Atomic Energy of Canada Limited (AECL). (author). 1 fig., 61 refs.

  6. A short history of the CANDU nuclear power system

    International Nuclear Information System (INIS)

    Brooks, G.L.

    1993-04-01

    This paper provides a short historical summary of the evolution of the CANDU nuclear power system with emphasis on the roles played by Ontario Hydro and private sector companies in Ontario in collaboration with Atomic Energy of Canada Limited (AECL). (author). 1 fig., 61 refs

  7. Improved containment isolation for CANDU plants

    Energy Technology Data Exchange (ETDEWEB)

    Stretch, A H [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

    1996-12-31

    The publication of Regulatory Policy Statement R- 7 in February 1991 by the Atomic Energy Control Board imposes new requirement for the isolation of fluid piping penetrating the containment boundary. The Appendix of R-7 describes the detailed requirements for metal extensions of the containment envelope, including the code classification qf the pressure retaining portions and isolation requirements for fluid piping and tubing.The application of these new requirements to the existing CANDU 6 design resulted in a number of design changes, including the addition of manual and automatic isolation valves, changes in code classification, and leakage monitoring considerations. (author). 3 refs., 3 figs.

  8. Operating experiences with Neutron Overpower Trip Systems in Ontario Hydro's CANDU nuclear plants

    International Nuclear Information System (INIS)

    Hnik, J.; Kozak, J.

    1991-01-01

    Operating experiences with Neutron Over Power Trip (NOP) Systems in different Ontario Hydro CANDU nuclear power plants are discussed. Lessons learned from the system operation and their impact on design improvements are presented. Retrofitting of additional tools, such as Shutdown System Monitoring computers, to improve operator interaction with the system is described. Experiences with the reliability of some of the NOP system components is also discussed. Options for future enhancements of system performance and operability are identified. (author)

  9. Improving chemistry performance in CANDU plants

    International Nuclear Information System (INIS)

    Turner, C.; Guzonas, D.

    2010-01-01

    There is a strong interplay between coolant chemistry and materials selection in any nuclear power plant system. To achieve the design life of reactor components it is necessary to monitor and control relevant chemistry parameters, such as ionic conductivity, pH, concentrations of dissolved ions and redox species (e.g., hydrogen, hydrazine, oxygen) and the concentrations of suspended corrosion products. Chemistry specifications are set to achieve a balance between the sometimes conflicting requirements to minimize corrosion and radiological dose and to minimize operating and maintenance costs over the lifetime of the plant. For the past decade, Atomic Energy of Canada Limited (AECL) has taken a rigorous and disciplined approach to reviewing and updating all aspects of chemistry control in the CANDU® nuclear power plant (NPP). This approach has included proactively reviewing chemistry operating experience from existing CANDU® and other water-cooled NPPs worldwide to identify and address emerging issues, updating all of our chemistry control documentation to ensure that each chemistry parameter is linked to a specific requirement (e.g., reduce activity transport, monitor for condenser leak) and incorporating the latest results from our Research and Development (R and D) programs to ensure that all chemistry specifications are supported by a sound rationale. The results of this review and update have been incorporated into updated chemistry specifications and, in some cases, modified operating procedures for new and existing plants. In addition, recommendations have been made for design modifications to improve chemistry control in new build plants, especially during periods of shutdown and startup when chemistry control has traditionally been more challenging. Chemistry control in new-build CANDU® plants will rely increasingly on the use of on-line instrumentation interfaced directly to AECL's state-of-the-art chemistry monitoring, diagnostics and analysis

  10. Expert systems use in present and future CANDU nuclear power supply systems

    International Nuclear Information System (INIS)

    Lupton, L.R.; Basso, R.A.J.; Anderson, L.L.; Anderson, J.W.D.

    1989-11-01

    As CANDU nuclear power plants become more complex, and are operated under tighter constraints for longer periods between outages, plant operations staff will have to absorb more information to correctly and rapidly respond to upsets. A development program is underway at Atomic Energy of Canada Limited to use expert systems and interactive media tools to assist operations staff of existing and future CANDU plants. The complete system for plant information access and display, on-line advice and diagnosis, and interactive operating procedures is called the Operator Companion. A prototype, consisting of operator consoles, expert systems and simulation modules in a distributed architecture, is currently being developed to demonstrate the concepts of the Operator Companion. Specialized advisors are also being developed using expert system technology to meet specific operational and design needs

  11. Safety of CANDU nuclear power stations

    International Nuclear Information System (INIS)

    Snell, V.G.

    1978-11-01

    A nuclear plant contains a large amount of radioactive material which could be a potential threat to public health. The plant is therefore designed, built and operated so that the risk to the public is low. Careful design of the normal reactor systems is the first line of defense. These systems are highly resistant to an accident happening in the first place, and can also be effective in stopping it if it does happen. Independent and redundant safety sytems minimize the effects of an accident, or stop it completely. They include shutdown systems, emergency core cooling systems, and containment systems. Massive impairment of any one safety system together with an accident can be tolerated. This 'defence in depth' approach recognizes that men and machines are imperfect and that the unexpected happens. The nuclear power plant need not be perfect to be safe. To allow meaningful judgements we must know how safe the plant is. The Atomic Energy Control Board guidelines give one such measure, but they may overestimate the true risk. We interpret these guidelines as an upper limit to the total risk, and trace their evolution. (author)

  12. Current activities in support of CANDU plant life management: an industry perspective

    International Nuclear Information System (INIS)

    Shalaby, B.A.; Price, E.G.; Hopkins, J.; Charlebois, P.

    1998-01-01

    The current focus of the CANDU industry is to position the nuclear option as a cost competitive, safe and reliable means of electricity production. To achieve its goal the CANDU industry as a whole is undertaking steps to improve further its performance and safety of its nuclear power plants. A number of programs have been planned and implemented particularly for plants in the mid-life range. Some of these programs include life assessment studies of critical systems, Structures and Components (SSCS), refurbishment and upgrading programs and monitoring and periodic inspection programs. Some elements of the programs have been in place from station start up and some are being instituted as part of the aging management and performance improvement program. The industry recognizes that the key to sustaining high performance over the life of the plant is the implementation of an integrated aging management program that encompasses all elements of plant operation and maintenance. A systematically implemented program on optimized maintenance and inspection strategy, standardized work processes, component rehabilitation programs, and applying lessons learned are some of the elements of a sustainable high performance and an effective plant life assurance program. The paper will describe the elements of an integrated program, the multiphase approach defined for CANDU PLIM and some of the activities undertaken by the industry to further improve and sustain plant safety, reliability and performance. (author)

  13. Station black out analysis for CANDU 6 plant

    International Nuclear Information System (INIS)

    Baburajan, P.K.; Rao, R.S.; Gupta, S.K.

    2011-01-01

    As part of International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP), 'Benchmarking severe accident computer codes for pressurised heavy water reactor applications', thermal hydraulic analysis of severe accident station black out (SBO) is carried out for a generic CANDU 6 plant. The CRP is conducted in order to improve severe accident analysis capability for heavy water reactors (HWRs) through the benchmarking exercise. The plant simulation is carried out using RELAP5/Mod3.4 best estimate system thermal hydraulic code. The total thermal power of the plant is 2064 MW. There are 380 fuel channels in the core, 12 fuel bundles per channel and each bundle assembly has 37 fuel elements. The primary heat transport system (HTS) consists of two loops. Each loop consist of inlet and exit headers, feeder lines, fuel channels, hot leg and cold leg of steam generator, pumps, pump suction and discharge lines. Ninety five fuel channels in each pass of the loop are simulated as a single channel. The steam generator as the secondary side heat sink consists of annulus down-comer, riser, steam separator, steam drum, steam header and steam lines. Fuel channels (pressure tube) and calandria tube are simulated using SCDAPSIM to study the severe accident code behaviour. The SBO transient is initiated after obtaining the steady state conditions. Present analysis is carried out till the pressure tube failure. Analysis results show that the secondary inventory is lost in about 6500 seconds of the transient. The primary inventory is lost in 10370 seconds of the transient and subsequently the pressure tube failure is predicted as the tube wall temperature exceeded 900 K. Further analysis is to be carried out by incorporating changes in the calandria model and including the modeling of calandria vault and containment. (author)

  14. Prediction of power-ramp defects in CANDU fuel

    International Nuclear Information System (INIS)

    Gillespie, P.; Wadsworth, S.; Daniels, T.

    2010-01-01

    Power ramps result in fuel pellet expansion and can lead to fuel sheath failures by fission product induced stress corrosion cracking (SCC). Historically, empirical models fit to experimental test data were used to predict the onset of power-ramp failures in CANDU fuel. In 1988, a power-ramped fuel defect event at PNGS-1 led to the refinement of these empirical models. This defect event has recently been re-analyzed and the empirical model updated. The empirical model is supported by a physically based model which can be used to extrapolate to fuel conditions (density, burnup) outside of the 1988 data set. (author)

  15. The CANDU 3 containment structure

    International Nuclear Information System (INIS)

    1994-01-01

    The design of the CANDU 3 nuclear power plant is being developed by AECL CANDU's Saskatchewan office. There are 24 CANDU nuclear power units operating in Canada and abroad and eight units are under construction is Romania and South Korea. The design of the CANDU 3 plant has evolved on the basis of the proven CANDU design. The experiences gained during construction, commissioning and operation of the existing CANDU plants are considered in the design. Many technological enhancements have been implemented in the design processes in all areas. The object has been to develop an improved reactor design that is suitable for the current and the future markets worldwide. Throughout the design phase of CANDU 3, emphasis has been placed in reducing the cost and construction schedule of the plant. This has been achieved by implementing design improvements and using new construction techniques. Appropriate changes and improvements to the design to suit new requirements are also adopted. In CANDU plants, the containment structure acts as an ultimate barrier against the leakage of radioactive substances during normal operations and postulated accident conditions. The concept of the structural design of the containment structure has been examined in considerable detail. This has resulted in development of a new conceptual design for the containment structure for CANDU 3. This paper deals with this new design of the containment structure

  16. CANDU nuclear reactor technology

    International Nuclear Information System (INIS)

    Kakaria, B. K.

    1994-01-01

    AECL has over 40 years of experience in the nuclear field. Over the past 20 years, this unique Canadian nuclear technology has made a worldwide presence, In addition to 22 CANDU reactors in Canada, there are also two in India, one in Pakistan, one in Argentina, four in Korea and five in Romania. CANDU advancements are based on evolutionary plant improvements. They consist of system performance improvements, design technology improvements and research and development in support of advanced nuclear power. Given the good performance of CANOU plants, it is important that this CANDU operating experience be incorporated into new and repeat designs

  17. Explosive and corrosive concentration analysis of gases produced in a CANDU type (N2, D2, O2, H2) nuclear power plant

    International Nuclear Information System (INIS)

    Binetti, E.O.

    1987-01-01

    The primary loop gas of an automatic control system of a nuclear power plant is of great importance as regards conservation and safety of the plant. These gases are produced by dissociation due to radiation effects on heavy water. The system is based on a sample capture equipment, a chromatographic analyzer with its associated electronics, a sample separator and conditioner, a temperature and pressure control system of the transport gas, all included in the reactor building, apart from other supporting instrumentation. (Author)

  18. CANDU fuel - fifteen years of power reactor experience

    International Nuclear Information System (INIS)

    Fanjoy, G.R.; Bain, A.S.

    1977-05-01

    Analyses of performance statistics, supplemented by examinations of fuel from power reactors and experimental loops have yielded: (a) a thorough understanding of the fundamental behaviour of CANDU fuel; (b) data showing that the predicted high utilization of uranium has been achieved; (c) criteria for operation, which have led to the current very low defect rate of 0.03% of all assemblies and to 'CANLUB' fuel, which has a graphite interlayer between the fuel and sheath to reduce defects on power increases; (d) proof that the short length (500 mm), collapsible cladding features of the CANDU bundle are successful and that the fuel can operate at high-power output (current peak outer-element linear power is 58 +- 15% kW/m). As of mid-1976 over 3 x 10 6 individual elements have been built and over 2 x 10 6 elements have been irradiated. Only six defects have been attributed to faulty materials or fabrication, and the use of high-density UO 2 with low-moisture content precluded defects from hydrogen contamination and densification

  19. Enhancing plant performance in newer CANDU plants utilizing PLiM methodologies

    International Nuclear Information System (INIS)

    Azeez, S.; Krishnan, V.S.; Nickerson, J.H.; Kakaria, B.

    2002-01-01

    Over the past 5 years, Atomic Energy of Canada Ltd. (AECL) has been working with CANDU utilities on comprehensive and integrated CANDU PLiM programs for successful and reliable operation through design life and beyond. Considerable progress has been made in the development of CANDU PLiM methodologies and implementation of the outcomes at the plants. The basis of CANDU PLiM programs is to understand the ageing degradation mechanisms, prevent/minimize the effects of these phenomena in the Critical Structures, Systems and Components (CSSCs), and maintain the CSSC condition as close as possible in the best operating condition. Effective plant practices in surveillance, maintenance, and operations are the primary means of managing ageing. From the experience to date, the CANDU PLiM program will modify and enhance, but not likely replace, existing plant programs that address ageing. However, a successful PLiM program will provide assurance that these existing plant programs are both effective and can be shown to be effective, in managing ageing. This requires a structured and managed approach to both the assessment and implementation processes

  20. Dimensional response of CANDU fuel to power changes

    Energy Technology Data Exchange (ETDEWEB)

    Fehrenbach, P J [Fuel Engineering Branch, Chalk River Nuclear Laboratories, Atomic Energy of Canada Limited, Chalk River, ON (Canada); Hastings, I J; Morel, P A; Sage, R D; Smith, A D [Fuel Materials Branch, Chalk River Nuclear Laboratories, Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    1983-06-01

    The introduction of CANLUB-coated fuel cladding, modified fuel management schemes, and revisions to the sequence of control rod movements, have eliminated power ramping fuel failures in CANDU power reactors. However, an irradiation program continues at Chalk River Nuclear Laboratories to determine the effect of various design and operating parameters on the dimensional response of UO{sub 2} fuel elements to power changes, over a range of conditions outside those normally experienced by CANDU power reactor fuel. We have investigated the effect of power changes on element diameter for UO{sub 2} fuel with starting densities of 10.6 and 10.8 Mg/m{sup 3} clad in 0.4 mm thick Zircaloy, at burnups from 0 to 100 MW.h/kg U. Element diameter measurements were obtained at power using an In-Reactor Diameter Measuring Rig (IRDMR). Rates of power change over the range 0.0005 to 0.03 kW.m{sup -1}.s{sup -1} were achieved by a combination of reactor power control and use of a Helium-3 power cycling facility. Total diameter increases in unirradiated elements were about 1% at pellet interface locations for both fuel densities during the initial power increase to 60 kW/m. Diameter changes during subsequent power cycles of these elements from 55 to 100% maximum power were significantly larger for the higher density fuel, ranging from 0.3 to 0.5% compared to less than 0.1% for the standard density (10.6 Mg/m{sup 3}) fuel. In elements pre-irradiated at 27 kW/m to burnups of about 100 MW.h/kg U prior to power ramping, the diameter increases measured after ramping to 55 kW/m also varied with starting fuel density. Diameter changes at pellet interface locations were about 0.9% and 0.6% for higher density and standard density fuel respectively. (author)

  1. CANDU 3000

    International Nuclear Information System (INIS)

    Keillor, Mac

    1987-01-01

    In this article, the CANDU 300 design, and the team that designed it, are featured. The CANDU 300 will operate at an energy cost similar to that of the larger CANDU units, but is sized for emerging markets. Ease of construction is an important feature: for example, full 360-degree access is available to each of the five buildings during construction; and the whole plant consists of about 90 modules, which can be built in separate locations, and hoisted into place

  2. Improving CANDU plant operation and maintenance through retrofit information technology systems

    International Nuclear Information System (INIS)

    Lupton, L.R.; Judd, R.A.; MacBeth, M.J.

    1998-01-01

    CANDU plant owners are facing an increasingly competitive environment for the generation of electricity. To meet this challenge, all owners have identified that information technology offers opportunities for significant improvements in CANDU operation, maintenance and administration (OM and A) costs. Targeted information technology application areas include instrumentation and control, engineering, construction, operations and plant information management. These opportunities also pose challenges and issues that must be addressed if the full benefits of the advances in information technology are to be achieved. Key among these are system hardware and software maintenance, and obsolescence protection; AECL has been supporting CANDU stations with the initial development and evaluation of systems to improve plant performance and cost. Key initiatives that have been implemented or are in the process of being implemented in some CANDU plants to achieve operational benefits include: critical safety parameter monitor system; advanced computerized annunciation system; plant historical data system; and plant display system. Each system will be described in terms of its role in enhancing current CANDU plant performance and how they will contribute to future CANDU plant performance

  3. Improving CANDU plant operation and maintenance through retrofit information technology systems

    International Nuclear Information System (INIS)

    Lupton, L. R.; Judd, R. A.

    1998-01-01

    CANDU plant owners are facing an increasingly competitive environment for the generation of electricity. To meet this challenge, all owners have identified that information technology offers opportunities for significant improvements in CANDU operation, maintenance and administration (OM and A) costs. Targeted information technology application areas include instrumentation and control, engineering, construction, operations and plant information management. These opportunities also pose challenges and issues that must be addressed if the full benefits of the advances in information technology are to be achieved. Key among these are system hardware and software maintenance, and obsolescence protection. AECL has been supporting CANDU stations with the initial development and evaluation of systems to improve plant performance and cost. Five key initiatives that have been implemented or are in the process of being implemented in some CANDU plants to achieve cooperational benefits include: critical safety parameter monitor system; advanced computerized annunciation system; plant historical data system; plant display system; and digital protection system. Each system will be described in terms of its role in enhancing current CANDU plant performance and how they will contribute to future CANDU plant performance. (author). 8 refs., 3 figs

  4. R and D in support of CANDU plant life management

    International Nuclear Information System (INIS)

    Tapping, R.L.; Holt, R.A.

    1999-01-01

    One of the keys to the long-term success of CANDUs is a high capacity factor over the station design life. Considerable R and D in underway at AECL to develop technologies for assessing, monitoring and mitigating the effect of plant ageing and for improving plant performance and extending plant life. To achieve longer service life and to realize high capacity factor from CANDU stations, AECL is developing new technologies to enhance fuel channel and steam generator inspection capabilities, to monitor system health, and to allow preventive maintenance and cleaning (e.g., on-line chemical cleaning processes that produce small volumes of wastes). The life management strategy for fuel channels and steam generators requires a program to inspect components on a routine basis to identify mechanisms that could potentially affect fitness-for-service. In the case of fuel channels, the strategy includes inspections for dimensional changes, flaw detection, and deuterium concentration. New techniques are been developed to enhance these inspection capabilities; examples include accurate measurement of the gap between a pressure tube and its calandria tube and rapid full-length inspections of steam generator tubes for all known flaw types. Central to life management of components are Fitness-for-Service Guidelines (FFSG) that have been developed with the CANDU Owners Group (COG) that provide a standardized method to assess the potential for propagation of flaws detected during in-service inspections, and assessment of any change in fracture characteristics of the material. FFSG continue to be improved with the development of new technologies such as the capability to credit relaxation of stresses due to creep and non-rejectable flaws in pressure tubes. Effective management of plant systems throughout their lifetime requires much more than data acquisition and display - it requires that system health is continually monitored and managed. AECL has developed a system Health Monitor

  5. Enhanced CANDU 6 Reactor

    International Nuclear Information System (INIS)

    Azeez, S.; Alizadeh, A.; Girouard, P.

    2005-01-01

    Full text: The CANDU 6 power reactor is visionary in its approach, remarkable for its on-power refuelling capability and proven over years of safe, economical and reliable power production. Developed by Atomic Energy of Canada Ltd, the CANDU 6 design offers excellent performance utilizing state-of-the-art technology. The first CANDU 6 plants went into service in the early 1980's as leading edge technology and the design has been continuously advanced to maintain superior performance with an outstanding safety record. The first CANDU 6 plants- Gentilly 2 and Point Lepreau in Canada, Embalse in Argentina and Wolsong- Unit 1 in Korea have been in service for more than 21 years and are still producing electricity at peak performance and to the end of 2004, their average lifetime Capacity Factor was 83.2%. The newer CANDU 6 units in Romania (Cernavoda 1), Korea (Wolsong-Units 2, 3 and 4) and Qinshan (Phase III- Units 1 and 2) have also been performing at outstanding levels. The average lifetime Capacity Factor of the 10 CANDU 6 operating units around the world has been 87% to the end of 2004. Building on these successes, AECL is committed to the further development of this highly successful design, now focussing on meeting customer's needs for reduced costs, further improvements to plant operation and performance, enhanced safety and incorporating up-to-date technology as warranted. This has resulted in AECL embarking on improving the CANDU 6 design through an upgraded product termed as the 'Enhanced CANDU 6' (EC6)- which incorporates several attractive but proven features that will make the CANDU 6 reactor even more economical, safer and easier to operate. Some of the key features that will be incorporated in the EC6 include increasing the plant's power output, shortening the overall project schedule, decreasing the capital cost, dealing with obsolescence issues, optimizing maintenance outages and incorporating lessons learnt through feedback obtained from the

  6. Modularized construction, structural design and analysis of CANDU 3 plant

    Energy Technology Data Exchange (ETDEWEB)

    Biswas, J K; Wollin, S; Selvadurai, S; Saudy, A M [Atomic Energy of Canada Ltd., Saskatoon, SK (Canada)

    1996-12-31

    CANDU 3 is rated at 450 MW electric, and is a smaller and advanced version of CANDU reactors successfully operating in Canada and abroad. The design uses modularization to minimize the construction schedule and thereby reduce cost. The paper (which is published only as a long summary), deals with the concept of modularization, and with stress analysis of the various civil structures.

  7. Modularized construction, structural design and analysis of CANDU 3 plant

    International Nuclear Information System (INIS)

    Biswas, J.K.; Wollin, S.; Selvadurai, S.; Saudy, A.M.

    1995-01-01

    CANDU 3 is rated at 450 MW electric, and is a smaller and advanced version of CANDU reactors successfully operating in Canada and abroad. The design uses modularization to minimize the construction schedule and thereby reduce cost. The paper (which is published only as a long summary), deals with the concept of modularization, and with stress analysis of the various civil structures

  8. Experience of oil in CANDU moderator during A831 planned outage at Bruce Power

    International Nuclear Information System (INIS)

    Ma, G.; Nashiem, R.; Matheson, S.; Stuart, C.; Roberts, J.G.

    2011-01-01

    In their address to the Nuclear Plant Chemistry Conference 2009, Bruce Power staff will describe the effects of oil ingress to the moderator of a CANDU reactor. During the A831 planned outage of Bruce Power Unit 3, an incident of oil ingress into moderator was discovered on Oct 17, 2008. An investigation identified the cause of the oil ingress. Atomic Energy of Canada Ltd. (AECL) assessed operability of the reactor with the oil present and made recommendations with respect to the effect on unit start-up with oil present. The principal concern was the radiolytic generation of deuterium from the breakdown of the oil in-core. Various challenges were presented during start-up which were overcome via innovative approaches. The subsequent actions and consequential effects on moderator chemistry are discussed in this paper. Examination of the plant chemistry data revealed some interesting aspects of moderator system chemistry under upset conditions which will also be presented. (author)

  9. Experience of oil in CANDU moderator during A831 planned outage at Bruce Power

    Energy Technology Data Exchange (ETDEWEB)

    Ma, G.; Nashiem, R.; Matheson, S. [Bruce Power, Tiverton, Ontario (Canada); Stuart, C. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Roberts, J.G. [CANTECH Associates Ltd., Burlington, Ontario (Canada)

    2011-03-15

    In their address to the Nuclear Plant Chemistry Conference 2009, Bruce Power staff will describe the effects of oil ingress to the moderator of a CANDU reactor. During the A831 planned outage of Bruce Power Unit 3, an incident of oil ingress into moderator was discovered on Oct 17, 2008. An investigation identified the cause of the oil ingress. Atomic Energy of Canada Ltd. (AECL) assessed operability of the reactor with the oil present and made recommendations with respect to the effect on unit start-up with oil present. The principal concern was the radiolytic generation of deuterium from the breakdown of the oil in-core. Various challenges were presented during start-up which were overcome via innovative approaches. The subsequent actions and consequential effects on moderator chemistry are discussed in this paper. Examination of the plant chemistry data revealed some interesting aspects of moderator system chemistry under upset conditions which will also be presented. (author)

  10. The next generation CANDU 6

    International Nuclear Information System (INIS)

    Hopwood, J.M.

    1999-01-01

    AECL's product line of CANDU 6 and CANDU 9 nuclear power plants are adapted to respond to changing market conditions, experience feedback and technological development by a continuous improvement process of design evolution. The CANDU 6 Nuclear Power Plant design is a successful family of nuclear units, with the first four units entering service in 1983, and the most recent entering service this year. A further four CANDU 6 units are under construction. Starting in 1996, a focused forward-looking development program is under way at AECL to incorporate a series of individual improvements and integrate them into the CANDU 6, leading to the evolutionary development of the next-generation enhanced CANDU 6. The CANDU 6 improvements program includes all aspects of an NPP project, including engineering tools improvements, design for improved constructability, scheduling for faster, more streamlined commissioning, and improved operating performance. This enhanced CANDU 6 product will combine the benefits of design provenness (drawing on the more than 70 reactor-years experience of the seven operating CANDU 6 units), with the advantages of an evolutionary next-generation design. Features of the enhanced CANDU 6 design include: Advanced Human Machine Interface - built around the Advanced CANDU Control Centre; Advanced fuel design - using the newly demonstrated CANFLEX fuel bundle; Improved Efficiency based on improved utilization of waste heat; Streamlined System Design - including simplifications to improve performance and safety system reliability; Advanced Engineering Tools, -- featuring linked electronic databases from 3D CADDS, equipment specification and material management; Advanced Construction Techniques - based on open top equipment installation and the use of small skid mounted modules; Options defined for Passive Heat Sink capability and low-enrichment core optimization. (author)

  11. Occupational radiation exposures at Canadian CANDU nuclear power stations

    International Nuclear Information System (INIS)

    LeSurf, J.E.; Taylor, G.F.

    1982-09-01

    In Canada, methods to reduce the radiation exposure to workers at nuclear power reactors have been studied and implemented since the early days of the CANDU reactor program. Close collaboration between the designers, the operators, and the manufacturers has reduced the total exposure at each station, the dose requirement to operate and maintain each successive station compared with earlier stations, and the average annual exposure per worker. Specific methods developed to achieve dose reduction include water chemistry; corrosion resistant materials; low cobalt materials; decontamination; hot filtration, improved equipment reliability, maintainability, and accessibility; improved shielding design and location; planning of work for low exposure; improved operating and maintenance procedures; removal of tritium from D 2 O systems and work environments; improved protective clothing; on-power refuelling; worker awareness and training; and many other small improvements. The 1981 occupational dose productivity factors for Pickering A and Bruce A nuclear generating stations were respectively 0.43 and 0.2 rem/MW(e).a

  12. A feasible approach to implement a commercial scale CANDU fuel manufacturing plant in Egypt

    International Nuclear Information System (INIS)

    El-Shehawy, I.; El-Sharaky, M.; Yasso, K.; Selim, I.; Graham, N.; Newington, D.

    1995-01-01

    Many planning scenarios have been examined to assess and evaluate the economic estimates for implementing a commercial scale CANDU fuel manufacturing plant in Egypt. The cost estimates indicated strong influence of the annual capital costs on total fuel manufacturing cost; this is particularly evident in a small initial plant where the proposed design output is only sufficient to supply reload fuel for a single CANDU-6 reactor. A modular approach is investigated as a possible way, to reduce the capital costs for a small initial fuel plant. In this approach the plant would do fuel assembly operations only and the remainder of a plant would be constructed and equipped in the stages when high production volumes can justify the capital expenses. Such approach seems economically feasible for implementing a small scale CANDU fuel manufacturing plant in developing countries such as Egypt and further improvement could be achieved over the years of operation. (author)

  13. Improvement of Candu-1000 MW(e) power cycle by moderator heat recovery

    International Nuclear Information System (INIS)

    Fath, H.E.S.

    1988-01-01

    Four different moderator heat recovery circuits are proposed for CANDU-1000 MW(e) reactors. The proposed circuits utilize all, or part, of the 155 MW(th) moderator heat load (at 70 0 C moderator outlet temperature from calandria) to the first stage of the feed water heating system. An economics study was carried out and indicated that the direct circulation of feed water through the moderator heat exchanger (with full heat recovery) is the most economical scheme. For this scheme the saved steam from the turbine extraction was found to produce additional electric power of 8 MW(e). This additional power represents a 0.7% increase in the plants nominal electric output. The outstanding features and advantages of the selected scheme are also presented. (author)

  14. Operations quality assurance for nuclear power plants

    International Nuclear Information System (INIS)

    1987-01-01

    This standard covers the quality assurance of all activities concerned with the operation and maintenance of plant equipment and systems in CANDU-based nuclear power plants during the operations phase, the period between the completion of commissioning and the start of decommissioning

  15. Trends in CANDU licensing

    International Nuclear Information System (INIS)

    Snell, V.G.; Grant, S.D.

    1997-01-01

    Modern utilities view nuclear power more and more as a commodity - it must compete 'today' with current alternatives to attract their investment. With its long construction times and large capital investment, nuclear plants are vulnerable to delays once they have been committed. There are two related issues. Where the purchaser and the regulator are experienced in CANDU, the thrust is a very practical one: to identify and resolve major licensing risks at a very early stage in the project. Thus for a Canadian project, the designer (AECL) and the prospective purchaser would deal directly with the AECB. However CANDU has also been successfully licensed in other countries, including Korea, Romania, Argentina, India and Pakistan. Each of these countries has its own regulatory agency responsible for licensing the plant. In addition, however, the foreign customer and regulator may seek input from the AECB, up to and including a statement of licensability in Canada; this is not normally needed for a ''repeat'' plant and/or if the customer is experienced in CANDU, but can be requested if the plant configuration has been modified significantly from an already-operating CANDU. It is thus the responsibility of the designer to initiate early discussions with the AECB so the foreign CANDU meets the expectations of its customers

  16. Deuterium ingress at rolled joints in Embalse nuclear power plant

    International Nuclear Information System (INIS)

    Ramos Nervi, J. E.; Schroeter, F.

    2013-01-01

    Deuterium ingress model at the Rolled Joint has been extensively used for CANDU Nuclear Power Plants Operators in the Life Management of the Pressure Tubes. The importance of understanding the model is vital to avoid delayed hydride cracking at the Rolled Joint. This work reports the first step on develop the model presented on literature to be used in Argentinean CANDU 6, Embalse Nuclear Power Plant. (author)

  17. Evaluation of Required Water Sources during Extended Loss of All AC Power for CANDU NPPs

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Woo Jae; Lee, Kyung Jin; Kim, Min Ki; Kim, Keon Yeop; Park, Da Hee; Oh, Seo Bin [FNC Technology Co., Yongin (Korea, Republic of); Chang, Young Jin; Byun, Choong Seop [KHNP, Daejeon (Korea, Republic of)

    2016-10-15

    Fukushima accident was caused by lasting long hours of Station Black-Out (SBO) triggered from natural disaster. This accident had resulted in the reactor core damage. The purpose of this study is to evaluate the required water sources to maintain hot standby conditions until 72 hours during ELAP situation. The analysis was performed with CATHENA code. CATHENA code has been developed for the best-estimated transient simulation of CANDU plants. This study was carried out to evaluate the strategy to maintain hot standby conditions during ELAP situation in CANDU reactors. In this analysis, water was supplied to SG by MSSV open and by the gravity feed. It can cool the core without damage until the dousing tank depletion. Before dousing tank depletion, the emergency water supply pump was available by emergency power restoration. The pump continuously fed water to SG. So it is expected that the reactor core can be cooled down without damage for 72 hours if water source is enough to feed. This result is useful to make a strategy against SBO including ELAP situation.

  18. Team CANDU : ready for the marketplace

    International Nuclear Information System (INIS)

    Howieson, J.Q.

    2007-01-01

    This paper outlines the partnership between AECL and a number of leading global nuclear suppliers to market the Candu power reactor. The mission of the CANDU team is to develop market opportunities for CANDU technology and deliver successful CANDU projects

  19. Power generation by nuclear power plants

    International Nuclear Information System (INIS)

    Bacher, P.

    2004-01-01

    Nuclear power plays an important role in the world, European (33%) and French (75%) power generation. This article aims at presenting in a synthetic way the main reactor types with their respective advantages with respect to the objectives foreseen (power generation, resources valorization, waste management). It makes a fast review of 50 years of nuclear development, thanks to which the nuclear industry has become one of the safest and less environmentally harmful industry which allows to produce low cost electricity: 1 - simplified description of a nuclear power generation plant: nuclear reactor, heat transfer system, power generation system, interface with the power distribution grid; 2 - first historical developments of nuclear power; 3 - industrial development and experience feedback (1965-1995): water reactors (PWR, BWR, Candu), RBMK, fast neutron reactors, high temperature demonstration reactors, costs of industrial reactors; 4 - service life of nuclear power plants and replacement: technical, regulatory and economical lifetime, problems linked with the replacement; 5 - conclusion. (J.S.)

  20. Experimental validation of Pu-Sm evolution model for CANDU-6 power transients

    International Nuclear Information System (INIS)

    Coutsiers, Eduardo E.; Pomerantz, Marcelo E.; Moreno, Carlos A.

    2000-01-01

    Development of a methodology to evaluate the reactivity produced by Pu-Sm transient, effect displayed after power transients. This methodology allows to predict the behavior of liquid zones with which the fine control of CANDU reactor power is made. With this information, it is easier to foresee the refueling demand after power movements. The comparison with experimental results showed good agreement. (author)

  1. Report of the COG/IAEA international workshop on managing nuclear safety at CANDU (PHWR) plants. Working material

    International Nuclear Information System (INIS)

    1997-01-01

    The workshop, hosted by COG and co-sponsored by the International Atomic Energy Agency (IAEA, Vienna) was held in Toronto, April 28 - May 1st, 1997. The 40 participants included senior managers from IAEA member countries operating or constructing CANDU (PHWR) stations. All the offshore utilities with PHWR stations in Korea, Romania, India, Argentina, Pakistan, and China were present with their domestic counterparts from Ontario Hydro Nuclear, Hydro Quebec, New Brunswick Power, and AECL. The objectives of the workshop were to: provide a forum for exchange of ideas among nuclear safety managers operating CANDU (PHWR) stations and to learn from each other's experiences; to foster sharing of information on different operating approaches to managing safety and, in particular, to highlight the strategies for controlling the overall plant risk to a low level; to identify and discuss issues of mutual interest pertinent to PHWR stations and to define future follow-up activities. Refs, figs

  2. Report of the COG/IAEA international workshop on managing nuclear safety at CANDU (PHWR) plants. Working material

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-01

    The workshop, hosted by COG and co-sponsored by the International Atomic Energy Agency (IAEA, Vienna) was held in Toronto, April 28 - May 1st, 1997. The 40 participants included senior managers from IAEA member countries operating or constructing CANDU (PHWR) stations. All the offshore utilities with PHWR stations in Korea, Romania, India, Argentina, Pakistan, and China were present with their domestic counterparts from Ontario Hydro Nuclear, Hydro Quebec, New Brunswick Power, and AECL. The objectives of the workshop were to: provide a forum for exchange of ideas among nuclear safety managers operating CANDU (PHWR) stations and to learn from each other`s experiences; to foster sharing of information on different operating approaches to managing safety and, in particular, to highlight the strategies for controlling the overall plant risk to a low level; to identify and discuss issues of mutual interest pertinent to PHWR stations and to define future follow-up activities. Refs, figs.

  3. Non-electrical CANDU applications

    International Nuclear Information System (INIS)

    Hopwood, Jerry; Kuran, Sermet; Zhou, Xi; Ivanco, Michael; Rolfe, Brian; Mancuso, Connie; Duffey, Romney

    2005-01-01

    AECL has performed studies to utilize CANDU nuclear energy in areas other than electrical generation. The studies presented in this paper include using CANDU for applications in non-traditional areas which expand the use of zero-greenhouse gas energy source. The Oil sands industry demands significant energy input and the majority of the energy required for bitumen extraction is steam and hot water. As the primary production of a CANDU plant is steam, it can satisfy the steam and hot water requirement without a major modification to the Nuclear Steam Plant (NSP). Reverse Osmosis (RO) has been identified by the IAEA as the most promising method for nuclear desalination. Since the RO desalination efficiency increases as its feedwater temperature rises, using condenser cooling water from a CANDU plant as the feedwater for a RO plant and sharing other facilities between these two plants results in significant benefits in capital and operating costs of a desalination plant. Electrolysis powered by nuclear-generated electricity is the technology currently available to produce hydrogen without greenhouse gas emissions. By using the cheaper electricity available at off-peak periods in an open electricity market, this technology could be economically competitive, improve overall energy system efficiency and reduce overall energy system carbon intensity. The paper summarizes the background, technical approaches, feasibility considerations, along with economic comparisons between CANDU nuclear energy and the traditional energy sources for each study. The results show that the CANDU technology is a promising energy source for various industries. (author)

  4. CANDU safety under severe accidents

    International Nuclear Information System (INIS)

    Snell, V.G.; Howieson, J.Q.; Frescura, G.M.; King, F.; Rogers, J.T.; Tamm, H.

    1988-01-01

    The characteristics of the CANDU reactor relevant to severe accidents are set first by the inherent properties of the design, and second by the Canadian safety/licensing approach. Probabilistic safety assessment studies have been performed on operating CANDU plants, and on the 4 x 880 MW(e) Darlington station now under construction; furthermore a scoping risk assessment has been done for a CANDU 600 plant. They indicate that the summed severe core damage frequency is of the order of 5 x 10 -6 /year. CANDU nuclear plant designers and owner/operators share information and operational experience nationally and internationally through the CANDU Owners' Group (COG). The research program generally emphasizes the unique aspects of the CANDU concept, such as heat removal through the moderator, but it has also contributed significantly to areas generic to most power reactors such as hydrogen combustion, containment failure modes, fission product chemistry, and high temperature fuel behaviour. Abnormal plant operating procedures are aimed at first using event-specific emergency operating procedures, in cases where the event can be diagnosed. If this is not possible, generic procedures are followed to control Critical Safety Parameters and manage the accident. Similarly, the on-site contingency plans include a generic plan covering overall plant response strategy, and a specific plan covering each category of contingency

  5. The CANDU 9 distributed control system design process

    International Nuclear Information System (INIS)

    Harber, J.E.; Kattan, M.K.; Macbeth, M.J.

    1997-01-01

    Canadian designed CANDU pressurized heavy water nuclear reactors have been world leaders in electrical power generation. The CANDU 9 project is AECL's next reactor design. Plant control for the CANDU 9 station design is performed by a distributed control system (DCS) as compared to centralized control computers, analog control devices and relay logic used in previous CANDU designs. The selection of a DCS as the platform to perform the process control functions and most of the data acquisition of the plant, is consistent with the evolutionary nature of the CANDU technology. The control strategies for the DCS control programs are based on previous CANDU designs but are implemented on a new hardware platform taking advantage of advances in computer technology. This paper describes the design process for developing the CANDU 9 DCS. Various design activities, prototyping and analyses have been undertaken in order to ensure a safe, functional, and cost-effective design. (author)

  6. Manufacturing opportunities in the Canadian CANDU and heavy water programs

    International Nuclear Information System (INIS)

    Reny, J.P.

    The volume of business available to Canadian manufacturers of CANDU power plant and heavy water plant components is analyzed over about the next 10 years. Implications of exported nuclear technology and plants are explored. (E.C.B.)

  7. Learning from experience: feedback to CANDU design

    International Nuclear Information System (INIS)

    Allen, P.J.; Hopwood, J.M.; Rousseau, G.P.

    1998-01-01

    AECL's main product line is based on two single unit CANDU nuclear power plant designs; CANDU 6 and CANDU 9, each of which is based on successfully operating CANDU plants. AECL's CANDU development program is based upon evolutionary improvement. The evolutionary design approach ensures the maximum degree of operational provenness. It also allows successful features of today's plants to be retained while incorporating improvements as they develop to the appropriate level of design maturity. A key component of this evolutionary development is a formal process of gathering and responding to feedback from: NPP operation, construction and commissioning; regulatory input; equipment supplier input; R and D results; market input. The progresses for gathering and implementing the experience feedback and a number of recent examples of design improvements from this feedback process are described in the paper. (author)

  8. Dura Seal recommendations for nuclear power plants

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    Reactor systems (BWR, PWR and Candu) are briefly reviewed with reference to the pumping services encountered in each system, to indicate the conditions imposed on mechanical seals for nuclear power plant liquid handling equipment. A description of the Dura Seals used in each service is included. (U.K.)

  9. Nuclear power - replacement of pressure tubes in CANDU reactors

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    The CANDU pressure tube reactor is an effective electricity generator. While most units have been built in Canada, units are successfully operated in Argentina and Korea as well as India and Pakistan, which have early versions of the same concept. Units are also under construction in Korea and Romania. The main constructional components of a CANDU core are the calandria vessel, the fuel channels and the reactivity control mechanisms. The fuel channel, in particular the pressure tubes, see an environment comprising high flux, high temperature water at high pressures, which induces changes in the properties and dimensions of the channel components. From the first, fuel channels were designed to be replaced because of the difficulty in predicting the behaviour of zirconium alloys in such service over a long period of time. In fact some phenomena, that were not known at the time of the earliest designs, have led to unacceptable changes in the properties of the channels and these early reactors have had to be retubed at half their intended life. These deficiencies have been corrected in the latest designs and fuel channels in reactors that have commenced operation over the last 10 years, are predicted to reach the intended 30 years life before replacement is necessary. The changing of fuel channels, the details and experience of which are explained, has been shown to be an effective way of refurbishing the CANDU reactor, extending its lifetime a further 25-30 years. (author)

  10. Enhanced candu 6 reactor: status

    International Nuclear Information System (INIS)

    Azeez, S.; Girouard, P.

    2006-01-01

    The CANDU 6 power reactor is visionary in its approach, renowned for its on-power refuelling capability and proven over years of safe, economical and reliable power production. Developed by Atomic Energy of Canada Limited (AECL), the CANDU 6 design offers excellent performance utilizing state-of-the-art technology. The first CANDU 6 plants went into service in the early 1980s as leading edge technology and the design has been continuously advanced to maintain superior performance with an outstanding safety record. The first set of CANDU 6 plants - Gentilly 2 and Point Lepreau in Canada, Embalse in Argentina and Wolsong- Unit 1 in Korea - have been in service for more than 22 years and are still producing electricity at peak performance; to the end of 2004, their average Lifetime Capacity Factor was 83.2%. The newer CANDU 6 units in Romania (Cernavoda 1), Korea (Wolsong-Units 2, 3 and 4) and Qinshan (Phase III- Units 1 and 2) have also been performing at outstanding levels. The average lifetime Capacity Factor of the 10 CANDU 6 operating units around the world has been 87% to the end of 2004. Building on these successes, AECL is committed to the further development of this highly successful design, now focussing on meeting customers' needs for reduced costs, further improvements to plant operation and performance, enhanced safety and incorporating up-to-date technology, as warranted. This has resulted in AECL embarking on improving the CANDU 6 design through an upgraded product termed the ''Enhanced CANDU 6'' (EC6), which incorporates several attractive but proven features that make the CANDU 6 reactor even more economical, safer and easier to operate. Some of the key features that are being incorporated into the EC6 include increasing the plant's power output, shortening the overall project schedule, decreasing the capital cost, dealing with obsolescence issues, optimizing maintenance outages and incorporating lessons learnt through feedback obtained from the

  11. Operation of CANDU power reactor in thorium self-sufficient fuel cycle

    Indian Academy of Sciences (India)

    This paper presents the results of calculations for CANDU reactor operation in thorium fuel cycle. Calculations are performed to estimate the feasibility of operation of heavy-water thermal neutron power reactor in self-sufficient thorium cycle. Parameters of active core and scheme of fuel reloading were considered to be the ...

  12. Cernavoda CANDU severe accident evaluation

    International Nuclear Information System (INIS)

    Negut, G.; Marin, A.

    1997-01-01

    The papers present the activities dedicated to Romania Cernavoda Nuclear Power Plant first CANDU Unit severe accident evaluation. This activity is part of more general PSA assessment activities. CANDU specific safety features are calandria moderator and calandria vault water capabilities to remove the residual heat in the case of severe accidents, when the conventional heat sinks are no more available. Severe accidents evaluation, that is a deterministic thermal hydraulic analysis, assesses the accidents progression and gives the milestones when important events take place. This kind of assessment is important to evaluate to recovery time for the reactor operators that can lead to the accident mitigation. The Cernavoda CANDU unit is modeled for the of all heat sinks accident and results compared with the AECL CANDU 600 assessment. (orig.)

  13. Some novel on-power refuelling features of CANDU stations

    International Nuclear Information System (INIS)

    Erwin, D.; Pendlebury, B.; Watson, J.F.; Welch, A.C.

    1976-01-01

    Part A of the paper describes the reasons for, and advantages resulting from, the use of flow assisted refuelling in the CANDU type nuclear reactors at the Pickering Generating Station. A separate fuel handling system is used for each reactor unit, as distinct from the system employed at the Bruce Generating station, where the fuel handling system is shared among several units. Part B of the paper describes some of the advantages of the shared concept with particular emphasis on the availability of the fuel handling system. (author)

  14. Advanced CANDU reactors

    International Nuclear Information System (INIS)

    Dunn, J.T.; Finlay, R.B.; Olmstead, R.A.

    1988-12-01

    AECL has undertaken the design and development of a series of advanced CANDU reactors in the 700-1150 MW(e) size range. These advanced reactor designs are the product of ongoing generic research and development programs on CANDU technology and design studies for advanced CANDU reactors. The prime objective is to create a series of advanced CANDU reactors which are cost competitive with coal-fired plants in the market for large electricity generating stations. Specific plant designs in the advanced CANDU series will be ready for project commitment in the early 1990s and will be capable of further development to remain competitive well into the next century

  15. Environmental effects on the response of self-powered flux detectors in CANDU reactors

    International Nuclear Information System (INIS)

    Lynch, G.F.; Shields, R.B.; Joslin, C.W.

    1976-01-01

    Self-powered flux detectors are playing an increasingly important role in the control and safety systems of CANDU-type reactors. In this paper we report on recent experiments to determine how local reactor conditions affect the output signals from self-powered detectors with vanadium, platinum and cobalt emitters. The results are interpreted in terms of variations in the local neutron, γ-ray and electron fluxes. (author)

  16. Development of techniques for radwaste systems in CANDU power stations

    International Nuclear Information System (INIS)

    Bourns, W.T.; Buckley, L.P.; Burrill, K.A.

    1979-04-01

    Techniques to reduce the volume of CANDU reactor wastes and to bitumenize them are being developed at Chalk River Nuclear Laboratories. Reverse osmosis is suitable for initial purification of dilute radioactive aqueous wastes. Tubular membranes are used to concentrate wastes to 5 weight percent solids, and while the membranes do foul, they may be cleaned mechanically, chemically, or with fresh feed. A wiped-film evaporator then concentrates the retentate to a 20 weight-percent slurry. A twin-screw extruder-evaporator has been used to bitumenize this slurry, and it will also handle ion exchange resin and dry incinerator ash. Work on a wiped-film evaporator as a bitumenizer for various feeds is in progress. More experience in handling solid feeds is needed before work can proceed to the demonstraton phase. (auth)

  17. CANDU technology for generation III + AND IV reactors

    International Nuclear Information System (INIS)

    Torgerson, D.F.

    2005-01-01

    Atomic Energy of Canada Limited (AECL) is the original developer of the CANDU?reactor, one of the three major commercial power reactor designs now used throughout the world. For over 60 years, AECL has continued to evolve the CANDU design from the CANDU prototypes in the 1950s and 1960s through to the second generation reactors now in operation, including the Generation II+ CANDU 6. The next phase of this evolution, the Generation III+ Advanced CANDU ReactorTM (ACRTM), continues the strategy of basing next generation technology on existing CANDU reactors. Beyond the ACR, AECL is developing the Generation IV CANDU Super Critical Water Reactor. Owing to the evolutionary nature of these advanced reactors, advanced technology from the development programs is also being applied to operating CANDU plants, for both refurbishments and upgrading of existing systems and components. In addition, AECL is developing advanced technology that covers the entire life cycle of the CANDU plant, including waste management and decommissioning. Thus, AECL maintains state-of-the-art expertise and technology to support both operating and future CANDU plants. This paper outlines the scale of the current core knowledge base that is the foundation for advancement and support of CANDU technology. The knowledge base includes advancements in materials, fuel, safety, plant operations, components and systems, environmental technology, waste management, and construction. Our approach in each of these areas is to develop the underlying science, carry out integrated engineering scale tests, and perform large-scale demonstration testing. AECL has comprehensive R and D and engineering development programs to cover all of these elements. The paper will show how the ongoing expansion of the CANDU knowledge base has led to the development of the Advanced CANDU Reactor. The ACR is a Generation III+ reactor with substantially reduced costs, faster construction, and enhanced passive safety and operating

  18. CANDU: study and review

    Energy Technology Data Exchange (ETDEWEB)

    Morad, César M., E-mail: cesar.morad@usp.br [Universidade de São Paulo (POLI/USP), SP (Brazil). Escola Politécnica; Stefani, Giovanni L. de, E-mail: giovanni.stefani@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil); Santos, Thiago A. dos, E-mail: thiago.santos@ufabc.edu.br [Universidade Federal do ABC (UFABC), Santo André, SP (Brazil)

    2017-07-01

    The CANDU (Canadian Deuterium Uranium) is a nuclear reactor developed by AECL (Atomic Energy of Canada Limited). The first small-scale reactor is known as NPD and was made in 1955 and commenced operation in 1962. It is a pressurized heavy water reactor and uses D2O as moderator and coolant and therefore uses natural uranium as fuel. There have been two major types of CANDU reactors, the original design of around 500 MWe that was intended to be used in multi-reactor installations in large plants, and the rationalized CANDU6 which has units in Argentina, South Korea, Pakistan, Romania and China. Throughout the 1980s and 90s the nuclear power market suffered a major crash, with few new plants being constructed in North America or Europe. Design work continued through, however, and a number of new design concepts were introduced that dramatically improved safety, capital costs, economics and overall performance. These Generation III+ and Generation IV machines became a topic of considerable interest in the early 2000s as it appeared a nuclear renaissance was underway and large numbers of new reactors would be built over the next decade. The present work aims to study the reactors of the CANDU type, exploring from its creation to studies directed to G-III and G-IV reactors. (author)

  19. CANDU: study and review

    International Nuclear Information System (INIS)

    Morad, César M.; Santos, Thiago A. dos

    2017-01-01

    The CANDU (Canadian Deuterium Uranium) is a nuclear reactor developed by AECL (Atomic Energy of Canada Limited). The first small-scale reactor is known as NPD and was made in 1955 and commenced operation in 1962. It is a pressurized heavy water reactor and uses D2O as moderator and coolant and therefore uses natural uranium as fuel. There have been two major types of CANDU reactors, the original design of around 500 MWe that was intended to be used in multi-reactor installations in large plants, and the rationalized CANDU6 which has units in Argentina, South Korea, Pakistan, Romania and China. Throughout the 1980s and 90s the nuclear power market suffered a major crash, with few new plants being constructed in North America or Europe. Design work continued through, however, and a number of new design concepts were introduced that dramatically improved safety, capital costs, economics and overall performance. These Generation III+ and Generation IV machines became a topic of considerable interest in the early 2000s as it appeared a nuclear renaissance was underway and large numbers of new reactors would be built over the next decade. The present work aims to study the reactors of the CANDU type, exploring from its creation to studies directed to G-III and G-IV reactors. (author)

  20. CANDU fuel

    International Nuclear Information System (INIS)

    MacEwan, J.R.; Notley, M.J.F.; Wood, J.C.; Gacesa, M.

    1982-09-01

    The direction of CANDU fuel development was set in 1957 with the decision to build pressure tube reactors. Short - 50 cm long - rodded bundles of natural UO 2 clad in Zircaloy were adopted to facilitate on-power fuelling to improve uranium utilization. Progressive improvements were made during 25 years of development, involving 650 man years and 180 million dollars. Today's CANDU bundle is based on the knowledge gained from extensive irradiation testing and experience in power reactors. The main thrust of future development is to demonstrate that the present bundle is suitable, with minor modifications, for thorium fuels

  1. CANDU project development

    Energy Technology Data Exchange (ETDEWEB)

    Hedges, K R [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

    1996-12-31

    Advanced CANDU reactor design strategy follows an evolutionary approach, taking manageable steps in the development of power plants from today`s available designs, and in parallel carrying out longer-term studies to develop future-generation reactor concepts. The major emphasis is on safety, on on reducing cost and schedule. New features are developed and thoroughly proof-tested before introduction into designs, in order to maximize owner confidence. (author). 4 figs.

  2. CANDU project development

    International Nuclear Information System (INIS)

    Hedges, K.R.

    1995-01-01

    Advanced CANDU reactor design strategy follows an evolutionary approach, taking manageable steps in the development of power plants from today's available designs, and in parallel carrying out longer-term studies to develop future-generation reactor concepts. The major emphasis is on safety, on on reducing cost and schedule. New features are developed and thoroughly proof-tested before introduction into designs, in order to maximize owner confidence. (author). 4 figs

  3. Advantages of butterfly valves for power plants

    International Nuclear Information System (INIS)

    Lapadat, J.T.

    1977-01-01

    Butterfly valves are increasingly used in nuclear power plants. They are used in CANDU reactors for class 2 and 3 service, to provide emergency and tight shutoff valves for all inlets and outlets of heat exchangers and all calandria penetrations. Guidelines for meeting nuclear power plant valve specifications are set out in ASME Section 3, Nuclear Power Plant Components. Some details of materials of construction, type of actuator, etc., for various classes of nuclear service are tabulated in the present article. The 'fishtail' butterfly valve is an improved design with reduced drag, as is illustrated and explained. (N.D.H.)

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  5. CANDU advanced fuel cycles

    International Nuclear Information System (INIS)

    Slater, J.B.

    1986-03-01

    This report is based on informal lectures and presentations made on CANDU Advanced Fuel Cycles over the past year or so, and discusses the future role of CANDU in the changing environment for the Canadian and international nuclear power industry. The changing perspectives of the past decade lead to the conclusion that a significant future market for a CANDU advanced thermal reactor will exist for many decades. Such a reactor could operate in a stand-alone strategy or integrate with a mixed CANDU-LWR or CANDU-FBR strategy. The consistent design focus of CANDU on enhanced efficiency of resource utilization combined with a simple technology to achieve economic targets, will provide sufficient flexibility to maintain CANDU as a viable power producer for both the medium- and long-term future

  6. Validation of the COBRA code for dry out power calculation in CANDU type advanced fuels

    International Nuclear Information System (INIS)

    Daverio, Hernando J.

    2003-01-01

    Stern Laboratories perform a full scale CHF testing of the CANFLEX bundle under AECL request. This experiment is modeled with the COBRA IV HW code to verify it's capacity for the dry out power calculation . Good results were obtained: errors below 10 % with respect to all data measured and 1 % for standard operating conditions in CANDU reactors range . This calculations were repeated for the CNEA advanced fuel CARA obtaining the same performance as the CANFLEX fuel. (author)

  7. Prediction of hydrogen distribution in the reactor building in CANDU6 plant

    International Nuclear Information System (INIS)

    Jin, Y.; Song, Y.

    2008-01-01

    The CANDU plants have a lot of zircaloy. The fuel cladding, calandria tubes and pressure tubes are made of zircaloy. The zircaloy can be oxidized and hydrogen is generated during severe accident progression. The detonation or deflagration to detonation transition (DDT) due to hydrogen combustion may occur if the local hydrogen concentration or global hydrogen concentration exceeds certain value. The detonation may result in the rupture of the reactor building. The inside of the reactor building of CANDU plants is complex. So prediction of hydrogen distribution in the reactor building is important. This prediction is made using ISAAC code and GOTHIC code. ISAAC code partitioned the reactor building in to 7 compartments. GOTHIC code modeled the CANDU6 reactor building using 12 nodes. The hydrogen concentrations in the various compartments in the reactor building are compared. GOTHIC code slightly underpredicts hydrogen concentration in the F/M rooms than ISAAC code, but trend is same. The hydrogen concentration in the boiler room and the moderator room shows almost same as for both codes. (author)

  8. Role of operator response guidelines in CANDU 9 design program

    International Nuclear Information System (INIS)

    Jaitly, R.K.

    2000-01-01

    The CANDU 9 is a large version of the CANDU Pressurized Heavy Water Reactor (PHWR) system developed in Canada. With an electrical output of approximately 935 MWe, the CANDU 9 complements the established mid-size CANDU 6 (700 MWe) and makes use of proven technology updated with state of the art features resulting from ongoing development. The CANDU 9 builds on the reactor and process system designs of the operating Darlington and Bruce B plants, and incorporates a modified CANDU 6 station layout, as well as improved construction methods and operational features. A high level of standardization has always been a feature of CANDU reactors. This theme is emphasized in the CANDU 9; all key components (reactor core, steam generators, coolant pumps, pressure tubes, etc.) are of the same design as those proven in service in the operating CANDU power stations. Including Probabilistic Safety Assessment (PSA) as part of the CANDU 9 design process from the outset of the program was seen as key to ensuring completeness of safety related requirements. The PSA work provided an in-depth understanding of the plant response to various postulated accidents. As well, the time frame for recovery and the related operator actions were identified. This information together with AECL's experience in supporting the development of Emergency Operating Procedures (EOPs) for the operating CANDU reactors are the basis for preparation of CANDU 9 Operator Response Guidelines (ORGs). Technical content, format and human factors considerations adopted for the ORGs are such that these can be readily converted to EOPs. The scope of ORGs includes generic as well as event specific ORGs. This dual approach is required to provide defense-in-depth. This paper describes the process used to prepare ORGs for the CANDU 9 reactor and discusses important benefits gained from the application of ORGs as input to the control center design and future preparation of the EOPs. (author)

  9. Assessment of System Behavior and Actions Under Loss of Electric Power For CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Kang, San Ha; Moon, Bok Ja; Kim, Seoung Rae [Nuclear Engineering Service and Solution Co., Ltd., Daejeon (Korea, Republic of)

    2014-05-15

    For the analysis, the CANDU-6 plant in Korea is considered and only the passive components are operable. The other systems are assumed to be at failed condition due to the loss of electric power. At this accident, only the inventories remained in the primary heat transport system (PHTS) and steam generator can be used for the decay heat removal. Due to the transfer of decay heat, the inventory of steam generator secondary side is discharged to the air through passive operation of main steam safety valves (MSSVs). After the steam generators are dried, the PHTS is over-pressurized and the coolant is discharged to fuelling machine vault through passive operation of degasser condenser tank relief valves (DCRVs). Under this situation, the maintenance of the integrity of PHTS is important for the protection of radionuclides release to the environment. Thus, deterministic analysis using CATHENA code is carried out for the simulation of the accident and the appropriate operator action is considered. The loss of electric power results in the depletion of steam generator inventory which is necessary for the decay heat removal. If only the passive system is credited, the PT can be failed after the steam generator is depleted. For the prevention of the PT failure, the feedwater should be supplied to the steam generator before 4,800s after the accident. The feedwater can be supplied using water in dousing tank if the steam generators are depressurized. The decay heat from the core is removed through natural circulation if the feedwater can be supplied continuously.

  10. Change in CANDU-6 reactivity following a power reduction at low PHT purity

    International Nuclear Information System (INIS)

    Whitlock, J.J.; Soulard, M.R.; Baudouin, A.

    1995-01-01

    The reactivity effect of a power reduction in CANDU-6 is examined using a three-dimensional, steady-state, coupled neutronics/thermalhydraulics methodology, starting from a global irradiation distribution matched to site data. The power reduction is sufficient to suppress coolant boiling in the fuel channels, and thus the significant parameters affecting reactivity are an increase in coolant density and a decrease in fuel temperature. These individual components are estimated using infinite-lattice-cell methodology. The effect of using newer methodology, particularly for the thermalhydraulic analysis, is examined by comparison with previous simulations. (author). 10 refs., 7 tabs., 1 fig

  11. The CANDU 80

    International Nuclear Information System (INIS)

    Hart, R.S.

    1998-01-01

    AECL has completed the conceptual design of a small CANDU plant with an output, in the range of 300 MWth (called the CANDU 80), suitable for a variety of electrical and co-generation applications including desalination, oil sands oil extraction and processing, and the provision of electricity and heat to areas with low demand. This paper provides a brief overview of the CANDU 80, and discusses key features contributing to safety and operational margins

  12. Canadian CANDU plant historical data systems. A review and look to the future

    International Nuclear Information System (INIS)

    Deverno, M.; De Grosbois, J.; Pothier, H.

    1996-01-01

    As part of several CANDU Owner's Group (COG) projects, AECL has conducted a review of current approaches and investigated solutions to plant process data collection, management, and use. Emphasis was placed on understanding the existing functionality and uses of plant data systems, their future needs and benefits. The result is a vision of a plant-wide Historical Data System (HDS) providing seamless access to both near real-time and historical data, user tool-kits for data visualization and analysis, and data management of the larger volume of data acquired during the life of a plant. HDS technology is critical to the implementation of technical surveillance and analysis, predictive and preventative maintenance programs, and other efforts necessary to enhance plant safety, availability, production and productivity. HDS technology will lead to higher capability and capacity factors while minimizing operations, maintenance, and administration (OM and A) costs. (author)

  13. Canadian CANDU plant historical data systems. A review and look to the future

    Energy Technology Data Exchange (ETDEWEB)

    Deverno, M; De Grosbois, J [Atomic Energy of Canada Limited, Chalk River, ON (Canada); Pothier, H [Atomic Energy of Canada Limited, Mississauga, ON (Canada); and others

    1997-12-31

    As part of several CANDU Owner`s Group (COG) projects, AECL has conducted a review of current approaches and investigated solutions to plant process data collection, management, and use. Emphasis was placed on understanding the existing functionality and uses of plant data systems, their future needs and benefits. The result is a vision of a plant-wide Historical Data System (HDS) providing seamless access to both near real-time and historical data, user tool-kits for data visualization and analysis, and data management of the larger volume of data acquired during the life of a plant. HDS technology is critical to the implementation of technical surveillance and analysis, predictive and preventative maintenance programs, and other efforts necessary to enhance plant safety, availability, production and productivity. HDS technology will lead to higher capability and capacity factors while minimizing operations, maintenance, and administration (OM and A) costs. (author) 15 refs., 7 figs.

  14. Diagnostic Technology Development for Core Internal Structure in CANDU reactor

    International Nuclear Information System (INIS)

    Jung, Hyun Kyu; Cheong, Y. M.; Lee, Y. S. and others

    2005-04-01

    Degradation of critical components of nuclear power plants has become important as the operating years of plants increase. The necessity of degradation study including measurement and monitoring technology has increased continuously. Because the fuel channels and the neighboring sensing tubes and control rods are particularly one of the critical components in CANDU nuclear plant, they are treated as a major research target in order to counteract the possible problems and establish the counterplan for the CANDU reactor safety improvement. To ensure the core structure integrity in CANDU nuclear plant, the following 2 research tasks were performed: Development of NDE technologies for the gap measurement between the fuel channels and LIN tubes. Development of vibration monitoring technology of the fuel channels and sensing tubes. The technologies developed in this study could contribute to the nuclear safety and estimation of the remaining life of operating CANDU nuclear power plants

  15. Computer functions in overall plant control of candu generating stations

    International Nuclear Information System (INIS)

    Chou, Q.B.; Stokes, H.W.

    1976-01-01

    System Planning Specifications form the basic requirements for the performance of the plant including its response to abnormal situations. The rules for the computer control programs are devised from these, taking into account limitations imposed by the reactor, heat transport and turbine-generator systems. The paper outlines these specifications and the limitations imposed by the major items of plant equipment. It describes the functions of each of the main programs, their interactions and the control modes used in the existing Ontario Hydro's nuclear station or proposed for future stations. Some simulation results showing the performance of the overall unit control system and plans for future studies are discussed. (orig.) [de

  16. CANDU development

    International Nuclear Information System (INIS)

    Brooks, G.L.

    1981-06-01

    Evolution of the 950 MW(e) CANDU reactor is summarized. The design was specifically aimed at the export market. Factors considered in the design were that 900-1000 MW is the maximum practical size for most countries; many countries have warmer condenser cooling water than Canada; the plant may be located on coastal sites; seismic requirements may be more stringent; and the requirements of international, as well as Canadian, standards must be satisfied. These considerations resulted in a 600-channel reactor capable of accepting condenser cooling water at 32 0 C. To satisfy the requirement for a proven design, the 950 MW CANDU draws upon the basic features of the Bruce and Pickering plants which have demonstrated high capacity factors

  17. Technical description of the CANDU nuclear power system

    International Nuclear Information System (INIS)

    Husseini, S.D.

    1977-01-01

    The lecture consists of: 1.) KANUPP description: general - Containment- Reactor Cooling System - Moderator Helium System. 2.) Operating experience: Plant Performances (as applicable to Kanupp) - Major failures of primary system - Performance of primary circulating pumps. - Heavy Water Leakage Control. - Radiation Dose Control - Inadvertant Operations. (orig.) [de

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

    International Nuclear Information System (INIS)

    Kim, Yonghee

    2015-01-01

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

  19. THE IMPACT OF POWER COEFFICIENT OF REACTIVITY ON CANDU 6 REACTORS

    Directory of Open Access Journals (Sweden)

    D. KASTANYA

    2013-10-01

    Full Text Available The combined effects of reactivity coefficients, along with other core nuclear characteristics, determine reactor core behavior in normal operation and accident conditions. The Power Coefficient of Reactivity (PCR is an aggregate indicator representing the change in reactor core reactivity per unit change in reactor power. It is an integral quantity which captures the contributions of the fuel temperature, coolant void, and coolant temperature reactivity feedbacks. All nuclear reactor designs provide a balance between their inherent nuclear characteristics and the engineered reactivity control features, to ensure that changes in reactivity under all operating conditions are maintained within a safe range. The CANDU® reactor design takes advantage of its inherent nuclear characteristics, namely a small magnitude of reactivity coefficients, minimal excess reactivity, and very long prompt neutron lifetime, to mitigate the demand on the engineered systems for controlling reactivity and responding to accidents. In particular, CANDU reactors have always taken advantage of the small value of the PCR associated with their design characteristics, such that the overall design and safety characteristics of the reactor are not sensitive to the value of the PCR. For other reactor design concepts a PCR which is both large and negative is an important aspect in the design of their engineered systems for controlling reactivity. It will be demonstrated that during Loss of Regulation Control (LORC and Large Break Loss of Coolant Accident (LBLOCA events, the impact of variations in power coefficient, including a hypothesized larger than estimated PCR, has no safety-significance for CANDU reactor design. Since the CANDU 6 PCR is small, variations in the range of values for PCR on the performance or safety of the reactor are not significant.

  20. The CANDU 9

    International Nuclear Information System (INIS)

    Hart, R.S.

    1994-01-01

    The CANDU 9 plants are single unit versions of the Bruce B design, incorporating relevant technical advances made in CANDU 6, and the newer Darlington and CANDU 3 designs. This paper describes the CANDU 9 480/SEU, with an electrical output of about 1050 MW. In this designation, 480 refers to the number of fuel channels, and SEU to slightly enriched uranium. Emphasis is placed on evolutionary design, and the use of well proven design features, to ensure regulatory licensability and reliable operation. Safety is enhanced through simplification and improvement of key systems and components. Relatively low energy costs result from reduced specific capital cost, reduced operating and maintenance cost, and reduced radiation exposure to personnel. Standardization is emphasized inasmuch as all key components (steam generators, heat transport pumps, pressure tubes fuelling machines etc.) ar of the same design as those in operating CANDU stations. Advanced CANDU fuel cycles are readily accommodated. 1 ref., 1 tab., 11 figs

  1. Candu Energy's Aging and Obsolescence Program and its Application to Operating Facilities and New Plant Design

    International Nuclear Information System (INIS)

    Dam, R.; Gold, R.; McCrea, L.

    2012-01-01

    While plant aging is inevitable, predictable and 'graceful 'aging' behavior can be achieved through the implementation of a comprehensive and integrated Plant Life Management (PLiM) program. Despite organizations like the IAEA and INPO placing more emphasis on equipment reliability, there is still a lack of completely integrated programs in the industry as evidenced by: - Piece-meal, often crisis-driven, implementation comprising many different, partial solutions; - Duplication of effort often seen when different groups work in 'silos'. A strategy which fits with existing plant processes and programs, and which coordinates a broad range of equipment reliability activities is key to achieving the desired results. An example of such a program is the Aging and Obsolescence Program (AOP). AOP follows application of INPO AP-913 guidance for equipment reliability. The program is augmented to include single point vulnerability identification, unified approach to short and long lived components, risk management, spare parts management, and the identification and resolution of obsolescence issues. The systematic nature of the program provides the needed foundation to old and new stations alike. For existing operating stations some of the key uses include outage interval extension, reduced forced outages, and/or outage time reduction, any of which can translate into improving plant performance, competitiveness, and significant dollars saved. Program elements applied to new plant design are commensurate with the industry direction to 'design for reliability', and has allowed Candu Energy to learn and to improve upon what it can offer to operating stations. This paper intends to describe the basic elements of Candu Energy's Aging and Obsolescence Program and will share some of the experience having applied it to existing operating stations, consider applications to support expanding regulatory requirements, and describe the integration into the design of new plants, promoting

  2. CANDU-BLW-250

    Energy Technology Data Exchange (ETDEWEB)

    Pon, G. A. [Atomic Energy of Canada Ltd, Sheridan Park, ON (Canada)

    1968-04-15

    The plant ''La Centrale nucleaire de Gentiliy'' is located between Montreal and Quebec City on the south shore of the St. Lawrence River. Startup is scheduled for 1971. A CANDU-BLW reactor is the nuclear steam generator. This reactor utilizes a heavy-water moderator, natural uranium oxide fuel, and a boiling light-water coolant. To be economic, this type of plant must have a minimum light-water inventory in the reactor core. A minimum inventory is obtained (a) by reducing the cross-sectional area for coolant flow to a minimum, and (b) by operating at a low coolant density. In CANDU-BLW-250, this is accomplished by operating a closed spaced fuel rod bundle at high steam quality. These features and others in the BLW concept lead to a number of areas of concern and they are summarized below: (i) Heat transfer. It is intended that under normal operating conditions the fuel sheaths will always be wetted with coolant. Some experiments and backup calculations are presented to support this specification. (ii) Hydrodynamic stability. Experiments and analysis indicate that the plant has a considerable over-power capacity before instability is predicted. (iii) Control. This plant does have a positive power coefficient and the transient performance with various disturbances is detailed. (iv) Safety. The positive power coefficient leads to concern over the loss of coolant accident. The results of some accident analyses are presented. (author)

  3. CANDU-BLW-250

    Energy Technology Data Exchange (ETDEWEB)

    Pon, G A

    1967-09-15

    The plant 'La Centrale nucleaire de Gentilly' is located between Montreal and Quebec City on the south shore of the St. Lawrence River and start-up is scheduled for 1971. A CANDU-BLW reactor is the nuclear steam generator. his reactor utilizes a heavy water moderator, natural uranium oxide fuel, and a boiling light water coolant. To be economic, this type of plant must have a minimum light water inventory in the reactor core. A minimum inventory is obtained (a) by reducing the cross-sectional area for coolant flow to a minimum, and (b) by operating at a low-coolant density. In CANDU-BLW-250, this is accomplished by operating a closed spaced fuel rod bundle at high steam quality. These features and others in the BLW concept lead to a number of areas of concern and they are summarized below: (1) Heat Transfer: It is intended that under normal operating conditions the fuel sheaths will always be wetted with coolant. (ii) Hydrodynamic Stability: Experiments and analysis indicate that the plant has a considerable over-power capacity before instability is predicted. (iii) Control: This plant does have a positive power coefficient and the transient performance with various disturbances are detailed. (iv) Safety: The positive power coefficient leads to concern over the loss of coolant accident. The results of some accident analysis are presented. (author)

  4. Numerical simulator of the CANDU fueling machine driving desk

    International Nuclear Information System (INIS)

    Doca, Cezar

    2008-01-01

    As a national and European premiere, in the 2003 - 2005 period, at the Institute for Nuclear Research Pitesti two CANDU fueling machine heads, no.4 and no.5, for the Nuclear Power Plant Cernavoda - Unit 2 were successfully tested. To perform the tests of these machines, a special CANDU fueling machine testing rig was built and was (and is) available for this goal. The design of the CANDU fueling machine test rig from the Institute for Nuclear Research Pitesti is a replica of the similar equipment operating in CANDU 6 type nuclear power plants. High technical level of the CANDU fueling machine tests required the using of an efficient data acquisition and processing Computer Control System. The challenging goal was to build a computer system (hardware and software) designed and engineered to control the test and calibration process of these fuel handling machines. The design takes care both of the functionality required to correctly control the CANDU fueling machine and of the additional functionality required to assist the testing process. Both the fueling machine testing rig and staff had successfully assessed by the AECL representatives during two missions. At same the time, at the Institute for Nuclear Research Pitesti was/is developed a numerical simulator for the CANDU fueling machine operators training. The paper presents the numerical simulator - a special PC program (software) which simulates the graphics and the functions and the operations at the main desk of the computer control system. The simulator permits 'to drive' a CANDU fueling machine in two manners: manual or automatic. The numerical simulator is dedicated to the training of operators who operate the CANDU fueling machine in a nuclear power plant with CANDU reactor. (author)

  5. New approach to derive linear power/burnup history input for CANDU fuel codes

    International Nuclear Information System (INIS)

    Lac Tang, T.; Richards, M.; Parent, G.

    2003-01-01

    The fuel element linear power / burnup history is a required input for the ELESTRES code in order to simulate CANDU fuel behavior during normal operating conditions and also to provide input for the accident analysis codes ELOCA and SOURCE. The purpose of this paper is to present a new approach to derive 'true', or at least more realistic linear power / burnup histories. Such an approach can be used to recreate any typical bundle power history if only a single pair of instantaneous values of bundle power and burnup, together with the position in the channel, are known. The histories obtained could be useful to perform more realistic simulations for safety analyses for cases where the reference (overpower) history is not appropriate. (author)

  6. Response characteristics of self-powered flux detectors in CANDU reactors

    International Nuclear Information System (INIS)

    Allan, C.J.

    1978-05-01

    As part of the development of a new flux-detector assembly for future CANDU reactors, the sensitivities of a variety of vanadium, cobalt and platinum self-powered detectors have been determined in a simulated CANDU core installed in the ZED-2 test reactor at CRNL. While the vanadium and cobalt detectors had solid emitters, the platinum detectors were of two types, having either solid platinum emitters, or emitters consisting of a platinum sheath over an Inconel core. Almost all of the signal from the cobalt and vanadium detectors is due to neutron events in the emitters. For these detectors we have measured the total sensitivities per unit length. For the platinum detectors, reactor γ-rays and neutrons both contribute appreciably to the output signal, and in addition to the total sensitivity, we have determined the individual neutron and γ-ray sensitivities for these detectors. It was found that the detector sensitivities depend primarily on emitter diameter and that the observed variations can be fitted by means of power laws. (author)

  7. Seismic design features of the ACR Nuclear Power Plant

    International Nuclear Information System (INIS)

    Elgohary, M.; Saudy, A.; Aziz, T.

    2003-01-01

    Through their worldwide operating records, CANDU Nuclear Power Plants (NPPs) have repeatedly demonstrated safe, reliable and competitive performance. Currently, there are fourteen CANDU 6 single unit reactors operating or under construction worldwide. Atomic Energy of Canada Limited's (AECL) Advanced CANDU Reactor - the ACR. - is the genesis of a new generation of technologically advanced reactors founded on the CANDU reactor concept. The ACR is the next step in the evolution of the CANDU product line. The ACR products (ACR-700 and ACR-1000) are based on CANDU 6 (700 MWe class) and CANDU 9 (900 MWe class) reactors, therefore continuing AECL's successful approach of offering CANDU plants that appeal to a broad segment of the power generation market. The ACR products are based on the proven CANDU technology and incorporate advanced design technologies. The ACR NPP seismic design complies with Canadian standards that were specifically developed for nuclear seismic design and also with relevant International Atomic Energy Agency (IAEA) Safety Design Standards and Guides. However, since the ACR is also being offered to several markets with many potential sites and different regulatory environments, there is a need to develop a comprehensive approach for the seismic design input parameters. These input parameters are used in the design of the standard ACR product that is suitable for many sites while also maintaining its economic competitiveness. For this purpose, the ACR standard plant is conservatively qualified for a Design Basis Earthquake (DBE) with a peak horizontal ground acceleration of 0.3g for a wide range of soil/rock foundation conditions and Ground Response Spectra (GRS). These input parameters also address some of the current technical issues such as high frequency content and near field effects. In this paper, the ACR seismic design philosophy and seismic design approach for meeting the safety design requirements are reviewed. Also the seismic design

  8. Recent advances in self-powered flux detector development for CANDU reactors

    International Nuclear Information System (INIS)

    Allan, C.J.; Drewell, N.H.; Hall, D.S.

    1983-01-01

    The characteristics of self-powered flux detectors used in CANDU reactors are reviewed. Detectors with emitters of vanadium, platinum, platinum-clad Inconel and Inconel are used. Data on dynamic response, relative neutron and gamma-ray sensitivities, and burnout, obtained both from experiments and from the Monte Carlo code ICARES, are presented. Since the response of a detector depends on the relative magnitudes of the various current-producing mechanisms, the operating principles of self-powered detectors are briefly reviewed. Current research programmes are discussed. These include modifying the design of the platinum-clad Inconel detector in order to match its dynamic response to that of the fuel power and developing a prompt-responding flux-mapping detector. (author)

  9. Nuclear power plants

    International Nuclear Information System (INIS)

    1985-01-01

    Data concerning the existing nuclear power plants in the world are presented. The data was retrieved from the SIEN (Nuclear and Energetic Information System) data bank. The information are organized in table forms as follows: nuclear plants, its status and type; installed nuclear power plants by country; nuclear power plants under construction by country; planned nuclear power plants by country; cancelled nuclear power plants by country; shut-down nuclear power plants by country. (E.G.) [pt

  10. Applicability of Operational Research Techniques in CANDU Nuclear Plant Maintenance

    International Nuclear Information System (INIS)

    Doyle, E. Kevin

    2002-01-01

    As previously reported at ICONE 6 in New Orleans, 1996, and ICONE 9 in Niece, 2001, the use of various maintenance optimization techniques at Bruce has lead to cost effective preventive maintenance applications for complex systems. Innovative practices included greatly reducing Reliability Centered Maintenance (RCM) costs while maintaining the accuracy of the analysis. The optimization strategy has undergone further evolution and at the present an Integrated Maintenance Program (IMP) is being put in place. Further cost refinement of the station preventive maintenance strategy whereby decisions are based on statistical analysis of historical failure data is being evaluated. A wide range of Operational Research (OR) literature was reviewed for implementation issues and several encouraging areas were found that will assist in the current effort of evaluating maintenance optimization techniques for nuclear power production. The road ahead is expected to consist first of resolving 25 years of data issues and preserving the data via appropriate knowledge system techniques while post war demographics permit experts to input into the system. Subsequent analytical techniques will emphasize total simplicity to obtain the requisite buy in from Corporate Executives who possibly are not trained in Operational Research. Case studies of containment airlock seal failures are used to illustrate the direct applicability of stochastic processes. Airlocks and transfer chambers were chosen as they have long been known as high maintenance items. Also, the very significant financial consequences of this type of failure will help to focus the attention of Senior Management on the effort. Despite substantial investment in research, improvement in the design of the seal material or configuration has not been achieved beyond the designs completed in the 1980's. Overall, the study showed excellent agreement of the relatively quick stochastic methods with the maintenance programs produced at

  11. [Karachi Nuclear Power Plant (KANUPP), Safety Management

    Energy Technology Data Exchange (ETDEWEB)

    Hasan, S M [Karachi Nuclear Power Plant (KANUPP), Karachi (Pakistan)

    1997-12-01

    The present regime for CANDU safety management in Pakistan has evolved in line with contemporary international practice, and is essential adequate to ensure the continued safety of KANUPP and other future CANDU reactors, as confirmed by international reviews as well. But the small size of Pakistan nuclear power program poses limitations in developing - expert judgment in analysis of in-service inspection data; and own methodology for CANDU safety analysis.

  12. [Karachi Nuclear Power Plant (KANUPP), Safety Management

    International Nuclear Information System (INIS)

    Hasan, S.M.

    1997-01-01

    The present regime for CANDU safety management in Pakistan has evolved in line with contemporary international practice, and is essential adequate to ensure the continued safety of KANUPP and other future CANDU reactors, as confirmed by international reviews as well. But the small size of Pakistan nuclear power program poses limitations in developing - expert judgment in analysis of in-service inspection data; and own methodology for CANDU safety analysis

  13. The CANDU 6

    International Nuclear Information System (INIS)

    Hopwood, J.M.; Hum, J.

    1999-01-01

    The CANDU 6 is a modem nuclear power plant designed and developed under the aegis of Atomic Energy of Canada, Limited (AECL) for domestic use and for export to other countries. This design has successfully met criteria for operation and redundant safety features required by Canada and by the International Atomic Energy Agency (IAEA) and has an estimable record of performance in all applications to date. Key to this success is a defined program of design enhancement in which changes are made while retaining fundamental features proven by operating experience. Basic design features and progress toward improvements are presented here. (author)

  14. Comparative Analysis of Thermohydraulic Margins in Embalse Power Station, CARA Vs. CANDU with Cobra IV-HW

    International Nuclear Information System (INIS)

    Daverio, H; Juanico, L

    2000-01-01

    Comparative analysis of thermohydraulic margins were studied of the CANDU 37 and CARA fuel bundles (FB) in Embalse power station with COBRA IV-HW code ., the geometry of the bundle laying on the channel was particularly modeled and discussing the results in comparison with former calculations with 1/6 simetry .The CARA design with enriched uranium (0.9 %) and extended burn up lets maintain the current thermohydraulic nominal margins , while compared with CANDU 37 rods FB enriched , the CARA design permits widely improve the current margins

  15. Kansas Power Plants

    Data.gov (United States)

    Kansas Data Access and Support Center — The Kansas Power Plants database depicts, as point features, the locations of the various types of power plant locations in Kansas. The locations of the power plants...

  16. Technology transfer: The CANDU approach

    International Nuclear Information System (INIS)

    Hart, R.S.

    1998-01-01

    The many and diverse technologies necessary for the design, construction licensing and operation of a nuclear power plant can be efficiently assimilated by a recipient country through an effective technology transfer program supported by the firm long term commitment of both the recipient country organizations and the supplier. AECL's experience with nuclear related technology transfer spans four decades and includes the construction and operation of CANDU plants in five countries and four continents. A sixth country will be added to this list with the start of construction of two CANDU 6 plants in China in early 1997. This background provides the basis for addressing the key factors in the successful transfer of nuclear technology, providing insights into the lessons learned and introducing a framework for success. This paper provides an overview of AECL experience relative to the important factors influencing technology transfer, and reviews specific country experiences. (author)

  17. Analysis of power variation in a CANDU-6 with a loss of moderator

    International Nuclear Information System (INIS)

    Fan, Y.

    2008-01-01

    A loss of heavy water in a postulated small failure in the horizontal unpressurized calandria vessel of a CANDU-6 reactor will lead to a drop in the moderator level in the reactor core. The STEPBACK and SETBACK functions at the initial moment of the drop in moderator level ensure a reactor shutdown and a reduction in total reactor power during this 900 seconds postulated transient. If the STEPBACK and SETBACK functions are unavailable, the reactor's regulating system will try to compensate for the negative reactivity resulting from the loss of the moderator. This kind of compensation will lead to power distortions from top to bottom in the reactor core. .Comparisons of different moderator leakage rates were used in the analysis to determine the relationships between the power and the moderator leakage rates. Maximum bundle and channel powers obtained were insensitive to the moderator leakage rate. .In a complete analysis for a moderator leakage rate of 40 1/s, it was found that, without the STEPBACK and SETBACK functions, serious power distortions would occur during the 900 seconds transient. The maximization of bundle and channel power during this transient happened in the bottom part of the reactor , and the regulating system worsened this power distortion. .From the above analysis, it was concluded that the maximum bundle power attained during the loss of the moderator was 1.18% of its initial value. The risk of bundle dryout was, therefore, quite small. (author)

  18. CANDU 3 - Modularization

    International Nuclear Information System (INIS)

    McAskie, M.J.

    1991-01-01

    The CANDU 3 Heavy Water Reactor is the newest design developed by AECL CANDU. It has set as a major objective, the achievement of significant reductions in both cost and schedule over previous designs. The basic construction strategy is to incorporate extensive modularization of the plant in order to parallel the civil and mechanical installation works. This results in a target 38 month construction schedule from first concrete to in-service compared to 68 months for the Wolsong-1 CANDU 6 actually achieved and the 54 months envisaged for an improved CANDU 6. This paper describes the module concepts that have been developed and explains how they contribute to the overall construction program and achieve the desired cost and schedule targets set for the CANDU 3. (author). 7 figs, 2 tabs

  19. Nuclear power plant licensing in Canada

    International Nuclear Information System (INIS)

    Tong, J.S.C.; Waddington, J.G.

    1997-01-01

    The Canadian nuclear power plant licensing practice which has evolved over three decades provides a regulatory framework that promotes safe design and operation of CANDU power plants. From the very outset, it recognizes the need for simple and reliable safety systems which are separate from the systems that are normally used to produce electricity. Further, it requires the reliability of safety systems be demonstrated by routine tests during plant operation. Over the three decades, the analysis requirements to demonstrate the performance and reliability of plant systems that have a role in the detection and mitigating of accidents have also evolved. Today's requirements are defined in consultative documents C-6 and C-98. One recurring theme throughout the evolution of the licensing practice is the maxim of prescribing only basic safety requirements and rules so that designers and operators have the freedom to devise the best possible design features and operating practices

  20. A study to develop the domestic functional requirements of the specific safety systems of CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Man Woong [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Lee, Jae Young; Park, Kun Chul [Handong Global Univ., Pohang (Korea, Republic of)] (and others)

    2003-03-15

    The present research has been made to develop and review critically the functional requirements of the specific safety systems of CANDU such as SDS-1, SDS2, ECCS, and containment. Based on R documents for this, a systematic study was made to develop the domestic regulation statements. Also, the conventional laws are carefully reviewed to see the compatibility to CANDU. Also, the safety assessment method for CANDU was studied by reviewing C documents and recommendation of IAEA. Through the present works, the vague policy in the CANDU safety regulation is cleaning up in a systematic form and a new frame to measure the objective risk of nuclear power plants was developed.

  1. A study to develop the domestic functional requirements of the specific safety systems of CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Man Woong; Lee, Jae Young; Bang, Kwang Hyun [Handong Global Univ., Pohang (Korea, Republic of)] (and others)

    2001-03-15

    The present research has been made to develop and review critically the functional requirements of the specific safety systems of CANDU such as SOS-1, SOS-2, ECCS and containment. Based on R documents for this, a systematic study was made to develop the domestic regulation statements. Also, the conventional laws are carefully reviewed to see the compatibility to CANDU. Also, the safety assessment method for CANDU was studied by reviewing C documents and recommendation of IAEA. Through the present works, the vague policy in the CANDU safety regulation is cleaning up in a systematic form and a new frame to measure the objective risk of nuclear power plants was developed.

  2. Development of a web-based CANDU core management procedures automation system

    International Nuclear Information System (INIS)

    Lee, S.; Park, D.; Yeom, C.; Suh, H.

    2007-01-01

    Introduce CANDU core management procedures automation system (COMPAS) - A web-based application which semi-automates several CANDU core management tasks. It provides various functionalities including selection and evaluation of refueling channel, detector calibration, coolant flow estimation and thermal power calculation through automated interfacing with analysis codes (RFSP, NUCIRC, etc.) and plant data. It also utilizes brand new .NET computing technology such as ASP.NET, smart client, web services and so on. Since almost all functions are abstracted from the previous experiences of the current working members of the Wolsong Nuclear Power Plant (NPP), it will lead to an efficient and safe operation of CANDU plants. (author)

  3. Development of a web-based CANDU core management procedures automation system

    Energy Technology Data Exchange (ETDEWEB)

    Lee, S.; Park, D.; Yeom, C. [Inst. for Advanced Engineering (IAE), Yongin (Korea, Republic of); Suh, H. [Korea Hydro and Nuclear Power (KHNP), Wolsong (Korea, Republic of)

    2007-07-01

    Introduce CANDU core management procedures automation system (COMPAS) - A web-based application which semi-automates several CANDU core management tasks. It provides various functionalities including selection and evaluation of refueling channel, detector calibration, coolant flow estimation and thermal power calculation through automated interfacing with analysis codes (RFSP, NUCIRC, etc.) and plant data. It also utilizes brand new .NET computing technology such as ASP.NET, smart client, web services and so on. Since almost all functions are abstracted from the previous experiences of the current working members of the Wolsong Nuclear Power Plant (NPP), it will lead to an efficient and safe operation of CANDU plants. (author)

  4. Project planning and scheduling techniques for the CANDU programme - an overview

    International Nuclear Information System (INIS)

    Wong, P.T.; Sebastian, P.R.

    1978-01-01

    The energy crisis and higher costs have imposed the need for tighter control of completion time for the construction of CANDU nuclear power plants. System procedures and techniques to meet this challenge are described

  5. Localization of CANDU technology

    Energy Technology Data Exchange (ETDEWEB)

    Alizadeh, Ala

    2010-09-15

    The CANDU pressurized heavy water reactor's principal design features suit it particularly well for technology transfer and localization. When the first commercial CANDU reactors of 540 MWe entered service in 1971, Canada's population of less than 24 million supported a 'medium' level of industrial development, lacking the heavy industrial capabilities of larger countries like the USA, Japan and Europe. A key motivation for Canada in developing the CANDU design was to ensure that Canada would have the autonomous capacity to build and operate nuclear power reactors without depending on foreign sources for key components or enriched fuel.

  6. CANDU in the next century

    International Nuclear Information System (INIS)

    Meneley, D.A.; Torgerson, D.F.

    1997-01-01

    AECL's main product line is available today in two designs, designated as CANDU 6 and CANDU 9. Each of these is based on successfully operating pressurized-heavy-water nuclear plants. Several new CANDU stations are under construction or planned around the world. The author presents plant concepts which may evolve from today's products during the 21st century, indicating the particular development directions which might be followed by AECL product development depending on the future competitive environment, economics, and market circumstances. This study shows that the CANDU energy supply system is sufficiently flexible to be adapted into widely varying circumstances over the next century and beyond

  7. Disturbance analysis in nuclear power plants

    International Nuclear Information System (INIS)

    Sillamaa, M.A.

    Disturbance analysis is any systematic procedure that helps an operator determine what has failed. This paper describes the typical information currently provided in CANDU power plants to help the operator respond to a disturbance. It presents a simplified model of how an operator could get into trouble, and briefly reviews development work on computerized disturbance analysis systems for nuclear power plants being done in various countries including Canada. Disturbance analysis systems promise to be useful tools in helping operators improve their response to complex situations. However, the originality and complexity of the work for a disturbance analysis system and the need to develop operator confidence and management support require a 'walk before you run' approach

  8. The final report of ''on-the-job training'' on the CANDU reactor

    International Nuclear Information System (INIS)

    Kim, D.H.; Koh, B.J.

    1983-01-01

    This is the final Report for the technical ''on-the-job traning'' for the Wolsung CANDU nuclear power plant which is the first Pressurized Heavy Water Reactor setting up in Korea. The technical ''on-the-job traning'' was established to increase the capability for the nuclear safety evaluation in order to contribute the future safe operation of the CANDU nuclear power plant. The training has been excuted through three level courses as elementary, intermediate and ''on-the-job training'' at Wolsung power plant. The elementary course was introduction to the CANDU basics and fundamentals. The intermediate course was the more advanced course, and the detailed concepts and engineering explanations of the CANDU system had been instructed. The third course was the ''on-the-job training'' at the Wolsung plant site, which was the most emphasized course during the project. (Author)

  9. Moderator heat recovery of CANDU reactors

    International Nuclear Information System (INIS)

    Fath, H.E.S.; Ahmed, S.T.

    1986-01-01

    A moderator heat recovery scheme is proposed for CANDU reactors. The proposed circuit utilizes all the moderator heat to the first stages of the plant feedwater heating system. CANDU-600 reactors are considered with moderator heat load varying from 120 to 160 MWsub(th), and moderator outlet temperature (from calandria) varying from 80 to 100 0 C. The steam saved from the turbine extraction system was found to produce an additional electric power ranging from 5 to 11 MW. This additional power represents a 0.7-1.7% increase in the plant electric output power and a 0.2-0.7% increase in the plant thermal efficiency. The outstanding features and advantages of the proposed scheme are presented. (author)

  10. Trends in the capital costs of CANDU generating stations

    International Nuclear Information System (INIS)

    Yu, A.M.

    1982-09-01

    This paper consolidates the actual cost experience gained by Atomic Energy of Canada Limited, Ontario Hydro, and other Canadian electric utlities in the planning, design and construction of CANDU-PHWR (CANada Deuterium Uranium-Pressurized Heavy Water Reactor) generating stations over the past 30 years. For each of the major CANDU-PHWR generating stations in operation and under construction in Canada, an analysis is made to trace the evolution of the capital cost estimates. Major technical, economic and other parameters that affect the cost trends of CANDU-PHWR generating stations are identified and their impacts assessed. An analysis of the real cost of CANDU generating stations is made by eliminating interest during construction and escalation, and the effects of planned deferment of in-service dates. An historical trend in the increase in the real cost of CANDU power plants is established. Based on the cost experience gained in the design and construction of CANDU-PHWR units in Canada, as well as on the assessment of parameters that influence the costs of such projects, the future costs of CANDU-PHWRs are presented

  11. CANDU reactors and greenhouse gas emissions

    International Nuclear Information System (INIS)

    Andseta, S.; Thompson, M.J.; Jarrell, J.P.; Pendergast, D.R.

    1998-01-01

    This paper was originally presented at the 11th Pacific Basin Nuclear Conference, Banff, Alberta, Canada, May 3-7, 1998. It has been updated to include additional lifecycle data on chemical releases from ore treatment and CANDU fuel fabrication. It is sometimes stated that nuclear power plants can supply electricity with zero emissions of greenhouse gases. In fact, consideration of the entire fuel cycle indicates that some greenhouse gases are generated during their construction and decommissioning and by the preparation of fuel and other materials required for their operation. This follows from the use of fossil fuels in the preparation of materials and during the construction and decommissioning of the plants. This paper reviews life cycle studies of several different kinds of power plants. Greenhouse gases generated by fossil fuels during the preparation of fuel and heavy water used by operating CANDU power plants are estimated. The total greenhouse gas emissions from CANDU nuclear plants, per unit of electricity ultimately produced, are very small in comparison with emissions from most other types of power plants. (author)

  12. CANDU reactors and greenhouse gas emissions

    International Nuclear Information System (INIS)

    Andseta, S.; Thompson, M.J.; Jarrell, J.P.; Pendergast, D.R.

    1999-01-01

    This paper was originally presented at the 11th Pacific Basin Nuclear Conference, Banff, Alberta, Canada, May 3-7, 1998. It has been updated to include additional lifecycle data on chemical releases from ore treatment and CANDU fuel fabrication. It is sometimes stated that nuclear power plants can supply electricity with zero emissions of greenhouse gases. In fact, consideration of the entire fuel cycle indicates that some greenhouse gases are generated during their construction and decommissioning and by the preparation of fuel and other materials required for their operation. This follows from the use of fossil fuels in the preparation of materials and during the construction and decommissioning of the plants. This paper reviews life cycle studies of several different kinds of power plants. Greenhouse gases generated by fossil fuels during the preparation of fuel and heavy water used by operating CANDU power plants are estimated. The total greenhouse gas emissions from CANDU nuclear plants, per unit of electricity ultimately produced, are very small in comparison with emissions from most other types of power plants. (author)

  13. CANDU severe accident management guidance update

    International Nuclear Information System (INIS)

    Jones, L.; Popov, N.; Gilbert, L.; Weed, J.

    2014-01-01

    The CANDU Owners Group (COG) developed a set of generic and initial station-specific Severe Accident Management Guidance (SAMG) documents to mitigate the consequences to the public in the event of a severe accident. The generic portion of the COG SAMG was completed in 2006; the overall project including the station-specific phase was completed in April 2007. Over the years, the CANDU industry and utilities have continuously increased the knowledge base for SAMG and have incorporated various engineered features based on the knowledge obtained. As a result of the event that occurred at the Fukushima Daiiachi nuclear power plant (NPP) in Japan, the Canadian Nuclear Safety Commission (CNSC) established the CNSC Fukushima Task Force. The results of the task force were documented in INFO-0828, CNSC Staff Action Plan on the CNSC Fukushima Task Force Recommendations. Among the recommendation documented in INFO-828 were Fukushima Action Items (FAIs) directed towards the CANDU utilities in Canada; a portion of which are related to SAMG documentation updates and directed at enhancing SAM response. A COG joint project was established to support the closure of the CNSC FAIs and to revise the current CANDU documentation accordingly. This paper provides a high level summary of the COG project scope and results. It also demonstrates that the CANDU SAMG programs in Canada provide robust protection and mitigation of severe accidents. (author)

  14. CANDU severe accident management guidance update

    Energy Technology Data Exchange (ETDEWEB)

    Jones, L., E-mail: lisa.m.jones@opg.com [Ontario Power Generation, Pickering, ON (Canada); Popov, N., E-mail: nik.popov@rogers.com [Candu Owners Group, Toronto, ON (Canada); Gilbert, L., E-mail: lovell.gilbert@brucepower.com [Bruce Power, Tiverton, ON (Canada); Weed, J., E-mail: jeff.weed@candu.gov [Candu Owners Group, Toronto, ON (Canada)

    2014-07-01

    The CANDU Owners Group (COG) developed a set of generic and initial station-specific Severe Accident Management Guidance (SAMG) documents to mitigate the consequences to the public in the event of a severe accident. The generic portion of the COG SAMG was completed in 2006; the overall project including the station-specific phase was completed in April 2007. Over the years, the CANDU industry and utilities have continuously increased the knowledge base for SAMG and have incorporated various engineered features based on the knowledge obtained. As a result of the event that occurred at the Fukushima Daiiachi nuclear power plant (NPP) in Japan, the Canadian Nuclear Safety Commission (CNSC) established the CNSC Fukushima Task Force. The results of the task force were documented in INFO-0828, CNSC Staff Action Plan on the CNSC Fukushima Task Force Recommendations. Among the recommendation documented in INFO-828 were Fukushima Action Items (FAIs) directed towards the CANDU utilities in Canada; a portion of which are related to SAMG documentation updates and directed at enhancing SAM response. A COG joint project was established to support the closure of the CNSC FAIs and to revise the current CANDU documentation accordingly. This paper provides a high level summary of the COG project scope and results. It also demonstrates that the CANDU SAMG programs in Canada provide robust protection and mitigation of severe accidents. (author)

  15. Long-term performance of the CANDU-type of vanadium self-powered neutron detectors in NRU

    Energy Technology Data Exchange (ETDEWEB)

    Leung, T.C. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)]. E-mail: leungt@aecl.ca

    2007-07-01

    The CANDU-type of in-core vanadium self-powered neutron flux detectors have been installed in NRU to monitor the axial neutron flux distributions adjacent to the loop fuel test sites since 1996. This paper describes how the thermal neutron fluxes were measured at two monitoring sites, and presents a method of correcting the vanadium burn-up effect, which can be up to 2 to 3% per year, depending on the detector locations in the reactor. It also presents the results of measurements from neutron flux detectors that have operated for over eight-years in NRU. There is good agreement between the measured and simulated neutron fluxes, to within {+-} 6.5%, and the long-term performance of the CANDU-type of vanadium neutron flux detectors in NRU is satisfactory. (author)

  16. Development of distributed computer systems for future nuclear power plants

    International Nuclear Information System (INIS)

    Yan, G.; L'Archeveque, J.V.R.

    1978-01-01

    Dual computers have been used for direct digital control in CANDU power reactors since 1963. However, as reactor plants have grown in size and complexity, some drawbacks to centralized control appear such as, for example, the surprisingly large amount of cabling required for information transmission. Dramatic changes in costs of components and a desire to improve system performance have stimulated a broad-based research and development effort in distribution systems. This paper outlines work in this area

  17. The evolution of the CANDU energy system - ready for Europe's energy future

    International Nuclear Information System (INIS)

    Hedges, K. R.; Hopwood, J. M.

    2001-01-01

    As air quality and climate change issues receive increasing attention, the opportunity for nuclear to play a larger role in the coming decades also increases. The good performance of the current fleet of nuclear plants is crucial evidence of nuclear's potential. The excellent record of Cernavoda-1 is an important part of this, and demonstrates the maturity of the Romanian program and of the CANDU design approach. However, the emerging energy market also presents a stringent economic challenge. Current NPP designs, while established as reliable electricity producers, are seen as limited by high capital costs. In some cases, the response to the economic challenge is to consider radical changes to new design concepts, with attendant development risks from lack of provenness. Because of the flexibility of the CANDU system, it is possible to significantly extend the mid-size CANDU design, creating a Next Generation product, without sacrificing the extensive design, delivery and operations information base for CANDU. This enables a design with superior safety characteristics while at the same time meeting the economic challenge of emerging markets. The Romanian nuclear program has progressed successfully forward, leading to the successful operation of Cernavoda-1, and the project to bring Cernavoda-2 to commercial operation. The Romanian nuclear industry has become a full-fledged member of the CANDU community, with all areas of nuclear technology well established and benefiting from international cooperation with other CANDU organizations. AECL is an active partner with Romanian nuclear organizations, both through cooperative development programs, commercial contracts, and also through the activities of the CANDU owners' Group (COG). The Cernavoda project is part of the CANDU 6 family of nuclear power plants developed by AECL. The modular fuel channel reactor concept can be modified extensively, through a series of incremental changes, to improve economics, safety

  18. Hydroelectric Power Plants Dobsina

    International Nuclear Information System (INIS)

    Majercak, V.; Srenkelova, Z.; Kristak, J.G.

    1997-01-01

    In this brochure the Hydroelectric Power Plants Dobsina, (VED), subsidiary of the utility Slovenske Elektrarne, a.s. (Slovak Electric, plc. Bratislava) are presented. VED is mainly aimed at generating peak-load electrical energy and maintenance of operational equipment. Reaching its goals, company is first of all focused on reliability of production, economy and effectiveness, keeping principles of work safety and industry safety standards and also ecology. VED operates eight hydroelectric power plants, from which PVE Ruzin I and PVE Dobsina I are pump storage ones and they are controlled directly by the Slovak Energy Dispatch Centre located in Zilina thought the system LS 3200. Those power plants participate in secondary regulation of electrical network of Slovakia. They are used to compensate balance in reference to foreign electrical networks and they are put into operation independently from VED. Activity of the branch is focused mainly on support of fulfilment of such an important aim as electric network regulation. Beginnings of the subsidiary Hydroelectric Power Plants Dobsina are related to the year of 1948. After commissioning of the pump storage Hydroelectric Power Plants Dobsina in 1953, the plant started to carry out its mission. Since that time the subsidiary has been enlarged by other seven power plants, through which it is fulfilling its missions nowadays. The characteristics of these hydroelectric power plants (The pump-storage power plant Dobsina, Small hydroelectric power plant Dobsina II, Small hydroelectric power plant Rakovec, Small hydroelectric power plant Svedlar, Hydroelectric power plant Domasa, The pump-storage power plant Ruzin, and Small hydroelectric power plant Krompachy) are described in detail. Employees welfare and public relations are presented

  19. The formation, composition and structure of corrosion products in CANDU nuclear power reactors

    International Nuclear Information System (INIS)

    Rummery, T.E.

    1978-01-01

    To gain a better understanding of the formation and transport of corrosion products in CANDU-PHW power reactors, and the role played by these products in the generation and subsequent fixation of radioactive species, we have examined in detail several surfaces removed from the Douglas Point Generating Station (Douglas Point, Ontario). Results are given for the surface of the primary-side of a Monel-400 boiler tube, and surfaces of carbon steel piping at the inlet and outlet of the boiler. The experimental techniques that were used included sequential acid stripping, X-ray diffractometry, scanning electron microscopy and energy dispersive X-ray spectrometry. The corrosion products on the Monel-400 were mainly nickel, copper, nickel oxide and nickel-deficient nickel ferrite and varied in composition and quantity as a function of both distance from the boiler inlet, and depth in the corrosion layer. The radioactive cobalt ( 60 Co) content was localized in 'streaks' deposited in the straight sections of the boiler tube, but distributed uniformly over the whole surface in the downstream bend section. The material covering the carbon steel surface comprised three phases: magnetite, aluminosilicate particles at the outermost surface, and a mixed cation spinel phase uniformly distributed over the surface at the corrosion film-water interface. The formation, composition and structure of the corrosion products are discussed. (author)

  20. Joint studies on large CANDU

    International Nuclear Information System (INIS)

    Lee, Ikhwan; Yu, S. K. W.

    1994-01-01

    CANDU PHWRs have demonstrated generic benefits which will be continued in future designs. These include economic benefits due to low operating costs, business potential, strategic benefits due to fuel cycle flexibility and operational benefits. These benefits have been realized in Korea through the operation of Wolsong 1, resulting in further construction of PHWRs at the same site. The principal benefit, low electricity cost, is due to the high capacity factor and the low fuel cost for CANDU. The CANDU plant at Wolsong has proven to be a safe, reliable and economical electricity producer. The ability of PHWR to burn natural uranium ensures security of fuel supply. Following successful Technology Transfer via the Wolsong 2,3 and 4 project, future opportunity exists between Korea and Canada for continuing co-operation in research and development to improve the technology base, for product development partnerships, and business opportunities in marketing and building PHWR plants in third countries. High reliability, through excellent design, well-controlled operation, efficient maintenance and low operating costs is critical to the economic viability of nuclear plants. CANDU plants have an excellent performance record. The four operating CANDU 6 plants, operated by four utilities in three countries, are world performance leaders. The CANDU 9 design, with higher output capacity, will help to achieve better site utilization and lower electricity costs. Being an evolutionary design, CANDU 9 assures high performance by utilizing proven systems, and component designs adapted from operating CANDU plants (Bruce B, Darlington and CANDU 6). All system and operating parameters are within the operating proven range of current plants. KAERI and AECL have an agreement to perform joint studies on future PHWR development. The objective of the joint studies is to establish the requirements for the design of future advanced CANDU PHWR including the utility need for design improvements

  1. Operating performance and reliability of CANDU PHWR fuel channels in Canada

    International Nuclear Information System (INIS)

    Strachan, B.; Brown, D.R.

    1983-03-01

    CANDU nuclear plants use many small-diameter high-pressure fuel channels. Good operating performance from the CANDU fuel channels has made a major contribution to the world-leading operating record of the CANDU nuclear power plants. As of 1982 December 31, there were 7,480 fuel channels installed in 18 CANDU reactors over 500 MW(e) in size. Eight of these reactors have been declared in-service and have accumulated 24,000 fuel channel-years of operation. The only significant operating problems with fuel channels have been the occurrence of leaking cracks in 70 fuel channels and a larger amount of axial creep on the early reactors than was originally provided for in the design. Both of these problems have been corrected on all CANDU reactors built since the Bruce GS 'A' station and the newer reactors should exhibit even better performance

  2. MATLAB/SIMULINK platform for simulation of CANDU reactor control system

    International Nuclear Information System (INIS)

    Javidnia, H.; Jiang, J.

    2007-01-01

    In this paper a simulation platform for CANDU reactors' control system is presented. The platform is built on MATLAB/SIMULINK interactive graphical interface. Since MATLAB/SIMULINK are powerful tools to describe systems mathematically, all the subsystems in a CANDU reactor are represented in MATLAB's language and are implemented in SIMULINK graphical representation. The focus of the paper is on the flux control loop of CANDU reactors. However, the ideas can be extended to include other parts in CANDU power plants and the same technique can be applied to other types of nuclear reactors and their control systems. The CANDU reactor model and xenon feedback model are also discussed in this paper. (author)

  3. Object-oriented simulator of the dynamics of Embalse nuclear power plant

    International Nuclear Information System (INIS)

    Boroni, Gustavo A.; Cuadrado, M.; Clausse, Alejandro

    2000-01-01

    LUDWIG is an object-oriented simulator of the dynamics of the CANDU Nuclear power plant Embalse Rio Tercero. The tool consists in a numerical plant analyzer by means of a model of the plant dynamics during normal operation, and a graphic environment for configuration and visualization of results. The simulator was validated against plant transients occurred in the plant and recorded in the past. (author)

  4. Qinshan CANDU NPP outage performance improvement through benchmarking

    International Nuclear Information System (INIS)

    Jiang Fuming

    2005-01-01

    With the increasingly fierce competition in the deregulated Energy Market, the optimization of outage duration has become one of the focal points for the Nuclear Power Plant owners around the world. People are seeking various ways to shorten the outage duration of NPP. Great efforts have been made in the Light Water Reactor (LWR) family with the concept of benchmarking and evaluation, which great reduced the outage duration and improved outage performance. The average capacity factor of LWRs has been greatly improved over the last three decades, which now is close to 90%. CANDU (Pressurized Heavy Water Reactor) stations, with its unique feature of on power refueling, of nuclear fuel remaining in the reactor all through the planned outage, have given raise to more stringent safety requirements during planned outage. In addition, the above feature gives more variations to the critical path of planned outage in different station. In order to benchmarking again the best practices in the CANDU stations, Third Qinshan Nuclear Power Company (TQNPC) have initiated the benchmarking program among the CANDU stations aiming to standardize the outage maintenance windows and optimize the outage duration. The initial benchmarking has resulted the optimization of outage duration in Qinshan CANDU NPP and the formulation of its first long-term outage plan. This paper describes the benchmarking works that have been proven to be useful for optimizing outage duration in Qinshan CANDU NPP, and the vision of further optimize the duration with joint effort from the CANDU community. (authors)

  5. A JAVA applet to simulate a CANDU reactor

    International Nuclear Information System (INIS)

    Varin, E.; Desarmenien, J.

    2004-01-01

    Here we present a CANDU nuclear power plant simulator, directly available on a web page. The developed applet has two mains objectives: to expose the CANDU technology to a large public on the internet; and to construct a realistic simulator to be used as a pedagogical tool for nuclear introduction to high school or under-graduate students. The neutronic behavior and control algorithms of the reactor are simulated. Java programming language enables a very flexible environment for public information and user interaction with the plant. Examples of shutdown and power maneuver are explained. (author)

  6. Incentives for improvement of CANDU

    International Nuclear Information System (INIS)

    Hart, R.S.; Dunn, J.T.; Finlay, R.B.

    1988-12-01

    CANDU is a relatively young technology which has demonstrated many achievements as an electrical power generation system. These achievements include an unsurpassed safety record, high annual and lifetime capacity factors, low electricity cost and a broad range of other performance strengths which together indicate that the CANDU technology is fundamentally sound. Known capabilities not yet fully exploited, such as advanced fuel cycle options, indicate that CANDU technology will continue to pay strong dividends on research, development and design investment. This provides a strong incentive for the improvement of CANDU on a continuing basis

  7. Small hydroelectric power plants

    International Nuclear Information System (INIS)

    Helgesen, Boerre

    2002-01-01

    Small hydroelectric power plants are power plants of 1 - 10 MW. For a supplier, this is an unnatural limit. A more natural limit involves compact engine design and simplified control system. The article discusses most of the engine and electrotechnical aspects in the development, construction and operation of such a plant

  8. The Canadian R and D program targeted at CANDU reactors

    International Nuclear Information System (INIS)

    Moeck, E.O.

    1988-01-01

    CANDU reactors produce electricity cheaply and reliably, with miniscule risk to the population and minimal impact on the environment. About half of Ontario's electricity and a third of New Brunswick's are generated by CANDU power plants. Hydro Quebec and utilities in Argentina, India, Pakistan, and the Republic of Korea also successfully operate CANDU reactors. Romania will soon join their ranks. The proven record of excellent performance of CANDUs is due in part to the first objective of the vigorous R and D program: namely, to sustain and improve existing CANDU power-plant technology. The second objective is to develop improved nuclear power plants that will remain competitive compared with alternative energy supplies. The third objective is to continue to improve our understanding of the processes underlying reactor safety and develop improved technology to mitigate the consequences of upset conditions. These three objectives are addressed by individual R and D programs in the areas of CANDU fuel channels, reduced operating costs, reduced capital costs, reactor safety research, and IAEA safeguards. The work is carried out mainly at three centres of Atomic Energy of Canada Limited--the Chalk River Nuclear Laboratories, the Whiteshell Nuclear Research Establishment, and the Sheridan Park Engineering Laboratories--and at Ontario Hydro's Research Laboratories. Canadian universities, consultants, manufacturers, and suppliers also provide expertise in their areas of specialization

  9. Less power plants

    International Nuclear Information System (INIS)

    TASR

    2003-01-01

    In the Slovak Republic the number of company power plants decreased as against 2001 by two sources. In present time only 35 companies have their own power plants. The companies Slovnaft, Kappa Sturovo, Slovensky hodvab Senica, Matador Puchov, Maytex Liptovsky MikuIas, Kovohuty Krompachy, Chemko Strazske and some Slovak sugar factories belong to the largest company power plants in force of distributing companies. Installed output of present 35 company sources is 531 MW. The largest of separate power plants as Paroplynovy cyklus Bratislava (218 MW) and VD Zilina (72 MW) belong to independent sources. Total installed output of Slovak sources was 8306 MW in the end of last year

  10. Requirements for containment system components in CANDU nuclear power plants

    International Nuclear Information System (INIS)

    1988-02-01

    This Standard specifies the requirements and establishes the rules for design, fabrication, and installation of pressure-retaining containment system components. In this Standard the term 'components' includes non registered items

  11. Requirements for containment system components in CANDU nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1988-02-01

    This Standard specifies the requirements and establishes the rules for design, fabrication, and installation of pressure-retaining containment system components. In this Standard the term `components` includes non registered items.

  12. AECL's advanced CANDU reactor - the ACR

    International Nuclear Information System (INIS)

    Alizadeh, Ala; Allsop, Peter; Hedges, Ken; Hopwood, Jerry; Yu, Stephen

    2003-01-01

    The ACR, the next generation CANDU design, represents the next step in development of the CANDU family of designs. AECL has achieved significant incremental improvements to the mid-size CANDU 6 nuclear power plant through successive projects, both in design and in project delivery. Building on this knowledge base, AECL is continuing to adapt the CANDU design to develop the ACR. This paper summarizes the ACR design features, which include major improvements in economics, inherent safety characteristics, performance and construction methods. Aimed at producing electrical power at a capital cost significantly less than that of the current reactor designs, the ACR is an evolutionary design based on the very successful CANDU 6 reactor. The new ACR product is specifically designed to produce power at a cost competitive with other forms of power generation while achieving short construction times, improved safety, international licensability, high investor returns, and low investor risk. It achieves these targets by taking advantage of the latest advances in both pressure-tube and pressure-vessel reactor technologies and experience. The flexibility and development potential of the fuel channel approach also enables designs to be developed that address priorities identified in international long-term specification programs such as the US Department of Energy (DOE) sponsored Generation IV program and IAEA hosted INPRO program. ACR-700 can be built in 36 months with a 48 month project duration, and deliver a lifetime capacity factor in excess of 90%. Overall, the ACR design represents a balance of proven design basis and innovations to give step improvements in safety, reliability and economics. The ACR development program, now well into the detail design stage, includes parallel formal licensing in the USA and Canada. Based on the status of the ACR design and AECL's on-going experience delivering reactor projects on-time and on-budget, the first ACR could be in service by

  13. The Application of Best Estimate and Uncertainty Analysis Methodology to Large LOCA Power Pulse in a CANDU 6 Reactor

    International Nuclear Information System (INIS)

    Abdul-Razzak, A.; Zhang, J.; Sills, H.E.; Flatt, L.; Jenkins, D.; Wallace, D.J.; Popov, N.

    2002-01-01

    The paper describes briefly a best estimate plus uncertainty analysis (BE+UA) methodology and presents its proto-typing application to the power pulse phase of a limiting large Loss-of-Coolant Accident (LOCA) for a CANDU 6 reactor fuelled with CANFLEX R fuel. The methodology is consistent with and builds on world practice. The analysis is divided into two phases to focus on the dominant parameters for each phase and to allow for the consideration of all identified highly ranked parameters in the statistical analysis and response surface fits for margin parameters. The objective of this analysis is to quantify improvements in predicted safety margins under best estimate conditions. (authors)

  14. Solar thermal power plants

    International Nuclear Information System (INIS)

    Schnatbaum, L.

    2009-01-01

    The solar thermal power plant technology, the opportunities it presents and the developments in the market are outlined. The focus is on the technology of parabolic trough power plants, a proven technology for solar power generation on a large scale. In a parabolic trough power plant, trough-shaped mirrors concentrate the solar irradiation onto a pipe in the focal line of the collector. The thermal energy thus generated is used for electricity generation in a steam turbine. Parabolic trough plants can be combined with thermal storage and fossil or biomass fired heat exchangers to generate electricity even when the sun is not shining. Solar Millennium AG in Erlangen has developed the first power plant of this kind in Europe. After two years of construction the plant started operation in Southern Spain in 2008. This one and its sister projects are important steps leading the way for the whole market. The paper also covers the technological challenges, the key components used and the research and development activities concerning this technology. Solar thermal power plants are ideal for covering peak and medium loads in power grids. In hybrid operation they can also cover base-load. The Solar Chimney power plant, another striking technology for the conversion of solar into electric energy, is described briefly. The paper concludes with a look at the future - the import of solar energy from the deserts of North Africa to central Europe. (author)

  15. Power plant chemical technology

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-12-01

    17 contributions covering topies of fossil fuel combustion, flue gas cleaning, power plant materials, corrosion, water/steam cycle chemistry, monitoring and control were presented at the annual meeting devoted to Power Plant Chemical Technology 1996 at Kolding (Denmark) 4-6 September 1996. (EG)

  16. Nuclear Power Plants. Revised.

    Science.gov (United States)

    Lyerly, Ray L.; Mitchell, Walter, III

    This publication is one of a series of information booklets for the general public published by the United States Atomic Energy Commission. Among the topics discussed are: Why Use Nuclear Power?; From Atoms to Electricity; Reactor Types; Typical Plant Design Features; The Cost of Nuclear Power; Plants in the United States; Developments in Foreign…

  17. NUCLEAR POWER PLANT

    Science.gov (United States)

    Carter, J.C.; Armstrong, R.H.; Janicke, M.J.

    1963-05-14

    A nuclear power plant for use in an airless environment or other environment in which cooling is difficult is described. The power plant includes a boiling mercury reactor, a mercury--vapor turbine in direct cycle therewith, and a radiator for condensing mercury vapor. (AEC)

  18. The Kuroshio power plant

    CERN Document Server

    Chen, Falin

    2013-01-01

    By outlining a new design or the Kuroshio power plant, new approaches to turbine design, anchorage system planning, deep sea marine engineering and power plant operations and maintenance are explored and suggested. The impact on the local environment, particularly in the face of natural disasters, is also considered to provide a well rounded introduction to plan and build a 30MW pilot power plant. Following a literature review, the six chapters of this book propose a conceptual design by focusing on the plant's core technologies and establish the separate analysis logics for turbine design and

  19. Nuclear power plants

    International Nuclear Information System (INIS)

    Margulova, T.Ch.

    1976-01-01

    The textbook focuses on the technology and the operating characteristics of nuclear power plants equiped with pressurized water or boiling water reactors, which are in operation all over the world at present. The following topics are dealt with in relation to the complete plant and to economics: distribution and consumption of electric and thermal energy, types and equipment of nuclear power plants, chemical processes and material balance, economical characteristics concerning heat and energy, regenerative preheating of feed water, degassing and condenser systems, water supply, evaporators, district heating systems, steam generating systems and turbines, coolant loops and pipes, plant siting, ventilation and decontamination systems, reactor operation and management, heat transfer including its calculation, design of reactor buildings, and nuclear power plants with gas or sodium cooled reactors. Numerous technical data of modern Soviet nuclear power plants are included. The book is of interest to graduate and post-graduate students in the field of nuclear engineering as well as to nuclear engineers

  20. CANDU lectures

    International Nuclear Information System (INIS)

    Rouben, B.

    1984-06-01

    This document is a compilation of notes prepared for two lectures given by the author in the winter of 1983 at the Institut de Genie Nucleaire, Ecole Polytechnique, Montreal. The first lecture gives a physical description of the CANDU reactor core: the nuclear lattice, the reactivity mechanisms, their functions and properties. This lecture also covers various aspects of reactor core physics and describes different calculational methods available. The second lecture studies the numerous facets of fuel management in CANDU reactors. The important variables in fuel management, and the rules guiding the refuelling strategy, are presented and illustrated by means of results obtained for the CANDU 600

  1. Results of fuel management at Embalse nuclear power plant. Analysis of performance at other plants

    International Nuclear Information System (INIS)

    Paz, A.O. de; Moreno, C.A.; Vinez, J.C.

    1987-01-01

    The operating experience of fuel management at the Embalse nuclear power plant from new core to the present situation (approximately 937 days at full power) is described. The average core burnup is about 4000 MW d/t U and the monthly averaged discharge burnup about 7800 MW d/t U. The neutron flux distribution is calculated by means of program PUMA-C, which is periodically checked by comparison between calculated and measured values of 102 vanadium detectors. A comparison of the performance of other reactors type CANDU 600 (Point Lepreau, Gentilly 2, Wolsung) from the point of view of fuel strategy is also presented. The data to perform the comparison were obtained by means of the CANDU system of information exchange between users (COG). (Author)

  2. Prediction model for exhausted point of ion exchange resin column of moderator purification circuit at Korean CANDU plant

    International Nuclear Information System (INIS)

    Sohn, Wook; Kang, Duck-Won; Ahn, Hyun Kyoung; Rhee, In Hyoung

    2005-01-01

    Most of the carbon-14 produced at CANDU plants are removed by an Ion eXchange (IX) resin column of the moderator purification circuit, and a column is replaced based on an empirical guideline. Since the amount of carbon-14 released from CANDU plants is governed by the performance of a column, optimal operation of IX resin columns through the timely replacement based on an objective criterion is very important. For this, the model for predicting the exhausted point of an IX resin column has been developed based on local chemical equilibrium. The performance evaluation at Wolsong Unit 3 showed that the model was able to simulate the removal of species by an IX resin column to such a high degree that the model could provide an objective criterion to replace an IX resin column timely. The derived maximum service time of a fresh IX resin column was 4,080 h, about twice that of the existing empirical guideline (up to 2,000h). Accordingly, if the maximum service time derived in this paper is applied to Wolsong Unit 3, it is expected to reduce the cost needed for the replacement of IX resin column by about 50%. (author)

  3. Italian steam power plants

    Energy Technology Data Exchange (ETDEWEB)

    von Rautenkranz, J

    1939-01-01

    A brief history of geothermal power production in Italy is presented. Boric acid has been produced on an industrial scale since 1818. The first electrical power was generated in 1904, and by 1939 the output of geothermal power plants had reached 500 GWh, with major expansion of facilities planned.

  4. Nuclear power plant siting

    International Nuclear Information System (INIS)

    Sulkiewicz, M.; Navratil, J.

    The construction of a nuclear power plant is conditioned on territorial requirements and is accompanied by the disturbance of the environment, land occupation, population migration, the emission of radioactive wastes, thermal pollution, etc. On the other hand, a nuclear power plant makes possible the introduction of district heating and increases the economic and civilization activity of the population. Due to the construction of a nuclear power plant the set limits of negative impacts must not be exceeded. The locality should be selected such as to reduce the unfavourable effects of the plant and to fully use its benefits. The decision on the siting of the nuclear power plant is preceded by the processing of a number of surveys and a wide range of documentation to which the given criteria are strictly applied. (B.H.)

  5. Systems analysis of the CANDU 3 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wolfgong, J.R.; Linn, M.A.; Wright, A.L.; Olszewski, M.; Fontana, M.H. [Oak Ridge National Lab., TN (United States)

    1993-07-01

    This report presents the results of a systems failure analysis study of the CANDU 3 reactor design; the study was performed for the US Nuclear Regulatory Commission. As part of the study a review of the CANDU 3 design documentation was performed, a plant assessment methodology was developed, representative plant initiating events were identified for detailed analysis, and a plant assessment was performed. The results of the plant assessment included classification of the CANDU 3 event sequences that were analyzed, determination of CANDU 3 systems that are ``significant to safety,`` and identification of key operator actions for the analyzed events.

  6. Integrated evolution of the medium power CANDU{sup MD} reactors; Evolution integree des reacteurs CANDU{sup MD} de moyenne puissance

    Energy Technology Data Exchange (ETDEWEB)

    Nuzzo, F. [AECL Accelerators, Kanata, ON (Canada)

    2002-07-01

    The aim of this document is the main improvements of the CANDU reactors in the economic, safety and performance domains. The presentation proposes also other applications as the hydrogen production, the freshening of water sea and the bituminous sands exploitation. (A.L.B.)

  7. CANDU combined cycles featuring gas-turbine engines

    International Nuclear Information System (INIS)

    Vecchiarelli, J.; Choy, E.; Peryoga, Y.; Aryono, N.A.

    1998-01-01

    In the present study, a power-plant analysis is conducted to evaluate the thermodynamic merit of various CANDU combined cycles in which continuously operating gas-turbine engines are employed as a source of class IV power restoration. It is proposed to utilize gas turbines in future CANDU power plants, for sites (such as Indonesia) where natural gas or other combustible fuels are abundant. The primary objective is to eliminate the standby diesel-generators (which serve as a backup supply of class III power) since they are nonproductive and expensive. In the proposed concept, the gas turbines would: (1) normally operate on a continuous basis and (2) serve as a reliable backup supply of class IV power (the Gentilly-2 nuclear power plant uses standby gas turbines for this purpose). The backup class IV power enables the plant to operate in poison-prevent mode until normal class IV power is restored. This feature is particularly beneficial to countries with relatively small and less stable grids. Thermodynamically, the advantage of the proposed concept is twofold. Firstly, the operation of the gas-turbine engines would directly increase the net (electrical) power output and the overall thermal efficiency of a CANDU power plant. Secondly, the hot exhaust gases from the gas turbines could be employed to heat water in the CANDU Balance Of Plant (BOP) and therefore improve the thermodynamic performance of the BOP. This may be accomplished via several different combined-cycle configurations, with no impact on the current CANDU Nuclear Steam Supply System (NSSS) full-power operating conditions when each gas turbine is at maximum power. For instance, the hot exhaust gases may be employed for feedwater preheating and steam reheating and/or superheating; heat exchange could be accomplished in a heat recovery steam generator, as in conventional gas-turbine combined-cycle plants. The commercially available GateCycle power plant analysis program was applied to conduct a

  8. The next generation of CANDU technologies: profiling the potential for hydrogen fuel

    International Nuclear Information System (INIS)

    Hopwood, J.M.

    2001-01-01

    This report discusses the Next-generation CANDU Power Reactor technologies currently under development at AECL. The innovations introduced into proven CANDU technologies include a compact reactor core design, which reduces the size by a factor of one third for the same power output; improved thermal efficiency through higher-pressure steam turbines; reduced use of heavy water (one quarter of the heavy water required for existing plants), thus reducing the cost and eliminating many material handling concerns; use of slightly enriched uranium to extend fuel life to three times that of existing natural uranium fuel and additions to CANDU's inherent passive safety. With these advanced features, the capital cost of constructing the plant can be reduced by up to 40 per cent compared to existing designs. The clean, affordable CANDU-generated electricity can be used to produce hydrogen for fuel cells for the transportation sector, thereby reducing emissions from the transportation sector

  9. Pressure test behaviour of embalse nuclear power plant containment structure

    International Nuclear Information System (INIS)

    Bruschi, S.; Marinelli, C.

    1984-01-01

    It's described the structural behaviour of the containment structure during the pressure test of the Embalse plant (CANDU type, 600MW), made of prestressed concrete with an epoxi liner. Displacement, strain, temperature, and pressure measurements of the containment structure of the Embalse Nuclear Power Plant are presented. The instrumentation set up and measurement specifications are described for all variables of interest before, during and after the pressure test. The analytical models to simulate the heat transfer due to sun heating and air convenction and to predict the associated thermal strains and displacements are presented. (E.G.) [pt

  10. CANDU 9 fuelling machine carriage

    Energy Technology Data Exchange (ETDEWEB)

    Ullrich, D J; Slavik, J F [Atomic Energy of Canada Ltd., Saskatoon, SK (Canada)

    1997-12-31

    Continuous, on-power refuelling is a key feature of all CANDU reactor designs and is essential to maintaining high station capacity factors. The concept of a fuelling machine carriage can be traced to the early CANDU designs, such as the Douglas Point Nuclear Generating Station. In the CANDU 9 480NU unit, the combination of a mobile carriage and a proven fuelling machine head design comprises an effective means of transporting fuel between the reactor and the fuel transfer ports. It is a suitable alternative to the fuelling machine bridge system that has been utilized in the CANDU 6 reactor units. The CANDU 9 480NU fuel handling system successfully combines features that meet the project requirements with respect to fuelling performance, functionality, seismic qualification and the use of proven components. The design incorporates improvements based on experience and applicable current technologies. (author). 4 figs.

  11. CANDU 9 fuelling machine carriage

    International Nuclear Information System (INIS)

    Ullrich, D.J.; Slavik, J.F.

    1996-01-01

    Continuous, on-power refuelling is a key feature of all CANDU reactor designs and is essential to maintaining high station capacity factors. The concept of a fuelling machine carriage can be traced to the early CANDU designs, such as the Douglas Point Nuclear Generating Station. In the CANDU 9 480NU unit, the combination of a mobile carriage and a proven fuelling machine head design comprises an effective means of transporting fuel between the reactor and the fuel transfer ports. It is a suitable alternative to the fuelling machine bridge system that has been utilized in the CANDU 6 reactor units. The CANDU 9 480NU fuel handling system successfully combines features that meet the project requirements with respect to fuelling performance, functionality, seismic qualification and the use of proven components. The design incorporates improvements based on experience and applicable current technologies. (author). 4 figs

  12. Applying operating experience to design the CANDU 3 process

    International Nuclear Information System (INIS)

    Harris, D.S.; Hinchley, E.M.; Pauksens, J.; Snell, V.; Yu, S.K.W.

    1991-01-01

    The CANDU 3 is an advanced, smaller (450 MWe), standardized version of the CANDU now being designed for service later in the decade and beyond. The design of this evolutionary nuclear power plant has been carefully planned and organized to gain maximum benefits from new technologies and from world experience to date in designing, building, commissioning and operating nuclear power stations. The good performance record of existing CANDU reactors makes consideration of operating experience from these plants a particularly vital component of the design process. Since the completion of the first four CANDU 6 stations in the early 1980s, and with the continuing evolution of the multi-unit CANDU station designs since then, AECL CANDU has devised several processes to ensure that such feedback is made available to designers. An important step was made in 1986 when a task force was set up to review and process ideas arising from the commissioning and early operation of the CANDU 6 reactors which were, by that time, operating successfully in Argentina and Korea, as well as the Canadian provinces of Quebec and New Brunswick. The task force issued a comprehensive report which, although aimed at the design of an improved CANDU 6 station, was made available to the CANDU 3 team. By that time also, the Institute of Power Operations (INPO) in the U.S., of which AECL is a Supplier Participant member, was starting to publish Good Practices and Guidelines related to the review and the use of operating experiences. In addition, details of significant events were being made available via the INPO SEE-IN (Significant Event Evaluation and Information Network) Program, and subsequently the CANNET network of the CANDU Owners' Group (COG). Systematic review was thus possible by designers of operations reports, significant event reports, and related documents in a continuing program of design improvement. Another method of incorporating operations feedback is to involve experienced utility

  13. Applying operating experience to design the CANDU 3 process

    Energy Technology Data Exchange (ETDEWEB)

    Harris, D S; Hinchley, E M; Pauksens, J; Snell, V; Yu, S K.W. [AECL-CANDU, Ontario (Canada)

    1991-04-01

    The CANDU 3 is an advanced, smaller (450 MWe), standardized version of the CANDU now being designed for service later in the decade and beyond. The design of this evolutionary nuclear power plant has been carefully planned and organized to gain maximum benefits from new technologies and from world experience to date in designing, building, commissioning and operating nuclear power stations. The good performance record of existing CANDU reactors makes consideration of operating experience from these plants a particularly vital component of the design process. Since the completion of the first four CANDU 6 stations in the early 1980s, and with the continuing evolution of the multi-unit CANDU station designs since then, AECL CANDU has devised several processes to ensure that such feedback is made available to designers. An important step was made in 1986 when a task force was set up to review and process ideas arising from the commissioning and early operation of the CANDU 6 reactors which were, by that time, operating successfully in Argentina and Korea, as well as the Canadian provinces of Quebec and New Brunswick. The task force issued a comprehensive report which, although aimed at the design of an improved CANDU 6 station, was made available to the CANDU 3 team. By that time also, the Institute of Power Operations (INPO) in the U.S., of which AECL is a Supplier Participant member, was starting to publish Good Practices and Guidelines related to the review and the use of operating experiences. In addition, details of significant events were being made available via the INPO SEE-IN (Significant Event Evaluation and Information Network) Program, and subsequently the CANNET network of the CANDU Owners' Group (COG). Systematic review was thus possible by designers of operations reports, significant event reports, and related documents in a continuing program of design improvement. Another method of incorporating operations feedback is to involve experienced utility

  14. Lessons learned from current Qinshan CANDU project and the impact on future NPP's

    International Nuclear Information System (INIS)

    Hedges, K. R.; Didsbury, R.; Yu, S. K. W.

    2000-01-01

    AECL has adopted an evolutionary approach to the development of the CANDU 6 and CANDU 9 Nuclear Power Plant (NPP) designs. Each new NPP project benefits from previous projects and contains an increasing number of fully proven enhancements. In accordance with this evolutionary design approach, AECL has built on the Wolsong and Qinshan successes and the solid performance of the reference CANDU stations to define, review and implement the enhancements for the CANDU 9 NPP. Some of these enhancements include fully integrated project information systems and databases, safety enhancements coming from PSA studies and licensing activities, distributed control systems for plant-wide control and an advanced control center which addresses human factors engineering concepts. Examples of the Qinshan CANDU project delivery enhancements are the utilization of electronic engineering tools for the complete plant, and the linking of these tools with the project material management system and document management systems. The project information is reviewed and approved at the engineering office in Canada and then transmitted to site electronically. Once the electronic data is at site the information packages are extracted as necessary to enable construction and facilitate contract needs with minimum effort. This paper will provide details of the CANDU Qinshan project experiences as well as describing some of the corresponding CANDU 9 enhancements. (author)

  15. Application of classroom simulators in the training of managers at CANDU plants

    Energy Technology Data Exchange (ETDEWEB)

    Bereznai, G.; Rizk, K., E-mail: george.bereznai@uoit.ca, E-mail: khalid.rizk@uoit.ca [Univ. of Ontario Inst. of Technology, Oshawa, ON (Canada)

    2015-07-01

    Technical managers of Canada's nuclear power plants are required to have in-depth knowledge of the normal and abnormal integrated unit operations typical of the plant at which they work. The Advanced Operations Overview for Managers (AOOM) training program was developed by Ontario Power Generation to fulfil this need for many of its managers. The program makes extensive use of 'classroom' simulators that have the same software models as the full-scope training simulators, but use graphical user interface to replicate the control room devices. For the last several years the AOOM program has been delivered by the University of Ontario Institute of Technology. (author)

  16. Application of classroom simulators in the training of managers at CANDU plants

    International Nuclear Information System (INIS)

    Bereznai, G.; Rizk, K.

    2015-01-01

    Technical managers of Canada's nuclear power plants are required to have in-depth knowledge of the normal and abnormal integrated unit operations typical of the plant at which they work. The Advanced Operations Overview for Managers (AOOM) training program was developed by Ontario Power Generation to fulfil this need for many of its managers. The program makes extensive use of 'classroom' simulators that have the same software models as the full-scope training simulators, but use graphical user interface to replicate the control room devices. For the last several years the AOOM program has been delivered by the University of Ontario Institute of Technology. (author)

  17. A compact, low cost, tritium removal plant for CANDU-6 reactors

    International Nuclear Information System (INIS)

    Sood, S.K.; Fong, C.; Kalyanam; Woodall, K.B.

    1997-01-01

    Tritium concentrations in CANDU-6 reactors are currently around 40 Ci/kg in moderator systems and around 1.5 Ci/kg in primary heat transport (PHT) systems. It is expected that tritium concentrations in moderator systems will continue to rise and will reach about 80 Ci/kg at maturity. A more detailed description of the increase in tritium concentrations in the moderator and PHT systems of CANDU-6 reactors is given in the next section of this paper. While moderator systems currently contribute more than 50% to tritium emissions, the impact of acute releases of moderator water is more severe at higher tritium concentrations. This impact can be substantially reduced by the addition of an isotope separation system for lowering the tritium level in the moderator system. In addition, lower tritium levels in CANDU systems will inevitably result in reduced occupational exposures, or will provide economic benefits due to ease of maintenance because less protective measures are required and maintenance activities can be more efficient

  18. Methodology used to calculate moderator-system heat load at full power and during reactor transients in CANDU reactors

    International Nuclear Information System (INIS)

    Aydogdu, K.

    1998-01-01

    Nine components determine the moderator-system heat load during full-power operation and during a reactor power transient in a CANDU reactor. The components that contribute to the total moderator-system heat load at any time consist of the heat generated in the calandria tubes, guide tubes and reactivity mechanisms, moderator and reflector; the heat transferred from calandria shell, the inner tubesheets and the fuel channels; and the heat gained from moderator pumps and heat lost from piping. The contributions from each of these components will vary with time during a reactor transient. The sources of heat that arise from the deposition of nuclear energy can be divided into two categories, viz., a) the neutronic component (which is directly proportional to neutronic power), which includes neutron energy absorption, prompt-fission gamma absorption and capture gamma absorption; and b) the fission-product decay-gamma component, which also varies with time after initiation of the transient. An equation was derived to calculate transient heat loads to the moderator. The equation includes two independent variables that are the neutronic power and fission-product decay-gamma power fractions during the transient and a constant term that represents the heat gained from moderator pumps and heat lost from piping. The calculated heat load in the moderator during steady-state full-power operation for a CANDU 6 reactor was compared with available measurements from the Point Lepreau, Wolsong 1 and Gentilly-2 nuclear generating stations. The calculated and measured values were in reasonably good agreement. (author)

  19. Nuclear power plant construction

    International Nuclear Information System (INIS)

    Lima Moreira, Y.M. de.

    1979-01-01

    The legal aspects of nuclear power plant construction in Brazil, derived from governamental political guidelines, are presented. Their evolution, as a consequence of tecnology development is related. (A.L.S.L.) [pt

  20. Nuclear power plant decommissioning

    International Nuclear Information System (INIS)

    Yaziz Yunus

    1986-01-01

    A number of issues have to be taken into account before the introduction of any nuclear power plant in any country. These issues include reactor safety (site and operational), waste disposal and, lastly, the decommissioning of the reactor inself. Because of the radioactive nature of the components, nuclear power plants require a different approach to decommission compared to other plants. Until recently, issues on reactor safety and waste disposal were the main topics discussed. As for reactor decommissioning, the debates have been academic until now. Although reactors have operated for 25 years, decommissioning of retired reactors has simply not been fully planned. But the Shippingport Atomic Power Plant in Pennysylvania, the first large scale power reactor to be retired, is now being decommissioned. The work has rekindled the debate in the light of reality. Outside the United States, decommissioning is also being confronted on a new plane. (author)

  1. Advanced stellarator power plants

    International Nuclear Information System (INIS)

    Miller, R.L.

    1994-01-01

    The stellarator is a class of helical/toroidal magnetic fusion devices. Recent international progress in stellarator power plant conceptual design is reviewed and comparisons in the areas of physics, engineering, and economics are made with recent tokamak design studies

  2. International power plant business

    Energy Technology Data Exchange (ETDEWEB)

    Grohe, R.

    1986-03-03

    At the Brown Boveri press seminar 'Energy' in Baden-Baden Rainer Grohe, member of the Brown Boveri board, Mannheim, gave a survey of the activities on the international power plant market in recent years. He showed the vacuities which must be taken into account in this sector today. The drastic escalation of demands on power plant suppliers has lead not to a reduction of protagonists but to an increase. (orig.).

  3. Offshore atomic power plants

    International Nuclear Information System (INIS)

    Anon.

    1975-01-01

    Various merits of offshore atomic power plants are illustrated, and their systems are assessed. The planning of the offshore atomic power plants in USA is reviewed, and the construction costs of the offshore plant in Japan were estimated. Air pollution problem may be solved by the offshore atomic power plants remarkably. Deep water at low temperature may be advantageously used as cooling water for condensers. Marine resources may be bred by building artificial habitats and by providing spring-up equipments. In the case of floating plants, the plant design can be standardized so that the construction costs may be reduced. The offshore plants can be classified into three systems, namely artificial island system, floating system and sea bottom-based system. The island system may be realized with the present level of civil engineering, but requires the development of technology for the resistance of base against earthquake and its calculation means. The floating system may be constructed with conventional power plant engineering and shipbuilding engineering, but the aseismatic stability of breakwater may be a problem to be solved. Deep water floating system and deep water submerging system are conceivable, but its realization may be difficult. The sea bottom-based system with large caissons can be realized by the present civil engineering, but the construction of the caissons, stability against earthquake and resistance to waves may be problems to be solved. The technical prediction and assessment of new plant sites for nuclear power plants have been reported by Science and Technology Agency in 1974. The construction costs of an offshore plant has been estimated by the Ministry of International Trade and Industry to be yen71,026/kW as of 1985. (Iwakiri, K.)

  4. Power plants 2010. Lectures

    International Nuclear Information System (INIS)

    2010-01-01

    The proceedings include the following lectures: Facing the challenges - new structures for electricity production. Renewable energies in Europe - chances and challenges. Nuclear outlook in the UK. Sustainable energy for Europe. Requirements of the market and the grid operator at the electricity production companies. Perspectives for the future energy production. Pumped storage plants - status and perspectives. Nuclear power/renewable energies -partners or opponents? New fossil fired power stations in Europe - status and perspectives. Nuclear energy: outlook for new build and lifetime extension in Europe. Biomass in the future European energy market - experiences for dong energy. Meeting the EU 20:20 renewable energy targets: the offshore challenges. DESERTEC: sustainable electricity for Europe, Middle East and North Africa. New power plants in Europe - a challenge for project and quality management. Consideration of safely in new build activities of power plants. Challenges to an integrated development in Maasvlakte, Netherlands. Power enhancement in EnBW power plants. Operational experiences of CCS pilot plants worldwide. Two years of operational experiences with Vattenfall's oxyfuel pilot plant. Pre-conditions for CCS. Storage technologies for a volatile generation. Overview: new generation of gas turbines.

  5. Risk monitor riskangel for risk-informed applications in nuclear power plants

    International Nuclear Information System (INIS)

    Wang, Fang; Wang, Jiaqun; Wang, Jin; Li, Yazhou; Hu, Liqin; Wu, Yican

    2016-01-01

    Highlights: • A general risk monitor riskangel with high-speed cutsets generator engine. • Benchmarks of actual nuclear power plant (NPP) instantaneous risk models. • Applications in daily operation, maintenance plan and component out of service. - Abstract: This paper studied the requirements of risk monitor software and its applications as a plant specific risk monitor, which supports risk-informed configuration risk management for the two CANDU 6 units at the Third Qinshan nuclear power plant (TQNPP) in China. It also describes the regulatory prospective on risk-informed Probabilistic Safety Assessment (PSA) applications and the use of risk monitor at operating nuclear power plants, high level technical and functional requirements for the development of CANDU specific risk monitor software, and future development trends.

  6. Thermal Hydraulic Assessment for Loss of SDCS Event During the Outage of CANDU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jonghyun [Gnest, Inc. Taejon (Korea, Republic of); Lee, Kwangho; Oh, Haechol; Jun, Hwangyong [KEPRI, Taejon (Korea, Republic of)

    2006-07-01

    During the outage(overhaul) of the nuclear power plant, there are several operating states other than the full power state, that is 'Hot-Zero Power', 'Depressurized-Cooldown', and 'Partially Drained'. Until now safety assessment has not been done much for this operating state of CANDU type reactor worldwide. For the accuracy and confidence of PSA for the CANDU outage, the safety analysis is necessary. At the first stage, we analyzed the thermal hydraulic characteristics and safety of the postulated event of loss of shutdown cooling system (SDCS) during the partially drained state which is the longest one in the middle of outage period. As an analysis tool, this study uses the best estimate thermal hydraulic code, RELAP5/CANDU which was modified according to the CANDU specific characteristics and based on RELAP5.Mod3.

  7. Candu reactors with thorium fuel cycles

    International Nuclear Information System (INIS)

    Hopwood, J.M.; Fehrenbach, P.; Duffey, R.; Kuran, S.; Ivanco, M.; Dyck, G.R.; Chan, P.S.W.; Tyagi, A.K.; Mancuso, C.

    2006-01-01

    Over the last decade and a half AECL has established a strong record of delivering CANDU 6 nuclear power plants on time and at budget. Inherently flexible features of the CANDU type reactors, such as on-power fuelling, high neutron economy, fuel channel based heat transport system, simple fuel bundle configuration, two independent shut down systems, a cool moderator and a defence-in-depth based safety philosophy provides an evolutionary path to further improvements in design. The immediate milestone on this path is the Advanced CANDU ReactorTM** (ACRTM**), in the form of the ACR-1000TM**. This effort is being followed by the Super Critical Water Reactor (SCWR) design that will allow water-cooled reactors to attain high efficiencies by increasing the coolant temperature above 550 0 C. Adaptability of the CANDU design to different fuel cycles is another technology advantage that offers an additional avenue for design evolution. Thorium is one of the potential fuels for future reactors due to relative abundance, neutronics advantage as a fertile material in thermal reactors and proliferation resistance. The Thorium fuel cycle is also of interest to China, India, and Turkey due to local abundance that can ensure sustainable energy independence over the long term. AECL has performed an assessment of both CANDU 6 and ACR-1000 designs to identify systems, components, safety features and operational processes that may need to be modified to replace the NU or SEU fuel cycles with one based on Thorium. The paper reviews some of these requirements and the associated practical design solutions. These modifications can either be incorporated into the design prior to construction or, for currently operational reactors, during a refurbishment outage. In parallel with reactor modifications, various Thorium fuel cycles, either based on mixed bundles (homogeneous) or mixed channels (heterogeneous) have been assessed for technical and economic viability. Potential applications of a

  8. Steps to Advanced CANDU 600

    International Nuclear Information System (INIS)

    Oh, Yongshick; Brooks, G. L.

    1988-01-01

    The CANDU nuclear power system was developed from merging of AECL heavy water reactor technology with Ontario Hydro electrical power station expertise. The original four units of Ontario Hydro's Pickering Generating Station are the first full-scale commercial application of the CANDU system. AECL and Ontario Hydro then moved to the next evolutionary step, a more advanced larger scale design for four units at the Bruce Generating Station. CANDU 600 followed as a single unit nuclear electric power station design derived from an amalgam of features of the multiple unit Pickering and Bruce designs. The design of the CANDU 600 nuclear steam supply system is based on the Pickering design with improvements derived from the Bruce design. For example, most CANDU 600 auxiliary systems are based on Bruce systems, whereas the fuel handling system is based on the Pickering system. Four CANDU 600 units are in operation, and five are under construction in Romania. For the additional four units at Pickering Generating Station 'B', Ontario Hydro selected a replica of the Pickering 'A' design with limited design changes to maintain a high level of standardization across all eight units. Ontario Hydro applied a similar policy for the additional four units at Bruce Generating Station 'B'. For the four unit Darlington station, Ontario Hydro selected a design based on Bruce with improvements derived from operating experience, the CANDU 600 design and development programs

  9. Development of Off-take Model, Subcooled Boiling Model, and Radiation Heat Transfer Input Model into the MARS Code for a Regulatory Auditing of CANDU Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, C.; Rhee, B. W.; Chung, B. D. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Ahn, S. H.; Kim, M. W. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2009-05-15

    Korea currently has four operating units of the CANDU-6 type reactor in Wolsong. However, the safety assessment system for CANDU reactors has not been fully established due to a lack of self-reliance technology. Although the CATHENA code had been introduced from AECL, it is undesirable to use a vendor's code for a regulatory auditing analysis. In Korea, the MARS code has been developed for decades and is being considered by KINS as a thermal hydraulic regulatory auditing tool for nuclear power plants. Before this decision, KINS (Korea Institute of Nuclear Safety) had developed the RELAP5/MOD3/CANDU code for CANDU safety analyses by modifying the model of the existing PWR auditing tool, RELAP5/MOD3. The main purpose of this study is to transplant the CANDU models of the RELAP5/MOD3/CANDU code to the MARS code including a quality assurance of the developed models.

  10. Development of Off-take Model, Subcooled Boiling Model, and Radiation Heat Transfer Input Model into the MARS Code for a Regulatory Auditing of CANDU Reactors

    International Nuclear Information System (INIS)

    Yoon, C.; Rhee, B. W.; Chung, B. D.; Ahn, S. H.; Kim, M. W.

    2009-01-01

    Korea currently has four operating units of the CANDU-6 type reactor in Wolsong. However, the safety assessment system for CANDU reactors has not been fully established due to a lack of self-reliance technology. Although the CATHENA code had been introduced from AECL, it is undesirable to use a vendor's code for a regulatory auditing analysis. In Korea, the MARS code has been developed for decades and is being considered by KINS as a thermal hydraulic regulatory auditing tool for nuclear power plants. Before this decision, KINS (Korea Institute of Nuclear Safety) had developed the RELAP5/MOD3/CANDU code for CANDU safety analyses by modifying the model of the existing PWR auditing tool, RELAP5/MOD3. The main purpose of this study is to transplant the CANDU models of the RELAP5/MOD3/CANDU code to the MARS code including a quality assurance of the developed models

  11. Power plant at sea

    International Nuclear Information System (INIS)

    Roggen, M.

    2000-01-01

    Drilling platforms are rather inefficient when it comes to their own power supply. In view of ecotax and their environmental image, the offshore industry particularly the Norwegians is highly committed to changing this situation. An efficient power plant, specially designed for the offshore industry, might just prove to be the answer to their prayers

  12. Nuclear power plants maintenance

    International Nuclear Information System (INIS)

    Anon.

    1988-01-01

    Nuclear power plants maintenance now appears as an important factor contributing to the competitivity of nuclea energy. The articles published in this issue describe the way maintenance has been organized in France and how it led to an actual industrial activity developing and providing products and services. An information note about Georges Besse uranium enrichment plant (Eurodif) recalls that maintenance has become a main data not only for power plants but for all nuclear industry installations. (The second part of this dossier will be published in the next issue: vol. 1 January-February 1989) [fr

  13. Nuclear power plant outages

    International Nuclear Information System (INIS)

    1998-01-01

    The Finnish Radiation and Nuclear Safety Authority (STUK) controls nuclear power plant safety in Finland. In addition to controlling the design, construction and operation of nuclear power plants, STUK also controls refuelling and repair outages at the plants. According to section 9 of the Nuclear Energy Act (990/87), it shall be the licence-holder's obligation to ensure the safety of the use of nuclear energy. Requirements applicable to the licence-holder as regards the assurance of outage safety are presented in this guide. STUK's regulatory control activities pertaining to outages are also described

  14. Nuclear Power Plants (Rev.)

    Energy Technology Data Exchange (ETDEWEB)

    Lyerly, Ray L.; Mitchell III, Walter [Southern Nuclear Engineering, Inc.

    1973-01-01

    Projected energy requirements for the future suggest that we must employ atomic energy to generate electric power or face depletion of our fossil-fuel resources—coal, oil, and gas. In short, both conservation and economic considerations will require us to use nuclear energy to generate the electricity that supports our civilization. Until we reach the time when nuclear power plants are as common as fossil-fueled or hydroelectric plants, many people will wonder how the nuclear plants work, how much they cost, where they are located, and what kinds of reactors they use. The purpose of this booklet is to answer these questions. In doing so, it will consider only central station plants, which are those that provide electric power for established utility systems.

  15. LNG plant combined with power plant

    Energy Technology Data Exchange (ETDEWEB)

    Aoki, I; Kikkawa, Y [Chiyoda Chemical Engineering and Construction Co. Ltd., Tokyo (Japan)

    1997-06-01

    The LNG plant consumers a lot of power of natural gas cooling and liquefaction. In some LNG plant location, a rapid growth of electric power demand is expected due to the modernization of area and/or the country. The electric power demand will have a peak in day time and low consumption in night time, while the power demand of the LNG plant is almost constant due to its nature. Combining the LNG plant with power plant will contribute an improvement the thermal efficiency of the power plant by keeping higher average load of the power plant, which will lead to a reduction of electrical power generation cost. The sweet fuel gas to the power plant can be extracted from the LNG plant, which will be favorable from view point of clean air of the area. (Author). 5 figs.

  16. LNG plant combined with power plant

    International Nuclear Information System (INIS)

    Aoki, I.; Kikkawa, Y.

    1997-01-01

    The LNG plant consumers a lot of power of natural gas cooling and liquefaction. In some LNG plant location, a rapid growth of electric power demand is expected due to the modernization of area and/or the country. The electric power demand will have a peak in day time and low consumption in night time, while the power demand of the LNG plant is almost constant due to its nature. Combining the LNG plant with power plant will contribute an improvement the thermal efficiency of the power plant by keeping higher average load of the power plant, which will lead to a reduction of electrical power generation cost. The sweet fuel gas to the power plant can be extracted from the LNG plant, which will be favorable from view point of clean air of the area. (Author). 5 figs

  17. Study of Containment Vent Strategies During Severe Accident Progression for the CANDU6 Plant

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Youngho; Ahn, K. I. [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    In March, 2011, Fukushima daichi nuclear power plants experienced a long term station blackout. Severe core damage occurred and a large amount of radioactive materials are released outside of the plants. After this terrible accident Nuclear Safety and Security Commission (NSSC) enforced to increase nuclear safety for all operating plants in Korea. To increase plant safety, both hardware reinforcement and software improvement are encouraged. Hardware reinforcement includes the preparation of the external water injection paths to the RCS and the spent fuel pool, a filtered containment venting system (CFVS), and AC power generating truck. Software improvement includes the increase of the effectiveness of the severe accident management guidance (SAMG) and plant staff training. To comply with NSSC's request, Wolsong Unit 1 has fulfilled the hardware reinforcement including the installation of a CFVS and started the extension of a SAMG to the low power and shutdown operation mode. Current SAMG deals accident occurred during full power operation only. The CFVS is designed to open and to close isolation valves manually. It does not require AC power. The operation of the CFVS prevents the reactor containment building failure due to the over-pressurization but it may release radioactive materials out of the reactor containment building. This paper discusses the radiological source terms for the containment vent strategy during severe accident progression which occurred during shutdown operation mode. This work is a part of the development of shutdown SAMG.. The CFVS is an effective means to control the containment pressure when the local air coolers are unavailable. Radioactive materials may release through the CFVS, but their amounts are reduced significantly. The alternative means, i.e., containment vent through the ventilation system which does not have an effective filter, is not a good choice to control the containment condition. It can maintain the containment

  18. CANDU 9 Design improvements based on experience feedback

    International Nuclear Information System (INIS)

    Yu, S. K. W.; Bonechi, M.; Snell, V. G.

    2000-01-01

    An evolutionary approach utilizing advance technologies has been implenented for the enhancement introduced in the CANDU 9 Nuclear Power Plant (NPP) design. The design of these systems and associated equipment has also benfited from experience feedback from operating CANDU stations and from including advanced products from CANDU engineering and research programs. This paper highlights the design features that contribute to the safety improvements of the CANDU 9 design, summarizes the analysis results which demonstrate the improved performance and also emphasizes design features which reduce operation and maintenance (Q and M) costs. The safety design features highlighted include the increased use of passive devices and heat sinks to achieve extensive system simplification; this also improves reliability and reduces maintenance workloads. System features that contribute to improved operability are also described. The CANDU 9 Control Center provides plant staff with enhanced operating, maintenance and diagnostics features which significantly improve operability, testing and maintainability due to the integration of human factors engineering with a systematic design process. (author)

  19. Application of fuel management calculation codes for CANDU reactor

    International Nuclear Information System (INIS)

    Ju Haitao; Wu Hongchun

    2003-01-01

    Qinshan Phase III Nuclear Power Plant adopts CANDU-6 reactors. It is the first time for China to introduce this heavy water pressure tube reactor. In order to meet the demands of the fuel management calculation, DRAGON/DONJON code is developed in this paper. Some initial fuel management calculations about CANDU-6 reactor of Qinshan Phase III are carried out using DRAGON/DONJON code. The results indicate that DRAGON/DONJON can be used for the fuel management calculation for Qinshan Phase III

  20. Atomic Energy Control Board (AECB) staff assessment and views of current maintenance practices of a four unit CANDU plant

    International Nuclear Information System (INIS)

    Malek, I.

    1995-01-01

    This paper discusses the AECB practices in assessing maintenance activities at one four unit CANDU nuclear plant.-it outlines the authority of the AECB in enforcing the licence condition concerned with maintenance, and how this is interpreted by AECB site staff to measure and report maintenance activities. The AECB staff attaches great importance to proper maintenance as it affects safe operation. Programs used by the licensee staff to identify safety important components, or to predict degradations and failures are of particular interest. In our experience, the application of such programs has been generally good. However, their integration into an overall maintenance scheme can be improved, and the possibilities of integration are not well understood. This paper includes examples of such integration to illustrate our views and to highlight the resultant benefits that AECB staff believes are possible. (author)

  1. Evolution of CANDU reactor design

    International Nuclear Information System (INIS)

    Pon, G.A.

    1978-08-01

    The CANDU (CANada Deuterium Uranium) design had its begin-ings in the early 1950's with the preliminary engineering studies that led to the 20 MW(e) NPD (Nuclear Power Demonstration) and the 200 MW(e) Douglas Point station . The next decade saw the first operation of both these stations and the commitment of the 2000 MW(e) Pickering and 3000 MW(e) Bruce plants. The present decade has witnessed the excellent performance of Pickering and Bruce and commitments to construct Gentilly-2, Cordoba, Pt. Lepreau, Wolsung, Pickering B, Bruce B and Darlington. In most cases, successive CANDU designs have meant an increase in plant output. Evolutionary developments have been made to fit the requirements of higher ratings and sizes, new regulations, better reliability and maintainability and lower costs. These changes, which are described system by system, have been introduced in the course of engineering parallel reactor projects with overlapping construction schedules -circumstances which ensure close contact with the practical realities of economics, manufacturing functions, construction activities and performance in commissioning. Features for one project furnished alternative concepts for others still on the drawing board and the experience gained in the first application yielded a sound basis for its re-use in succeeding projects. Thus the experiences gained in NPD, Douglas Point, Gentilly-1 and KANUPP have contributed to Pickering and Bruce, which in turn have contributed to the design of Gentilly-2. (author)

  2. The European source-term evaluation code ASTEC: status and applications, including CANDU plant applications

    International Nuclear Information System (INIS)

    Van Dorsselaere, J.P.; Giordano, P.; Kissane, M.P.; Montanelli, T.; Schwinges, B.; Ganju, S.; Dickson, L.

    2004-01-01

    Research on light-water reactor severe accidents (SA) is still required in a limited number of areas in order to confirm accident-management plans. Thus, 49 European organizations have linked their SA research in a durable way through SARNET (Severe Accident Research and management NETwork), part of the European 6th Framework Programme. One goal of SARNET is to consolidate the integral code ASTEC (Accident Source Term Evaluation Code, developed by IRSN and GRS) as the European reference tool for safety studies; SARNET efforts include extending the application scope to reactor types other than PWR (including VVER) such as BWR and CANDU. ASTEC is used in IRSN's Probabilistic Safety Analysis level 2 of 900 MWe French PWRs. An earlier version of ASTEC's SOPHAEROS module, including improvements by AECL, is being validated as the Canadian Industry Standard Toolset code for FP-transport analysis in the CANDU Heat Transport System. Work with ASTEC has also been performed by Bhabha Atomic Research Centre, Mumbai, on IPHWR containment thermal hydraulics. (author)

  3. Wind power plant

    Energy Technology Data Exchange (ETDEWEB)

    Caneghem, A.E. von

    1975-07-24

    The invention applies to a wind power plant in which the wind is used to drive windmills. The plant consists basically of a vertical tube with a lateral wind entrance opening with windmill on its lower end. On its upper end, the tube carries a nozzle-like top which increases the wind entering the tube by pressure decrease. The wind is thus made suitable for higher outputs. The invention is illustrated by constructional examples.

  4. CANDU passive shutdown systems

    Energy Technology Data Exchange (ETDEWEB)

    Hart, R S; Olmstead, R A [AECL CANDU, Sheridan Park Research Community, Mississauga, ON (Canada)

    1996-12-01

    CANDU incorporates two diverse, passive shutdown systems, independent of each other and from the reactor regulating system. Both shutdown systems function in the low pressure, low temperature, moderator which surrounds the fuel channels. The shutdown systems are functionally different, physically separate, and passive since the driving force for SDS1 is gravity and the driving force for SDS2 is stored energy. The physics of the reactor core itself ensures a degree of passive safety in that the relatively long prompt neutron generation time inherent in the design of CANDU reactors tend to retard power excursions and reduces the speed required for shutdown action, even for large postulated reactivity increases. All passive systems include a number of active components or initiators. Hence, an important aspect of passive systems is the inclusion of fail safe (activated by active component failure) operation. The mechanisms that achieve the fail safe action should be passive. Consequently the passive performance of the CANDU shutdown systems extends beyond their basic modes of operation to include fail safe operation based on natural phenomenon or stored energy. For example, loss of power to the SDS1 clutches results in the drop of the shutdown rods by gravity, loss of power or instrument air to the injection valves of SDS2 results in valve opening via spring action, and rigorous self checking of logic, data and timing by the shutdown systems computers assures a fail safe reactor trip through the collapse of a fluctuating magnetic field or the discharge of a capacitor. Event statistics from operating CANDU stations indicate a significant decrease in protection system faults that could lead to loss of production and elimination of protection system faults that could lead to loss of protection. This paper provides a comprehensive description of the passive shutdown systems employed by CANDU. (author). 4 figs, 3 tabs.

  5. Benchmarking Nuclear Power Plants

    International Nuclear Information System (INIS)

    Jakic, I.

    2016-01-01

    One of the main tasks an owner have is to keep its business competitive on the market while delivering its product. Being owner of nuclear power plant bear the same (or even more complex and stern) responsibility due to safety risks and costs. In the past, nuclear power plant managements could (partly) ignore profit or it was simply expected and to some degree assured through the various regulatory processes governing electricity rate design. It is obvious now that, with the deregulation, utility privatization and competitive electricity market, key measure of success used at nuclear power plants must include traditional metrics of successful business (return on investment, earnings and revenue generation) as well as those of plant performance, safety and reliability. In order to analyze business performance of (specific) nuclear power plant, benchmarking, as one of the well-established concept and usual method was used. Domain was conservatively designed, with well-adjusted framework, but results have still limited application due to many differences, gaps and uncertainties. (author).

  6. Wind power plant

    Energy Technology Data Exchange (ETDEWEB)

    Kling, A

    1977-01-13

    The wind power plant described has at least one rotor which is coupled to an electricity generator. The systems are fixed to a suspended body so that it is possible to set up the wind power plant at greater height where one can expect stronger and more uniform winds. The anchoring on the ground or on a floating body is done by mooring cables which can simultaneously have the function of an electric cable. The whole system can be steered by fins. The rotor system itself consists of at least one pair of contrarotating, momentum balanced rotors.

  7. Ardennes nuclear power plant

    International Nuclear Information System (INIS)

    1974-12-01

    The SENA nuclear power plant continued to operate, as before, at authorized rated power, namely 905MWth during the first half year and 950MWth during the second half year. Net energy production:2028GWh; hours phased to the line: 7534H; availability factor: 84%; utilization factor: 84%; total shutdowns:19; number of scrams:10; cost per KWh: 4,35 French centimes. Overall, the plant is performing very satisfactory. Over the last three years net production has been 5900GWh, corresponding to in average utilization factor of 83%

  8. Nuclear power plant

    International Nuclear Information System (INIS)

    Orlov, V.V.; Rineisky, A.A.

    1975-01-01

    The invention is aimed at designing a nuclear power plant with a heat transfer system which permits an accelerated fuel regeneration maintaining relatively high initial steam values and efficiency of the steam power circuit. In case of a plant with three circuits the secondary cooling circuit includes a steam generator with preheater, evaporator, steam superheater and intermediate steam superheater. At the heat supply side the latter is connected with its inlet to the outlet of the evaporator and with its outlet to the low-temperature side of the secondary circuit

  9. TAPCHAN Wave Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    1983-10-01

    The Tapered Channel Wave Power Plant (TAPCHAN) is based on a new method for wave energy conversion. The principle of operation can be explained by dividing the system into the following four sub-systems: Firstly, a collector which is designed to concentrate the water energy and optimize collection efficiency for a range of frequencies and directions. Secondly, the energy converter, in which the energy of the collected waves is transformed into potential energy in an on-shore water reservoir. This is the unique part of the power plant. It consists of a gradually narrowing channel with wall heights equal to the filling level of the reservoir (typical heights 3-7 m). The waves enter the wide end of the channel and as they propagate down the narrowing channel, the wave height is amplified until the wavecrests spill over the walls. Thirdly, a reservoir which provides a stable water supply for the turbines. Finally, the hydroelectric power plant, where well established techniques are used for the generation of electric power. The water turbine driving the electric generator is of a low head type, such as a Kaplan or a tubular turbine. It must be designed for salt water operation and should have good regulation capabilities. Power plants based on the principle described, are now offered on a commercial basis.

  10. Performance of candu-6 fuel bundles manufactured in romania nuclear fuel plant

    International Nuclear Information System (INIS)

    Bailescu, A.; Barbu, A.; Din, F.; Dinuta, G.; Dumitru, I.; Musetoiu, A.; Serban, G.; Tomescu, A.

    2013-01-01

    The purpose of this article is to present the performance of nuclear fuel produced by Nuclear Fuel Plant (N.F.P.) - Pitesti during 1995 - 2012 and irradiated in units U1 and U2 from Nuclear Power Plant (N.P.P.) Cernavoda and also present the Nuclear Fuel Plant (N.F.P.) - Pitesti concern for providing technology to prevent the failure causes of fuel bundles in the reactor. This article presents Nuclear Fuel Plant (N.F.P.) - Pitesti experience on tracking performance of nuclear fuel in reactor and strategy investigation of fuel bundles notified as suspicious and / or defectives both as fuel element and fuel bundle, it analyzes the possible defects that can occur at fuel bundle or fuel element and can lead to their failure in the reactor. Implementation of modern technologies has enabled optimization of manufacturing processes and hence better quality stability of achieving components (end caps, chamfered sheath), better verification of end cap - sheath welding. These technologies were qualified by Nuclear Fuel Plant (N.F.P.) - Pitesti on automatic and Computer Numerical Control (C.N.C.) programming machines. A post-irradiation conclusive analysis which will take place later this year (2013) in Institute for Nuclear Research Pitesti (the action was initiated earlier this year by bringing a fuel bundle which has been reported defective by pool visual inspection) will provide additional information concerning potential damage causes of fuel bundles due to manufacturing processes. (authors)

  11. CANDU fuel performance

    International Nuclear Information System (INIS)

    Ivanoff, N.V.; Bazeley, E.G.; Hastings, I.J.

    1982-01-01

    CANDU fuel has operated successfully in Ontario Hydro's power reactors since 1962. In the 19 years of experience, about 99.9% of all fuel bundles have performed as designed. Most defects occurred before 1979 and subsequent changes in fuel design, fuel management, reactor control, and manufacturing quality control have reduced the current defect rate to near zero. Loss of power production due to defective fuel has been negligible. The outstanding performance continues while maintaining a low unit energy cost for fuel

  12. Cost and schedule reduction for next-generation Candu

    International Nuclear Information System (INIS)

    Hopwood, J.M.; Yu, S.; Pakan, M.; Soulard, M.

    2002-01-01

    AECL has developed a suite of technologies for Candu R reactors that enable the next step in the evolution of the Candu family of heavy-water-moderated fuel-channel reactors. These technologies have been combined in the design for the Advanced Candu Reactor TM1 (ACRTM), AECL's next generation Candu power plant. The ACR design builds extensively on the existing Candu experience base, but includes innovations, in design and in delivery technology, that provide very substantial reductions in capital cost and in project schedules. In this paper, main features of next generation design and delivery are summarized, to provide the background basis for the cost and schedule reductions that have been achieved. In particular the paper outlines the impact of the innovative design steps for ACR: - Selection of slightly enriched fuel bundle design; - Use of light water coolant in place of traditional Candu heavy water coolant; - Compact core design with unique reactor physics benefits; - Optimized coolant and turbine system conditions. In addition to the direct cost benefits arising from efficiency improvement, and from the reduction in heavy water, the next generation Candu configuration results in numerous additional indirect cost benefits, including: - Reduction in number and complexity of reactivity mechanisms; - Reduction in number of heavy water auxiliary systems; - Simplification in heat transport and its support systems; - Simplified human-machine interface. The paper also describes the ACR approach to design for constructability. The application of module assembly and open-top construction techniques, based on Candu and other worldwide experience, has been proven to generate savings in both schedule durations and overall project cost, by reducing premium on-site activities, and by improving efficiency of system and subsystem assembly. AECL's up-to-date experience in the use of 3-D CADDS and related engineering tools has also been proven to reduce both engineering and

  13. Geothermal Power Generation Plant

    Energy Technology Data Exchange (ETDEWEB)

    Boyd, Tonya [Oregon Inst. of Technology, Klamath Falls, OR (United States). Geo-Heat Center

    2013-12-01

    Oregon Institute of Technology (OIT) drilled a deep geothermal well on campus (to 5,300 feet deep) which produced 196°F resource as part of the 2008 OIT Congressionally Directed Project. OIT will construct a geothermal power plant (estimated at 1.75 MWe gross output). The plant would provide 50 to 75 percent of the electricity demand on campus. Technical support for construction and operations will be provided by OIT’s Geo-Heat Center. The power plant will be housed adjacent to the existing heat exchange building on the south east corner of campus near the existing geothermal production wells used for heating campus. Cooling water will be supplied from the nearby cold water wells to a cooling tower or air cooling may be used, depending upon the type of plant selected. Using the flow obtained from the deep well, not only can energy be generated from the power plant, but the “waste” water will also be used to supplement space heating on campus. A pipeline will be construction from the well to the heat exchanger building, and then a discharge line will be construction around the east and north side of campus for anticipated use of the “waste” water by facilities in an adjacent sustainable energy park. An injection well will need to be drilled to handle the flow, as the campus existing injection wells are limited in capacity.

  14. Nuclear power plant analyzer

    International Nuclear Information System (INIS)

    Stritar, A.

    1986-01-01

    The development of Nuclear Power Plant Analyzers in USA is described. There are two different types of Analyzers under development in USA, the forst in Idaho and Los Alamos national Lab, the second in brookhaven National lab. That one is described in detail. The computer hardware and the mathematical models of the reactor vessel thermalhydraulics are described. (author)

  15. Nuclear power plant

    International Nuclear Information System (INIS)

    Wieser, R.

    1979-01-01

    The reactor pressure vessel consists of two parts. A cylindrical lower part with a hemispherical steel roof is placed at some distance within an equally shaped pressure vessel of concrete. Both vessels are standing on a common bottom plate. The interspace is kept at subpressure. It serves to contain ring galleries, elevator shafts, and power plant components. (GL) [de

  16. Nuclear Power Plant Technician

    Science.gov (United States)

    Randall, George A.

    1975-01-01

    The author recognizes a body of basic knowledge in nuclear power plant technoogy that can be taught in school programs, and lists the various courses, aiming to fill the anticipated need for nuclear-trained manpower--persons holding an associate degree in engineering technology. (Author/BP)

  17. Steam power plant

    International Nuclear Information System (INIS)

    Campbell, J.W.E.

    1981-01-01

    This invention relates to power plant forced flow boilers operating with water letdown. The letdown water is arranged to deliver heat to partly expanded steam passing through a steam reheater connected between two stages of the prime mover. (U.K.)

  18. Power plants 2009. Lectures

    International Nuclear Information System (INIS)

    2009-01-01

    Within the Annual Conference 2009 of the VGB PowerTech e.V. (Essen, Federal Republic of Germany) from 23rd to 25th May, 2009, in Lyon (France) the following lectures were held: (1) Electricity demand, consequences of the financial and economic crisis - Current overview 2020 for the EU-27 (Hans ten Berge); (2) Status and perspectives of the electricity generation mix in France (Bernard Dupraz); (3) European electricity grid - status and perspective (Dominique Maillard); (4) Technologies and acceptance in the European energy market (Gordon MacKerran); (5) EPR construction in Finland, China, France, (Claude Jaouen); (6) EPR Flamanville 3: A project on the path towards nuclear revival (Jacques Alary); (7) Worldwide nuclear Revival and acceptance (Luc Geraets); (8) An overview on the status of final disposal of radioactive wastes worldwide (Piet Zuidema); (9) Who needs pumped storage plants? PSP are partner to grid stability and renewable energies (Hans-Christoph Funke); (10) Sustainable use of water resources to generate electricity safely and efficiently (Patrick Tourasse); (11) The growth strategy of RWE Innogy - Role of RES in RWE strategy (Fritz Vahrenholt); (12) Solar technologies towards grid parity - key factors and timeframe (G. Gigliucci); (13) Overview on CCS technologies and results of Vattenfalls oxyfuel pilot plant (Philippe Paelinck); (14) Development perspectives of lignite-based IGCC-plants with CCS (Dietmar Keller); (15) Post combustion capture plants - concept and plant integration (Wolfgang Schreier); (16) CCS fossil power generation in a carbon constraint world (Daniel Hofmann); (17) CEZ group strategy in Central and South Eastern Europe (Jan Zizka); (18) Strategy and projects of DONG Energy (Jens Erik Pedersen); (19) E.ON coal-based power generation of the future - The highly efficient power plant and downstream separation of carbon dioxide (Gerhard Seibel); (20) Final sage of first supercritical 460 MW e l. CFB Boiler construction - firs

  19. Review on the application of system engineer model in nuclear power plant

    International Nuclear Information System (INIS)

    Chen Guocai

    2005-01-01

    system engineer was adopted deeply and play important roles in nuclear power plants in United States and Canada, the plant performance indicates that system engineer mode is a good practice. Qinshan CANDU nuclear power plant, established the system engineer mode since commissioning, as a core, system engineer took charge of the preparation of commissioning procedures, organization, coordination and guidance of commissioning execution. Unit 1 was put into commercial operation 43 days in advance and 112 days ahead of schedule for Unit 2 with excellent quality. Commissioning period are just 10.5 and 7.8 months for both Units respectively. Which is the shortest record in the history of CANDU nuclear power plant commissioning up to now. During operation, systems engineer has strength in routine operating and units reliability improvement. Based on the practice of Qinshan CANDU nuclear power plant commissioning and production technical management, the main form of the article in the era of knowledge: its characteristics and advantage and operating mode of the system engineer mode. System engineer is different from project engineer, he act as the master of systems and takes full responsibility for systems technical management. System engineer should do many jobs and improvement schedule to ensure his system in health status. System health monitor is a basic tool in system management, which is useful for equipment performance improvement. At last, the author made a forecast and comment on the prospects for the system engineer in the future. (author)

  20. Development of Operational Safety Monitoring System and Emergency Preparedness Advisory System for CANDU Reactors (I)

    International Nuclear Information System (INIS)

    Kim, Ma Woong; Shin, Hyeong Ki; Lee, Sang Kyu; Kim, Hyun Koon; Yoo, Kun Joong; Ryu, Yong Ho; Son, Han Seong; Song, Deok Yong

    2007-01-01

    As increase of operating nuclear power plants, an accident monitoring system is essential to ensure the operational safety of nuclear power plant. Thus, KINS has developed the Computerized Advisory System for a Radiological Emergency (CARE) system to monitor the operating status of nuclear power plant continuously. However, during the accidents or/and incidents some parameters could not be provided from the process computer of nuclear power plant to the CARE system due to limitation of To enhance the CARE system more effective for CANDU reactors, there is a need to provide complement the feature of the CARE in such a way to providing the operating parameters using to using safety analysis tool such as CANDU Integrated Safety Analysis System (CISAS) for CANDU reactors. In this study, to enhance the safety monitoring measurement two computerized systems such as a CANDU Operational Safety Monitoring System (COSMOS) and prototype of CANDU Emergency Preparedness Advisory System (CEPAS) are developed. This study introduces the two integrated safety monitoring system using the R and D products of the national mid- and long-term R and D such as CISAS and ISSAC code

  1. Beloyarsk Nuclear Power Plant

    International Nuclear Information System (INIS)

    1997-01-01

    The Beloyarsk Nuclear Power Plant (BNPP) is located in Zarechny, approximately 60 km east of Ekaterinberg along the Trans-Siberian Highway. Zarechny, a small city of approximately 30,000 residents, was built to support BNPP operations. It is a closed city to unescorted visitors. Residents must show identification for entry. BNPP is one of the first and oldest commercial nuclear power plants in Russia and began operations in 1964. As for most nuclear power plants in the Russian Federation, BNPP is operated by Rosenergoatom, which is subordinated to the Ministry of Atomic Energy of the Russian Federation (Minatom). BNPP is the site of three nuclear reactors, Units 1, 2, and 3. Units 1 and 2, which have been shut-down and defueled, were graphite moderated reactors. The units were shut-down in 1981 and 1989. Unit 3, a BN-600 reactor, is a 600 MW(electric) sodium-cooled fast breeder reactor. Unit 3 went on-line in April 1980 and produces electric power which is fed into a distribution grid and thermal power which provides heat to Zarechny. The paper also discusses the SF NIKIET, the Sverdiovsk Branch of NIKIET, Moscow, which is the research and development branch of the parent NIKEIT and is primarily a design institute responsible for reactor design. Central to its operations is a 15 megawatt IVV research reactor. The paper discusses general security and fissile material control and accountability at these two facilities

  2. Nuclear power plant safety

    International Nuclear Information System (INIS)

    Otway, H.J.

    1974-01-01

    Action at the international level will assume greater importance as the number of nuclear power plants increases, especially in the more densely populated parts of the world. Predictions of growth made prior to October 1973 [9] indicated that, by 1980, 14% of the electricity would be supplied by nuclear plants and by the year 2000 this figure would be about 50%. This will make the topic of international co-operation and standards of even greater importance. The IAEA has long been active in providing assistance to Member States in the siting design and operation of nuclear reactors. These activities have been pursued through advisory missions, the publication of codes of practice, guide books, technical reports and in arranging meetings to promote information exchange. During the early development of nuclear power, there was no well-established body of experience which would allow formulation of internationally acceptable safety criteria, except in a few special cases. Hence, nuclear power plant safety and reliability matters often received an ad hoc approach which necessarily entailed a lack of consistency in the criteria used and in the levels of safety required. It is clear that the continuation of an ad hoc approach to safety will prove inadequate in the context of a world-wide nuclear power industry, and the international trade which this implies. As in several other fields, the establishment of internationally acceptable safety standards and appropriate guides for use by regulatory bodies, utilities, designers and constructors, is becoming a necessity. The IAEA is presently planning the development of a comprehensive set of basic requirements for nuclear power plant safety, and the associated reliability requirements, which would be internationally acceptable, and could serve as a standard frame of reference for nuclear plant safety and reliability analyses

  3. Future generations of CANDU: advantages and development with passive safety

    International Nuclear Information System (INIS)

    Duffey, R. B.

    2006-01-01

    Atomic Energy of Canada Limited (AECL) advances water reactor and CANDLT technology using an evolutionary development strategy. This strategy ensures that innovations are based firmly on current experience and keeps our development programs focused on one reactor concept, reducing risks, development costs, and product development cycle times. It also assures our customers that our products will never become obsolete or unsupported, and the continuous line of water reactor development is secure and supported into the future. Using the channel reactor advantage of modularity, the subdivided core has the advantage of passive safety by heat removal to the low- pressure moderator. With continuous improvements, the Advanced CANDU Reactor TM (ACR-1000TM) concept will likely remain highly competitive for a number of years and leads naturally to the next phase of CANDU development, namely the Generation IV CANDU -SCWR concept. This is conventional water technology, since supercritical boilers and turbines have been operating for some time in coal-fired power plants. Significant cost, safety, and performance advantages would result from the CANDU-SCWR concept, plus the flexibility of a range of plant sizes suitable for both small and large electric grids, and the ability for co-generation of electric power, process heat, and hydrogen. In CANDU-SCWR, novel developments are included in the primary circuit layout and channel design. The R and D in Canada is integrated with the Generation IV international Forum (GIF) plans, and has started on examining replaceable insulating liners that would ensure channel life, and on providing completely passive reactor decay heat removal directly to the moderator heat sink without forced cooling. In the interests of sustainability, hydrogen production by a CANDU- SCWR is also be included as part of the system requirements, where the methods for hydrogen production will depend on the outlet temperature of the reactor

  4. Advancing the CANDU reactor: From generation to generation

    International Nuclear Information System (INIS)

    Hopwood, Jerry; Duffey, Romney B.; Yu, Steven; Torgerson, Dave F.

    2006-01-01

    Emphasizing safety, reliability and economics, the CANDU reactor development strategy is one of continuous improvement, offering value and assured support to customers worldwide. The Advanced CANDU Reactor (ACR-1000) generation, designed by Atomic Energy of Canada Limited (AECL), meets the new economic expectation for low-cost power generation with high capacity factors. The ACR is designed to meet customer needs for reduced capital cost, shorter construction schedule, high plant capacity factor, low operating cost, increased operating life, simple component replacement, enhanced safety features, and low environmental impact. The ACR-1000 design evolved from the internationally successful medium-sized pressure tube reactor (PTR) CANDU 6 and incorporates operational feedback from eight utilities that operate 31 CANDU units. This technical paper provides a brief description of the main features of the ACR-1000, and its major role in the development path of the generations of the pressure tube reactor concept. The motivation, philosophy and design approach being taken for future generation of CANDU pressure tube reactors are described

  5. Passive heat removal in CANDU

    International Nuclear Information System (INIS)

    Hart, R.S.

    1997-01-01

    CANDU has a tradition of incorporating passive systems and passive components whenever they are shown to offer performance that is equal to or better than that of active systems, and to be economic. Examples include the two independent shutdown systems that employ gravity and stored energy respectively, the dousing subsystem of the CANDU 6 containment system, and the ability of the moderator to cool the fuel in the event that all coolant is lost from the fuel channels. CANDU 9 continues this tradition, incorporating a reserve water system (RWS) that increases the inventory of water in the reactor building and profiles a passive source of makeup water and/or heat sinks to various key process systems. The key component of the CANDU 9 reserve water system is a large (2500 cubic metres) water tank located at a high elevation in the reactor building. The reserve water system, while incorporating the recovery system functions, and the non-dousing functions of the dousing tank in CANDU 6, embraces other key systems to significantly extend the passive makeup/heat sink capability. The capabilities of the reserve water system include makeup to the steam generators secondary side if all other sources of water are lost; makeup to the heat transport system in the event of a leak in excess of the D 2 O makeup system capability; makeup to the moderator in the event of a moderator leak when the moderator heat sink is required; makeup to the emergency core cooling (ECC) system to assure NPSH to the ECC pumps during a loss of coolant accident (LOCA), and provision of a passive heat sink for the shield cooling system. Other passive designs are now being developed by AECL. These will be incorporated in future CANDU plants when their performance has been fully proven. This paper reviews the passive heat removal systems and features of current CANDU plants and the CANDU 9, and briefly reviews some of the passive heat removal concepts now being developed. (author)

  6. Latina nuclear power plant

    International Nuclear Information System (INIS)

    1976-03-01

    In the period under review, the Latina power plant produced 1009,07 million kWh with a utilization factor of 72% and an availability factor of 80,51%. The disparity between the utilization and availability factors was mainly due to the shutdown of the plant owing to trade union strife. The reasons for non-availability (19,49%) were almost all related to the functioning of the conventional part and the general servicing of the plant (18 September-28 October). During the shutdown for maintenance, an inspection of the steel members and parts of the core stabilizing structure was made in order to check for the familiar oxidation phenomena caused by CO 2 ; the results of the inspection were all satisfactory. Operation of the plant during 1974 was marked by numerous power cutbacks as a result of outages of the steam-raising units (leaks from the manifolds) and main turbines (inspection and repairs to the LP rotors). Since it was first brought into commercial operation, the plant has produced 13,4 thousand million kWh

  7. Wind power plant

    Energy Technology Data Exchange (ETDEWEB)

    Weiss, G

    1975-11-20

    A wind power plant is proposed suitable for electicity generation or water pumping. This plant is to be self-adjusting to various wind velocities and to be kept in operation even during violent storms. For this purpose the mast, carrying the wind rotor and pivotable around a horizontal axis is tiltable and equipped with a wind blind. Further claims contain various configurations of the tilting base resp. the cut in of an elastic link, the attachment and design of the wind blind as well as the constructive arrangement of one or more dynamos.

  8. Power plant process computer

    International Nuclear Information System (INIS)

    Koch, R.

    1982-01-01

    The concept of instrumentation and control in nuclear power plants incorporates the use of process computers for tasks which are on-line in respect to real-time requirements but not closed-loop in respect to closed-loop control. The general scope of tasks is: - alarm annunciation on CRT's - data logging - data recording for post trip reviews and plant behaviour analysis - nuclear data computation - graphic displays. Process computers are used additionally for dedicated tasks such as the aeroball measuring system, the turbine stress evaluator. Further applications are personal dose supervision and access monitoring. (orig.)

  9. Fossil power plant automation

    International Nuclear Information System (INIS)

    Divakaruni, S.M.; Touchton, G.

    1991-01-01

    This paper elaborates on issues facing the utilities industry and seeks to address how new computer-based control and automation technologies resulting from recent microprocessor evolution, can improve fossil plant operations and maintenance. This in turn can assist utilities to emerge stronger from the challenges ahead. Many presentations at the first ISA/EPRI co-sponsored conference are targeted towards improving the use of computer and control systems in the fossil and nuclear power plants and we believe this to be the right forum to share our ideas

  10. Conceptual designs for very high-temperature CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bushby, S.J.; Dimmick, G.R.; Duffey, R.B. [Atomic Energy of Canada Ltd., Chalk River, Ontario (Canada)

    2000-07-01

    Although its environmental benefits are demonstrable, nuclear power must be economically competitive with other energy sources to ensure it retains, or increases, its share of the changing and emerging energy markets of the next decades. In recognition of this, AECL is studying advanced reactor concepts with the goal of significant reductions in capital cost through increased thermodynamic efficiency and plant simplification. The program, generically called CANDU-X, examines concepts for the future, but builds on the success of the current CANDU designs by keeping the same fundamental design characteristics: excellent neutron economy for maximum flexibility in fuel cycle; an efficient heavy-water moderator that provides a passive heat sink under upset conditions; and, horizontal fuel channels that enable on-line refueling for optimum fuel utilization and power profiles. Retaining the same design fundamentals takes maximum advantage of the existing experience base, and allows technological and design improvements developed for CANDU-X to be incorporated into more evolutionary CANDU plants in the short to medium term. Three conceptual designs have been developed that use supercritical water (SCW) as a coolant. The increased coolant temperature results in the thermodynamic efficiency of each CANDU-X concept being significantly higher than conventional nuclear plants. The first concept, CANDU-X Mark 1, is a logical extension of the current CANDU design to higher operating temperatures. To take maximum advantage of the high heat capacity of water at the pseudo-critical temperature, water at nominally 25 MPa enters the core at 310{sup o}C, and exits at {approx}410{sup o}C. The high specific heat also leads to high heat transfer coefficients between the fuel cladding and the coolant. As a result, Zr-alloys can be used as cladding, thereby retaining relatively high neutron economy. The second concept, CANDU-X NC, is aimed at markets that require smaller simpler distributed

  11. Conceptual designs for very high-temperature CANDU reactors

    International Nuclear Information System (INIS)

    Bushby, S.J.; Dimmick, G.R.; Duffey, R.B.

    2000-01-01

    Although its environmental benefits are demonstrable, nuclear power must be economically competitive with other energy sources to ensure it retains, or increases, its share of the changing and emerging energy markets of the next decades. In recognition of this, AECL is studying advanced reactor concepts with the goal of significant reductions in capital cost through increased thermodynamic efficiency and plant simplification. The program, generically called CANDU-X, examines concepts for the future, but builds on the success of the current CANDU designs by keeping the same fundamental design characteristics: excellent neutron economy for maximum flexibility in fuel cycle; an efficient heavy-water moderator that provides a passive heat sink under upset conditions; and, horizontal fuel channels that enable on-line refueling for optimum fuel utilization and power profiles. Retaining the same design fundamentals takes maximum advantage of the existing experience base, and allows technological and design improvements developed for CANDU-X to be incorporated into more evolutionary CANDU plants in the short to medium term. Three conceptual designs have been developed that use supercritical water (SCW) as a coolant. The increased coolant temperature results in the thermodynamic efficiency of each CANDU-X concept being significantly higher than conventional nuclear plants. The first concept, CANDU-X Mark 1, is a logical extension of the current CANDU design to higher operating temperatures. To take maximum advantage of the high heat capacity of water at the pseudo-critical temperature, water at nominally 25 MPa enters the core at 310 o C, and exits at ∼410 o C. The high specific heat also leads to high heat transfer coefficients between the fuel cladding and the coolant. As a result, Zr-alloys can be used as cladding, thereby retaining relatively high neutron economy. The second concept, CANDU-X NC, is aimed at markets that require smaller simpler distributed power plants

  12. Virtual power plant auctions

    International Nuclear Information System (INIS)

    Ausubel, Lawrence M.; Cramton, Peter

    2010-01-01

    Since their advent in 2001, virtual power plant (VPP) auctions have been implemented widely. In this paper, we describe the simultaneous ascending-clock auction format that has been used for virtually all VPP auctions to date, elaborating on other design choices that most VPP auctions have had in common as well as discussing a few aspects that have varied significantly among VPP auctions. We then evaluate the various objectives of regulators in requiring VPP auctions, concluding that the auctions have been effective devices for facilitating new entry into electricity markets and for developing wholesale power markets. (author)

  13. Virtual power plant auctions

    Energy Technology Data Exchange (ETDEWEB)

    Ausubel, Lawrence M.; Cramton, Peter [Department of Economics, University of Maryland, College Park, MD 20742 (United States)

    2010-12-15

    Since their advent in 2001, virtual power plant (VPP) auctions have been implemented widely. In this paper, we describe the simultaneous ascending-clock auction format that has been used for virtually all VPP auctions to date, elaborating on other design choices that most VPP auctions have had in common as well as discussing a few aspects that have varied significantly among VPP auctions. We then evaluate the various objectives of regulators in requiring VPP auctions, concluding that the auctions have been effective devices for facilitating new entry into electricity markets and for developing wholesale power markets. (author)

  14. Nuclear power plant

    International Nuclear Information System (INIS)

    Aisaka, Tatsuyoshi; Kamahara, Hisato; Yanagisawa, Ko.

    1982-01-01

    Purpose: To prevent corrosion stress cracks in structural materials in a BWR type nuclear power plant by decreasing the oxygen concentration in the reactor coolants. Constitution: A hydrogen injector is connected between the condensator and a condensate clean up system of a nuclear power plant. The injector is incorporated with hydrogenated compounds formed from metal hydrides, for example, of alloys such as lanthanum-nickel alloy, iron titanium alloy, vanadium, palladium, magnesium-copper alloy, magnesium-nickel alloy and the like. Even if the pressure of hydrogen obtained from a hydrogen bomb or by way of water electrolysis is changed, the hydrogen can always be injected into a reactor coolant at a pressure equal to the equilibrium dissociation pressure for metal hydride by introducing the hydrogen into the hydrogen injector. (Seki, T.)

  15. CANDU reactor - supporting the nuclear renaissance

    International Nuclear Information System (INIS)

    Oberth, R.

    2010-01-01

    'Full text:' The CANDU reactor has proven to be a strong performer in both the Canada, with 22 units constructed in Ontario, New Brunswick and Quebec, as well as in Argentina, Korea, Romania and China where a further nine units are operating and two in the planning stage. The average lifetime capacity factor of the CANDU reactor fleet is 89%. The last seven CANDU projects in Korea, China, and Romania have been completed on budget and on schedule. CANDU reactors have the highest uranium utilization efficiency measures as electricity output per ton of uranium mined. The CANDU fuel channel design using on-power fuelling and a heavy water moderator enables flexible fueling options - from the current natural uranium option to burning uranium recovered from used LWR reactor fuel and even a thorium-based fuel. AECL and the CANDU reactor are poised to participate in the worldwide construction at least 250 new reactors over the next 20 years. (author)

  16. Nuclear power plant

    International Nuclear Information System (INIS)

    Schabert, H.P.; Laurer, E.

    1976-01-01

    The invention concerns a quick-acting valve on the main-steam pipe of a nuclear power plant. The engineering design of the valve is to be improved. To the main valve disc, a piston-operated auxiliary valve disc is to be assigned closing a section of the area of the main valve disc. This way it is avoided that the drive of the main valve disc has to carry out different movements. 15 sub-claims. (UWI) [de

  17. Fusion power plant economics

    International Nuclear Information System (INIS)

    Miller, R.L.

    1996-01-01

    The rationale, methodology, and updated comparative results of cost projections for magnetic-fusion-energy central-station electric power plants are considered. Changing market and regulatory conditions, particularly in the U.S., prompt fundamental reconsideration of what constitutes a competitive future energy-source technology and has implications for the direction and emphasis of appropriate near-term research and development programs, for fusion and other advanced generation systems. 36 refs., 2 figs., 2 tabs

  18. Power plant emissions reduction

    Science.gov (United States)

    Anand, Ashok Kumar; Nagarjuna Reddy, Thirumala Reddy

    2015-10-20

    A system for improved emissions performance of a power plant generally includes an exhaust gas recirculation system having an exhaust gas compressor disposed downstream from the combustor, a condensation collection system at least partially disposed upstream from the exhaust gas compressor, and a mixing chamber in fluid communication with the exhaust gas compressor and the condensation collection system, where the mixing chamber is in fluid communication with the combustor.

  19. ROP design for Enhanced CANDU 6 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hu, J.; Scherbakova, D; Kastanya, D.; Ovanes, M. [Candu Energy Inc., Mississauga, Ontario (Canada)

    2011-07-01

    The Enhanced CANDU 6 (EC6) nuclear power plant is a mid-sized pressurized heavy water reactor design, based on the highly successful CANDU 6 (C6) family of power plants, upgraded to meet today's Canadian and international safety requirements and to satisfy Generation III expectations. The EC6 reactor is equipped with two independent Regional Overpower Protection (ROP) systems to prevent overpowers in the reactor fuel. The ROP system design, retaining the traditional C6 methodology, is determined to cover the End-of-Life (EOL) reactor core condition since the reactor operating/thermal margin gradually decreases as plant equipment ages. Several design changes have been incorporated into the reference C6 plant to mitigate the ageing effect on the ROP trip margin. This paper outlines the basis for the EC6 ROP physics design and presents the ROP related improvements made in the EC6 design to ensure that full power operation is not limited by the ROP throughout the entire life of the reactor. (author)

  20. Feasibility Study for Cobalt Bundle Loading to CANDU Reactor Core

    International Nuclear Information System (INIS)

    Park, Donghwan; Kim, Youngae; Kim, Sungmin

    2016-01-01

    CANDU units are generally used to produce cobalt-60 at Bruce and Point Lepreau in Canada and Embalse in Argentina. China has started production of cobalt-60 using its CANDU 6 Qinshan Phase III nuclear power plant in 2009. For cobalt-60 production, the reactor’s full complement of stainless steel adjusters is replaced with neutronically equivalent cobalt-59 adjusters, which are essentially invisible to reactor operation. With its very high neutron flux and optimized fuel burn-up, the CANDU has a very high cobalt-60 production rate in a relatively short time. This makes CANDU an excellent vehicle for bulk cobalt-60 production. Several studies have been performed to produce cobalt-60 using adjuster rod at Wolsong nuclear power plant. This study proposed new concept for producing cobalt-60 and performed the feasibility study. Bundle typed cobalt loading concept is proposed and evaluated the feasibility to fuel management without physics and system design change. The requirement to load cobalt bundle to the core was considered and several channels are nominated. The production of cobalt-60 source is very depend on the flux level and burnup directly. But the neutron absorption characteristic of cobalt bundle is too high, so optimizing design study is needed in the future

  1. Feasibility Study for Cobalt Bundle Loading to CANDU Reactor Core

    Energy Technology Data Exchange (ETDEWEB)

    Park, Donghwan; Kim, Youngae; Kim, Sungmin [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    CANDU units are generally used to produce cobalt-60 at Bruce and Point Lepreau in Canada and Embalse in Argentina. China has started production of cobalt-60 using its CANDU 6 Qinshan Phase III nuclear power plant in 2009. For cobalt-60 production, the reactor’s full complement of stainless steel adjusters is replaced with neutronically equivalent cobalt-59 adjusters, which are essentially invisible to reactor operation. With its very high neutron flux and optimized fuel burn-up, the CANDU has a very high cobalt-60 production rate in a relatively short time. This makes CANDU an excellent vehicle for bulk cobalt-60 production. Several studies have been performed to produce cobalt-60 using adjuster rod at Wolsong nuclear power plant. This study proposed new concept for producing cobalt-60 and performed the feasibility study. Bundle typed cobalt loading concept is proposed and evaluated the feasibility to fuel management without physics and system design change. The requirement to load cobalt bundle to the core was considered and several channels are nominated. The production of cobalt-60 source is very depend on the flux level and burnup directly. But the neutron absorption characteristic of cobalt bundle is too high, so optimizing design study is needed in the future.

  2. Feasibility study of CANDU-9 fuel handling system

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Jeong Ki; Jo, C. H.; Kim, H. M.

    1996-12-01

    CANDU`s combination of natural uranium, heavy water and on-power refuelling is unique in its design and remarkable for reliable power production. In order to offer more output, better site utilization, shorter construction time, improved station layout, safety enhancements and better control panel layout, CANDU-9 is now under development with design improvement added to all proven CANDU advantages or applicable technologies. One of its major improvement has been applied to fuel handling system. This system is very similar to that of CANDU-3, and some parts of the system are applied to those of the existing CANDU-6 or CANDU-9. Design concepts and design requirements of fuel handling system for CANDU-9 have been identified to compare with those of the existing CANDU and the design feasibilities have been evaluated. (author). 4 tabs., 13 figs., 9 refs.

  3. Atomic power plant

    International Nuclear Information System (INIS)

    Kawakami, Hiroto.

    1975-01-01

    Object: To permit decay heat to be reliably removed after reactor shut-down at such instance as occurrence of loss of power by means of an emergency water supply pump. Structure: An atomic power plant having a closed cycle constructed by connecting a vapor generator, a vapor valve, a turbine having a generator, a condenser, and a water supply pump in the mentioned order, and provided with an emergency water supply pump operated when there is a loss of power to the water supply pump, a degasifier pressure holding means for holding the pressure of the degasifier by introducing part of the vapor produced from said vapor generator, and a valve for discharge to atmosphere provided on the downstream side of said vapor generator. (Kamimura, M.)

  4. On nuclear power plant uprating

    International Nuclear Information System (INIS)

    Ho, S. Allen; Bailey, James V.; Maginnis, Stephen T.

    2004-01-01

    Power uprating for commercial nuclear power plants has become increasingly attractive because of pragmatic reasons. It provides quick return on investment and competitive financial benefits, while involving low risks regarding plant safety and public objection. This paper briefly discussed nuclear plant uprating guidelines, scope for design basis analysis and engineering evaluation, and presented the Salem nuclear power plant uprating study for illustration purposes. A cost and benefit evaluation of the Salem power uprating was also included. (author)

  5. Improved operability of the CANDU 9 control centre

    International Nuclear Information System (INIS)

    Macbeth, M. J.; Webster, A.

    1996-01-01

    The next generation CANDU nuclear power plant being designed by AECL is the 900 MWe class CANDU 9 station. It is based upon the Darlington CANDU station design which is among the world leaders in capacity factor with low Operation, Maintenance and Administration (OM and A) costs. This Control Centre design includes the proven functionality of existing CANDU control centres (including the Wolsong 2,3, and 4 control centre improvements, such as the Emergency Core Cooling panels), the characteristics identified by systematic design with human factors analysis of operations requirements and the advanced features needed to improve station operability which is made possible by the application of new technology. The CANDU 9 Control Centre provides plant staff with an improved operability capability due to the combination of proveness, systematic design with human factors engineering and enhanced operating features. Significant features which contribute to this improved operability include: · Standard NSP, BOP and F/H panels with controls and indicators integrated by a standard display/presentation philosophy. · Common plant parameter signal database for extensive monitoring, checking, display and annunciation. · Powerful annunciation system allowing alarm filtering, prioritizing and interrogation to enhance staff recognition of events, plant state and required corrective procedural actions. · The use of an overview display to present immediate and uncomplicated plant status information to facilitate operator awareness of unit status in a highly readable and recognizable format. · Extensive cross checking of similar process parameters amongst themselves, with the counterpart safety system parameters and as well as with 'signature' values obtained from known steady state conditions. · Powerful calculation capabilities, using the plant wide database, providing immediate recognizable and readable and readable output data on plant state information and plant state change

  6. 9th International conference on CANDU maintenance

    International Nuclear Information System (INIS)

    2011-01-01

    The 9th International Conference on CANDU Maintenance was held in Toronto, Ontario, Canada on December 4-6, 2011. The conference focused on Nuclear plant reliability and maintenance. Equipment reliability is a critical factor in achieving safe and reliable Nuclear Power Plant operations for many reasons. For one it reduces the challenges upon the operating staff and allows station personnel to 'go on the offence' instead of having to play defense in a reactive mode. Plant reliability ensures that there is time to study the issues in detail and develop solutions for long-term success. Let us not forget that the owner carries the burden of demonstrating the return on investment, and plant reliability goes a long way in helping to make the case for plant life extension. A good reputation for plant reliability provides the public capital necessary to build confidence and facilitate the licensing process. The proceedings papers and presentations given at the 9th International Conference on Candu Maintenance covered topics that include: Managing Worker; Radiation Dose; Full Life Cycle Management; Managing Maintenance Refurbishment; Designing for Maintainability; Inspection Techniques; and, Mitigating Degradation.

  7. Luncheon address: Development of the CANDU reactor

    International Nuclear Information System (INIS)

    Bain, A.S.

    1997-01-01

    The paper is a highlight of the some of the achievements in the development of the CANDU Reactor, taken from the book C anada Enters the Nuclear Age . The CANDU reactor is one of Canada's greatest scientific/engineering achievements, that started in the 1940's and bore fruit with the reactors of the 60's, 70's, and 80's. The Government decided in the 1950's to proceed with a demonstration nuclear power reactor (NPD), AECL invited 7 Canadian corporations to bid on a contract to design and construct the NPD plant. General Electric was selected. A utility was also essential for participation and Ontario Hydro was chosen. In May 1957 it was concluded that the minimum commercial size would be about 200MWe and it should use horizontal pressure tubes to contain the fuel and pressurized heavy water coolant. The book also talks of standard out-reactor components such as pumps, valves, steam generators and piping. A major in-reactor component of interest was the fuel, fuel channels and pressure tubes. A very high level of cooperation was required for the success of the CANDU program

  8. The future for CANDU

    International Nuclear Information System (INIS)

    Foster, J.S.

    1977-06-01

    Canada could have 60,000 MW(e) of installed nuclear-electric generating capacity by the year 2000 and have exported the plan to generate a further 5,000 MW(e). While the CANDU reactor can readily be scaled up to larger unit sizes, its real potential lies in the even greater efficiency that can be obtained by using alternative fuel cycles. The thorium - uranium-233 fuel cycle, for instance, makes it possible to attain a conversion factor of unity, or a little better, on a feed of pure thorium in a substantially unmodified CANDU reactor. Further developments, such as spallation, offer means of converting fertile to fissile material to provide a fissile inventory for an expanding system. The coincidence of expected future shortages of other energy supplies with continuing good experience in the nuclear field should assist in creating a climate that will permit accelerated nuclear power development. (author)

  9. Garigliano nuclear power plant

    International Nuclear Information System (INIS)

    1976-03-01

    During the period under review, the Garigliano power station produced 1,028,77 million kWh with a utilization factor of 73,41% and an availability factor of 85,64%. The disparity between the utilization and availability factors was mainly due to a shutdown of about one and half months owing to lack of staff at the plant. The reasons for nonavailability (14.36%) break down as follows: nuclear reasons 11,49%; conventional reasons 2,81%; other reasons 0,06%. During the period under review, no fuel replacements took place. The plant functioned throughout with a single reactor reticulation pump and resulting maximum available capacity of 150 MWe gross. After the month of August, the plant was operated at levels slightly below the maximum available capacity in order to lengthen the fuel cycle. The total number of outages during the period under review was 11. Since the plant was brought into commercial operation, it has produced 9.226 million kWh

  10. Mobile power plant units

    Energy Technology Data Exchange (ETDEWEB)

    Radtke, R

    1979-10-05

    Diesel engines of the MaK line 282 AK/332 with a cylinder power up to 160 kW are used, either as 6-cylinder or 8-cylinder in-line engine or as 12-cylinder V engine. Fuel consumption is between 207 and 212 g/kW. The engine is mounted on a frame, together with a generator. The fuel reserve in the tank will last for 8 hours. The lubricating system, the cooling water and starting air system, the switchboard system, and the frame are described. The switchboard plant is mounted either on a skid undercarriage or on the undercarriage. The plant can be operated independently or parallel to the network. The unit can be remote-controlled via push buttons or control knobs. A picture is presented of a mobile diesel aggregate which is in service in Libya.

  11. Comparative studies between nuclear power plants and hydroelectric power plants

    International Nuclear Information System (INIS)

    Menegassi, J.

    1984-01-01

    This paper shows the quantitative evolution of the power plants in the main countries of the world. The Brazilian situation is analysed, with emphasys in the technical and economical aspects related to power production by hidroelectric or nuclear power plants. The conclusion is that the electricity produced by hidro power plants becomes not economics when is intended to be produced at large distances from the demand centers. (Author) [pt

  12. Nuclear power plant disasters

    International Nuclear Information System (INIS)

    Trott, K.R.

    1979-01-01

    The possibility of a nuclear power plant disaster is small but not excluded: in its event, assistance to the affected population mainly depends on local practitioners. Already existing diseases have to be diagnosed and treated; moreover, these physicians are responsible for the early detection of those individuals exposed to radiation doses high enough to induce acute illness. Here we present the pathogenesis, clinical development and possible diagnostic and therapeutical problems related to acute radiation-induced diseases. The differentiation of persons according to therapy need and prognosis is done on the sole base of the clinical evidence and the peripheral blood count. (orig.) [de

  13. Demonstration tokamak power plant

    International Nuclear Information System (INIS)

    Abdou, M.; Baker, C.; Brooks, J.; Ehst, D.; Mattas, R.; Smith, D.L.; DeFreece, D.; Morgan, G.D.; Trachsel, C.

    1983-01-01

    A conceptual design for a tokamak demonstration power plant (DEMO) was developed. A large part of the study focused on examining the key issues and identifying the R and D needs for: (1) current drive for steady-state operation, (2) impurity control and exhaust, (3) tritium breeding blanket, and (4) reactor configuration and maintenance. Impurity control and exhaust will not be covered in this paper but is discussed in another paper in these proceedings, entitled Key Issues of FED/INTOR Impurity Control System

  14. Siting nuclear power plants

    International Nuclear Information System (INIS)

    Yellin, J.; Joskow, P.L.

    1980-01-01

    The first edition of this journal is devoted to the policies and problems of siting nuclear power plants and the question of how far commercial reactors should be placed from urban areas. The article is divided into four major siting issues: policies, risk evaluation, accident consequences, and economic and physical constraints. One concern is how to treat currently operating reactors and those under construction that were established under less-stringent criteria if siting is to be used as a way to limit the consequences of accidents. Mehanical cost-benefit analyses are not as appropriate as the systematic use of empirical observations in assessing the values involved. Stricter siting rules are justified because (1) opposition because of safety is growing: (2) remote siting will make the industry more stable; (3) the conflict is eliminated between regulatory policies and the probability basis for nuclear insurance; and (4) joint ownership of utilities and power-pooling are increasing. 227 references, 7 tables

  15. Commissioning of Qinshan phase III PHWR nuclear power plant (2 x 700 MW)

    International Nuclear Information System (INIS)

    Gu Jun

    2004-12-01

    As the first CANDU type NPP built in China, the commissioning team established a very efficient and high standard commissioning management system. Unit 1 was put into commercial operation 43 days in advance and 112 days ahead of schedule for Unit 2. Commissioning quality achieved international advanced level. The commissioning period created new world history record of heavy water nuclear power plants. A summary for the practice and the experience of TQNPC obtained in the commissioning of the two unit was given. (authors)

  16. Pulsed nuclear power plant

    International Nuclear Information System (INIS)

    David, C.V.

    1986-01-01

    This patent describes a nuclear power plant. This power plant consists of: 1.) a cavity; 2.) a detonatable nuclear device in a central region of the cavity; 3.) a working fluid inside of the cavity; 4.) a method to denote a nuclear device inside of the cavity; 5.) a mechanical projection from an interior wall of the cavity for recoiling to absorb a shock wave produced by the detonation of the nuclear device and thereby protecting the cavity from damage. A plurality of segments defines a shell within the cavity and a plurality of shock absorbers, each connecting a corresponding segment to a corresponding location on the wall of the cavity. Each of these shock absorbers regulate the recoil action of the segments; and 6.) means for permitting controlled extraction of a quantity of hot gases from the cavity produced by the vaporization of the working fluid upon detonation of the nuclear device. A method of generating power is also described. This method consists of: 1.) introducing a quantity of water in an underground cavity; 2.) heating the water in the cavity to form saturated steam; 3.) detonating a nuclear device at a central location inside the cavity; 4.) recoiling plate-like elements inside the cavity away from the central location in a mechanically regulated and controlled manner to absorb a shock wave produced by the nuclear device detonation and thereby protect the underground cavity against damage; 5.) extracting a quantity of superheated steam produced by the detonation of the nuclear device; and 6.) Converting the energy in the extracted superheated steam into electrical power

  17. Wind power plant system services

    DEFF Research Database (Denmark)

    Basit, Abdul; Altin, Müfit

    Traditionally, conventional power plants have the task to support the power system, by supplying power balancing services. These services are required by the power system operators in order to secure a safe and reliable operation of the power system. However, as in the future the wind power...... is going more and more to replace conventional power plants, the sources of conventional reserve available to the system will be reduced and fewer conventional plants will be available on-line to share the regulation burden. The reliable operation of highly wind power integrated power system might...... then beat risk unless the wind power plants (WPPs) are able to support and participate in power balancing services. The objective of this PhD project is to develop and analyse control strategies which can increase the WPPs capability to provide system services, such as active power balancing control...

  18. Supporting CANDU operators-CANDU owners group

    International Nuclear Information System (INIS)

    Collingwood, B.R.

    1997-01-01

    The CANDU Owners Group (COG) was formed in 1984 by the Canadian CANDU owning utilities and Atomic Energy of Canada limited (AECL). Participation was subsequently extended to all CANDU owners world-wide. The mandate of the COG organization is to provide a framework for co-operation, mutual assistance and exchange of information for the successful support, development, operation, maintenance and economics of CANDU nuclear electric generating stations. To meet these objectives COG established co-operative programs in two areas: 1. Station Support. 2. Research and Development. In addition, joint projects are administered by COG on a case by case basis where CANDU owners can benefit from sharing of costs

  19. Industrial safety in power plants

    International Nuclear Information System (INIS)

    1987-01-01

    The proceedings of the VGB conference 'Industrial safety in power plants' held in the Gruga-Halle, Essen on January 21 and 22, 1987, contain the papers reporting on: Management responsibility for and legal consequences of industrial safety; VBG 2.0 Industrial Accident Prevention Regulation and the power plant operator; Operational experience gained with wet-type flue gas desulphurization systems; Flue gas desulphurization systems: Industrial-safety-related requirements to be met in planning and operation; the effects of the Hazardous Substances Ordinance on power plant operation; Occupational health aspects of heat-exposed jobs in power plants; Regulations of the Industrial Accident Insurance Associations concerning heat-exposed jobs and industrial medical practice; The new VBG 30 Accident Prevention Regulation 'Nuclear power plants'; Industrial safety in nuclear power plants; safe working on and within containers and confined spaces; Application of respiratory protection equipment in power plants. (HAG) [de

  20. How and when will self-assessment improve maintenance of CANDU plants - evaluator's view

    International Nuclear Information System (INIS)

    Mika, K.

    1997-01-01

    Learning organizations use self-evaluation programs to compare their actual performance with their management's expectations and industry standards. In general, self-evaluations identify areas needing improvement. Self-evaluation activities in maintenance departments include self-assessments, management monitoring observations, root cause analyses, event investigations, benchmarking, use of operating experience, self-checking, and problem reporting systems. In Ontario Hydro Nuclear (OHN) we have used almost all forms of self-evaluation with mixed success. However, we did not use self-assessments. With emphasis on excellent maintenance this powerful form of self-evaluation is being introduced to all our plants. Because of its recency and relative inexperience with it, this paper is focused on self-assessment. The paper provides the author's perspective on how a self-assessment program would be evaluated by an independent internal evaluation and what attributes should be in place in order for maintenance departments to succeed in the implementation and successful continuation of the program. (author)

  1. CANDU channel flow verification

    International Nuclear Information System (INIS)

    Mazalu, N.; Negut, Gh.

    1997-01-01

    The purpose of this evaluation was to obtain accurate information on each channel flow that enables us to assess precisely the level of reactor thermal power and, for reasons of safety, to establish which channel is boiling. In order to assess the channel flow parameters, computer simulations were done with the NUCIRC code and the results were checked by measurements. The complete channel flow measurements were made in the zero power cold condition. In hot conditions there were made flow measurements using the Shut Down System 1 (SDS 1) flow devices from 0.1 % F.P. up to 100 % F.P. The NUCIRC prediction for CANDU channel flows and the measurements by Ultrasonic Flow Meter at zero power cold conditions and SDS 1 flow channel measurements at different reactor power levels showed an acceptable agreement. The 100 % F.P. average errors for channel flow of R, shows that suitable NUCIRC flow assessment can be made. So, it can be done a fair prediction of the reactor power distribution. NUCIRC can predict accurately the onset of boiling and helps to warn at the possible power instabilities at high powers or it can detect the flow blockages. The thermal hydraulic analyst has in NUCIRC a suitable tool to do accurate predictions for the thermal hydraulic parameters for different steady state power levels which subsequently leads to an optimal CANDU reactor operation. (authors)

  2. Nuclear Power Plant 1996

    International Nuclear Information System (INIS)

    1997-01-01

    Again this year, our magazine presents the details of the conference on Spanish nuclear power plant operation held in February and that was devoted to 1996 operating results. The Protocol for Establishment of a New Electrical Sector Regulation that was signed last December will undoubtedly represent a new challenge for the nuclear industry. By clearing stating that current standards of quality and safety should be maintained or even increased if possible, the Protocol will force the Sector to improve its productivity, which is already high as demonstrated by the results of the last few years described during this conference and by recent sectorial economic studies. Generation of a nuclear kWh that can compete with other types of power plants is the new challenge for the Sector's professionals, who do not fear the new liberalization policies and approaching competition. Lower inflation and the resulting lower interest rates, apart from being representative indices of our economy's marked improvement, will be very helpful in facing this challenge. (Author)

  3. Plutonium dispositioning in CANDU

    International Nuclear Information System (INIS)

    Boczar, P.G.; Feinroth, H.; Luxat, J.C.

    1995-07-01

    Recently, the U.S. Department of Energy (DOE) sponsored Atomic Energy of Canada Limited (AECL) to evaluate salient technical, strategic, schedule, and cost-related parameters of using CANDU reactors for dispositioning of weapons-grade plutonium in the form of Mixed OXide (MOX) fuel. A study team, consisting of key staff from the CANDU reactor designers and researchers (AECL), operators (Ontario Hydro) and fuel suppliers, analyzed all significant factors involved in such application, with the objective of identifying an arrangement that would permit the burning of MOX in CANDU at the earliest date. One of Ontario Hydro's multi-unit stations, Bruce A nuclear generating station (4x769 MW(e)), was chosen as the reference for the study. The assessment showed that no significant modifications of reactor or process systems are necessary to operate with a full MOX core. Plant modifications would be limited to fuel handling and modifications necessary to accommodate enhanced security and safeguards requirements. No safety limitations were identified

  4. Electrical, control and information systems in the Enhanced CANDU 6

    International Nuclear Information System (INIS)

    De Grosbois, J.; Raiskums, G.; Soulard, M.

    2011-01-01

    This paper describes the electrical, control, and information system (EC and I) design feature improvements of the Enhanced CANDU 6 (EC6). These additional features are carefully integrated into the EC6 design platform, and are engineered with consideration of operational feedback, human factors, and leveraging the advantages of digital instrumentation and control (I and C) technology to create a coherent I and C architecture in support of safe and high performance operation. The design drivers for the selection of advanced features are also discussed. The EC6 nuclear power plant is a mid-sized Pressurized Heavy Water Reactor design, based on the highly successful CANDU 6 family of power plants, and upgraded to meet today's Canadian and international safety requirements and to satisfy Generation 3 design expectations. (author)

  5. Computational fluid dynamics analysis for flow accelerated corrosion in CANDU6 feeder pipes

    International Nuclear Information System (INIS)

    Catana, A.; Pauna, E.; Ioan, M.

    2013-01-01

    CANDU6 plant management over a long time period includes various ageing and degradation mechanisms like FAC manifested mainly at first and second elbow of CANDU6 outlet feeders. FAC take place at all CANDU6 built before 2000 year with feeders made from SA106 grade B low alloy carbon-steel (with chromium at 0.02%). CFD method is used in this paper to investigate the feeder's wall thinning process taking place mainly due local flow conditions in complex 3D geometrical configurations. The 380 outlet feeders grouped in 2.5'' (320) and 2.0'' feeders (60). The objective of this paper is to help, as much as possible, to focus investigation on most probable maximum thinning rate locations through 3D distribution of some TH parameters. Application of CFD methods in CANDU6 nuclear reactors implies the knowledge of real plant operating data like: long term time averaged channel power and mass flow as well as temperature, pressure, pHa etc allowing the optimization and cost reduction of wall thinning monitoring process at CANDU6 nuclear power plants. (authors)

  6. Nuclear power plants

    International Nuclear Information System (INIS)

    Ushijima, Susumu.

    1984-01-01

    Purpose: To enable to prevent the degradation in the quality of condensated water in a case where sea water leakage should occur in a steam condenser of a BWR type nuclear power plant. Constitution: Increase in the ion concentration in condensated water is detected by an ion concentration detector and the leaking factor of sea water is calculated in a leaking factor calculator. If the sea water leaking factor exceeds a predetermined value, a leak generation signal is sent from a judging device to a reactor power control device to reduce the reactor power. At ehe same tiem, the leak generation signal is also sent to a steam condenser selection and isolation device to interrupt the sea water pump of a specified steam condenser based on the signal from the ion concentration detector, as well as close the inlet and outlet valves while open vent and drain valves to thereby forcively discharge the sea water in the cooling water pipes. This can keep the condensate desalting device from ion breaking and prevent the degradation in the quality of the reactor water. (Horiuchi, T.)

  7. The CANDU man-machine interface and simulator training

    International Nuclear Information System (INIS)

    Hinchley, E.M.; Yanofsky, N.

    1982-09-01

    The most significant features of the man-machine interface for CANDU power stations are the extensive use of computer-driven colour graphics displays and the small number of manual controls. The man-machine interface in CANDU stations is designed to present the operator with concise, easy-to-understand information. Future developments in the use of computers in safety shutdown systems, and the use of data highway technologies in plant regulating systems will present special requirements and new opportunities in the application of human factors engineering to the control room. Good man-machine interaction depends on operator training as much as on control room design. In Canada computerized training simulators, which indicate plant response to operator action, are being introducted for operator training. Such simulators support training in normal operation of all plant systems and also in the fault management tasks following malfunctions

  8. Three Mile Island - a review of the accident and its implications for CANDU safety

    International Nuclear Information System (INIS)

    Pannell, R.J.; Campbell, F.R.

    1980-03-01

    After the accident at the Three Mile Island-2 reactor all Canadian owners of CANDU nuclear power plants were asked by the Atomic Energy Control Board (AECB) to conduct a design review to assess the reliability of feedwater supply to boilers, the availability of backup cooling systems, and the adequacy of routine and emergency operating procedures. The authors studied the available information on the accident and the replies received from licensees. Their report is in three sections: a description of the accident, the authors' opinions of the underlying causes, and recommendations to the AECB regarding what might be done to confirm or improve the safety of CANDU plants

  9. Hybrid combined cycle power plant

    International Nuclear Information System (INIS)

    Veszely, K.

    2002-01-01

    In case of re-powering the existing pressurised water nuclear power plants by the proposed HCCPP solution, we can increase the electricity output and efficiency significantly. If we convert a traditional nuclear power plant unit to a HCCPP solution, we can achieve a 3.2-5.5 times increase in electricity output and the achievable gross efficiency falls between 46.8-52% and above, depending on the applied solution. These figures emphasise that we should rethink our power plant technologies and we have to explore a great variety of HCCPP solutions. This may give a new direction in the development of nuclear reactors and power plants as well.(author)

  10. Heat sink management during CANDU low level operation

    International Nuclear Information System (INIS)

    Wang Liansheng

    2008-01-01

    This paper introduces the practice of low-level operation with opening on the main heat transport system during an outage for a Candu-6 nuclear power plant, analyses the risks of losing heat sink during this condition, and points out the safety measures and management requirement for controlling such risks. This paper can be used as a reference for improving and optimizing the heat sink management for the coming outages. (author)

  11. Accident prevention in power plants

    International Nuclear Information System (INIS)

    Steyrer, H.

    Large thermal power plants are insured to a great extent at the Industrial Injuries Insurance Institute of Instrument and Electric Engineering. Approximately 4800 employees are registered. The accident frequency according to an evaluation over 12 months lies around 79.8 per year and 1000 employees in fossil-fired power plants, around 34.1 per year and 1000 employees in nuclear power plants, as in nuclear power plants coal handling and ash removal are excluded. Injuries due to radiation were not registered. The crucial points of accidents are mechanical injuries received on solid, sharp-edged and pointed objects (fossil-fired power plants 28.6%, nuclear power plants 41.5%), stumbling, twisting or slipping (fossil-fired power plants 21.8%, nuclear power plants 19.5%) and injuries due to moving machine parts (only nuclear power plants 12.2%). However, accidents due to burns or scalds obtain with 4.2% and less a lower portion than expected. The accident statistics can explain this fact in a way that the typical power plant accident does not exist. (orig./GL) [de

  12. CANDU 6 - the highly successful medium sized reactor

    International Nuclear Information System (INIS)

    Hedges, K. R.; Allen, P. J.; Hopwood, J. M.

    2000-01-01

    The CANDU 6 Pressurized Heavy Water Reactor system, featuring horizontal fuel channels and heavy water moderator continues to evolve, supported by AECL's strong commitment to comprehensive R and D programs. The initial CANDU 6 design started in the 1970's. The first plants went into service in 1983, and the latest version of the plant is under construction in China. With each plant the technology has evolved giving the dual advantages of proveness and modern technology. CANDU 6 delivers important advantages of the CANDU system with benefit to small and medium-sized grids. This technology has been successfully adopted by, and localized to varying extents in, each of the CANDU 6 markets. For example, all CANDU owners obtain their fuel from domestic suppliers. Progressive CANDU development continues at AECL to enhance this medium size product CANDU 6. There are three key CANDU development strategic thrusts: improved economics, fuel cycle flexibility, and enhanced safety. The CANDU 6 product is also enhanced by incorporating improvements and advanced features that will be arising from our CANDU Technology R and D programs in areas such as heavy water and tritium, control and instrumentation, fuel and fuel cycles, systems and equipment and safety and constructability. (author)

  13. Wuergassen nuclear power plant

    International Nuclear Information System (INIS)

    Anon.

    1989-01-01

    The decision of the Federal Court of Administration concerns an application for immediate decommissioning of a nuclear power plant (Wuergassen reactor): The repeal of the permit granted. The decision dismisses the appeal for non-admission lodged by the plaintiffs against the ruling of the Higher Court of Administration (OVG) of North-Rhine Westphalia of December 19th 1988 (File no. 21 AK 8/88). As to the matter in dispute, the Federal Court of Administration confirms the opinion of the Higher Court of Administration. As to the headnotes, reference can be made to that decision. Federal Court of Administration, decision of April 5th 1989 - 7 B 47.89. Lower instance: OVG NW, Az.: 21 AK 8/88. (orig./RST) [de

  14. Nuclear power plant

    International Nuclear Information System (INIS)

    Uruma, Hiroshi

    1998-01-01

    In the first embodiment of the present invention, elements less activated by neutrons are used as reactor core structural materials placed under high neutron irradiation. In the second embodiment of the present invention, materials less activated by neutrons when corrosive materials intrude to a reactor core are used as structural materials constituting portions where corrosion products are generated. In the third embodiment, chemical species comprising elements less activated by neutrons are used as chemical species to be added to reactor water with an aim of controlling water quality. A nuclear power plant causing less radioactivity can be provided by using structural materials comprising a group of specific elements hardly forming radioactivity by activation of neutrons or by controlling isotope ratios. (N.H.)

  15. Nuclear power plant

    International Nuclear Information System (INIS)

    Schabert, H.P.

    1976-01-01

    A nuclear power plant is described which includes a steam generator supplied via an input inlet with feedwater heated by reactor coolant to generate steam, the steam being conducted to a steam engine having a high pressure stage to which the steam is supplied, and which exhausts the steam through a reheater to a low pressure stage. The reheater is a heat exchanger requiring a supply of hot fluid. To avoid the extra load that would be placed on the steam generator by using a portion of its steam output as such heating fluid, a portion of the water in the steam generator is removed and passed through the reheater, this water having received at least adequate heating in the steam generator to make the reheater effective, but not at the time of its removal being in a boiling condition

  16. Power Plant Replacement Study

    Energy Technology Data Exchange (ETDEWEB)

    Reed, Gary

    2010-09-30

    This report represents the final report for the Eastern Illinois University power plant replacement study. It contains all related documentation from consideration of possible solutions to the final recommended option. Included are the economic justifications associated with the chosen solution along with application for environmental permitting for the selected project for construction. This final report will summarize the results of execution of an EPC (energy performance contract) investment grade audit (IGA) which lead to an energy services agreement (ESA). The project includes scope of work to design and install energy conservation measures which are guaranteed by the contractor to be self-funding over its twenty year contract duration. The cost recovery is derived from systems performance improvements leading to energy savings. The prime focus of this EPC effort is to provide a replacement solution for Eastern Illinois University's aging and failing circa 1925 central steam production plant. Twenty-three ECMs were considered viable whose net impact will provide sufficient savings to successfully support the overall project objectives.

  17. Power plant removal costs

    International Nuclear Information System (INIS)

    Ferguson, J.S.

    1998-01-01

    The financial, regulatory and political significance of the estimated high removal costs of nuclear power plants has generated considerable interest in recent years, and the political significance has resulted in the Nuclear Regulatory Commission (NRC) eliminating the use of conventional depreciation accounting for the decontamination portion of the removal (decommissioning). While nuclear plant licensees are not precluded from utilizing conventional depreciation accounting for the demolition of non-radioactive structures and site restoration, state and federal utility regulators have not been favorably inclined to requests for this distinction. The realization that steam-generating units will be more expensive to remove, relative to their original cost, predates the realization that nuclear units will be expensive. However, the nuclear issues have overshadowed this realization, but are unlikely to continue to do so. Numerous utilities have prepared cost estimates for steam generating units, and this presentation discusses the implications of a number of such estimates that are a matter of public record. The estimates cover nearly 400 gas, oil, coal and lignite generating units. The earliest estimate was made in 1978, and for analysis purposes the author has segregated them between gas and oil units, and coal and lignite units

  18. Power Plant Replacement Study

    Energy Technology Data Exchange (ETDEWEB)

    Reed, Gary

    2010-09-30

    This report represents the final report for the Eastern Illinois University power plant replacement study. It contains all related documentation from consideration of possible solutions to the final recommended option. Included are the economic justifications associated with the chosen solution along with application for environmental permitting for the selected project for construction. This final report will summarize the results of execution of an EPC (energy performance contract) investment grade audit (IGA) which lead to an energy services agreement (ESA). The project includes scope of work to design and install energy conservation measures which are guaranteed by the contractor to be self-funding over its twenty year contract duration. The cost recovery is derived from systems performance improvements leading to energy savings. The prime focus of this EPC effort is to provide a replacement solution for Eastern Illinois University’s aging and failing circa 1925 central steam production plant. Twenty-three ECMs were considered viable whose net impact will provide sufficient savings to successfully support the overall project objectives.

  19. Perspectives of nuclear power plants

    International Nuclear Information System (INIS)

    Vajda, Gy.

    2001-01-01

    In several countries the construction of nuclear power plants has been stopped, and in some counties several plants have been decommissioned or are planned to. Therefore, the question arises: have nuclear power plants any future? According to the author, the question should be reformulated: can mankind survive without nuclear power? To examine this challenge, the global power demand and its trends are analyzed. According to the results, traditional energy sources cannot be adequate to supply power. Therefore, a reconsideration of nuclear power should be imminent. The economic, environmental attractions are discussed as opposite to the lack of social support. (R.P.)

  20. Overview of nuclear power plant equipment qualification issues and practices

    International Nuclear Information System (INIS)

    Torr, K.G.

    1989-01-01

    This report presents a view of and commentary on the current status of equipment qualification (EQ) in nuclear industries of the major western nations. The introductory chapters discuss the concepts of EQ, the elements of EQ process and highlight some of the key issues in EQ. A brief review of industry practices and some of the prevalent industrial standards is presented, followed by an overview of current regulatory positions in the USA, France, Germany and Sweden. A summary and commentary on the latest research findings on issues relating to accident simulation, to aging simulation and some special topics related to EQ, has been contributed by Franklin Research Centre of Philadelphia. The last part of the report deals with equipment qualification in Canada and gives recommendations on EQ for new plants as well as currently operational CANDU nuclear power plants

  1. Wind-power plant

    Energy Technology Data Exchange (ETDEWEB)

    Kling, A

    1976-08-26

    The invention is concerned with a wind-power plant whose rotor axis is pivoted in the supporting structure and swingable around an axis of tilt, forming an angle with the rotor axis and the vertical axis, and allowing precession of the rotor. On changes of wind direction an electric positioning device is moving the rotor axis into the new direction in such a way that no precession forces are exerted on the supporting structure and this one may very easily be held. Instead of one rotor, also a type with two coaxial, co-planar countercurrent rotors may be used. Each of the two countercurrent rotors is carrying a number of magnetic poles, distributed all over the circumference, acting together with the magnetic poles of the other rotor. At least the poles of one rotor have electric line windings being connected by leads with a collector so that the two rotors form the two parts of a power generator being each rotatable with respect to the other ('stator' and 'rotor').

  2. Distributed control and instrumentation systems for future nuclear power plants

    International Nuclear Information System (INIS)

    Yan, G.; L'Archeveque, J.V.R.

    1976-01-01

    The centralized dual computer system philosophy has evolved as the key concept underlying the highly successful application of direct digital control in CANDU power reactors. After more than a decade, this basis philosophy bears re-examination in the light of advances in system concepts--notably distributed architectures. A number of related experimental programs, all aimed at exploring the prospects of applying distributed systems in Canadian nuclear power plants are discussed. It was realized from the outset that the successful application of distributed systems depends on the availability of a highly reliable, high capacity, low cost communications medium. Accordingly, an experimental facility has been established and experiments have been defined to address such problem areas as interprocess communications, distributed data base design and man/machine interfaces. The design of a first application to be installed at the NRU/NRX research reactors is progressing well

  3. Management of Spent Nuclear Fuel from Nuclear Power Plant Reactor

    International Nuclear Information System (INIS)

    Wati, Nurokhim

    2008-01-01

    Management of spent nuclear fuel from Nuclear Power Plant (NPP) reactor had been studied to anticipate program of NPP operation in Indonesia. In this paper the quantity of generated spent nuclear fuel (SNF) is predicted based on the national electrical demand, power grade and type of reactor. Data was estimated using Pressurized Water Reactor (PWR) NPP type 1.000 MWe and the SNF management overview base on the experiences of some countries that have NPP. There are four strategy nuclear fuel cycle which can be developed i.e: direct disposal, reprocessing, DUPlC (Direct Use of Spent PWR Fuel In Candu) and wait and see. There are four alternative for SNF management i.e : storage at the reactor building (AR), away from reactor (AFR) using wet centralized storage, dry centralized storage AFR and prepare for reprocessing facility. For the Indonesian case, centralized facility of the wet type is recommended for PWR or BWR spent fuel. (author)

  4. Delivery improvements for CANDU projects

    International Nuclear Information System (INIS)

    Stephen Yu; Ken Hedges

    1998-01-01

    Future CANDU design will continue to meet emerging design and performance requirements as expected by the operating utilities, and will integrate new technologies into both the product features and work processes. Elements of this risk reduction strategy include feedback of lessons learned from operating plants, project experiences from previous projects, and replication of successful systems and equipment. Project implementation risk is minimized by up-front engineering and licensing prior to contract start. Enhanced competitiveness of the CANDU products is ensured by incorporating improvements based on updated technology. This paper summarizes the strategy used to enhance competitiveness of the CANDU products and the measures introduced to minimize risk during project implementation. This strategy provides a balance between innovation and proven designs; and between the desire for safety and operational improvements and the cost to achieve the improvements

  5. Man-machine interaction in Canadian nuclear power plants

    International Nuclear Information System (INIS)

    Olmstead, R.A.

    1994-01-01

    The design of man-machine interaction in the CANDU plants has evolved considerably over several generations of plants, from the first Douglas Point plant through to the next generation of plants represented by new designs like CANDU 3. In the early plants, the control room configuration was based on designers' projections of control interface requirements. With succeeding generations of designs, there has been an evolution towards a more systematic consideration of human strengths and weaknesses, increasing attention to formal requirements definition, and incorporation of a larger base of operational experience. This paper describes the design of the man-machine interaction for third generation CANDU-3 control rooms for improved operator reliability and reduced costs. (author). 13 refs., 3 figs

  6. Risk analysis due to the extension of STI for CANDU diesel generators

    Energy Technology Data Exchange (ETDEWEB)

    Jee, Moon Hak; Choi, Kwang Hee; Jung, Hyun Jong; Choi, Seong Soo [Korea electric Power Research Institute, Taejon (Korea, Republic of); Lim, Jae Won [Atomic Creative Technology Company, Taejon (Korea, Republic of); Song, Jin Bae [KHNP, Kyungju (Korea, Republic of)

    2005-07-01

    The purpose of this study is to provide technical rationale for the extension of the Surveillance Test Interval (STI) of the Standby Diesel Generator (SDG) and the Emergency Power Supply Diesel Generator (EPSDG) of CANDU plants in Korea in a reliability aspect. The current STI of 2 weeks aims to be extended to 4 weeks through this study.

  7. Brighter for small power plants

    International Nuclear Information System (INIS)

    Haaland, Leif

    2003-01-01

    The article presents a small tunnel drilling machine aimed at using for the construction of small hydroelectric power plants and mentions briefly some advantages economically and environmentally of both the machine and the power production solution

  8. Multi-purpose use of the advanced CANDU compact simulator

    International Nuclear Information System (INIS)

    Lam, K.Y.; MacBeth, M.J.

    1997-01-01

    A near full-scope dynamic model of a CANDU-PHWR (Canadian Deuterium Uranium Pressurized Heavy Water) nuclear power plant was constructed as a multi-purpose advanced Compact Simulator using CASSIM (Cassiopeia Simulation) development system. This Compact Simulator has played an integral part in the design and verification of the CANDU 900 MW control centre mock-up located in the Atomic Energy of Canada (AECL) design office, providing CANDU plant process dynamic data to the Plant Display System (PDS) and the Distributed Control System (DCS), as well as mock-up panel devices. As a design tool, the Compact Simulator is intended to be used for control strategy development, human factors studies, analysis of overall plant control performance, tuning estimates for major control loops. As a plant commissioning and operational strategy development tool, the simulation is intended to be used to evaluate routine and non-routine operational procedures, practice 'what-if' scenarios for operational strategy development, practice malfunction recovery procedures and verify human factors activities

  9. Study of seismic responses of Candu-3 reactor building using isolator bearings

    International Nuclear Information System (INIS)

    Biswas, J.K.

    1992-01-01

    Seismic isolator bearings are known to increase reliability, reduce cost and increase the potential sitings for nuclear power plants located in regions of high seismicity. High seismic activities in Canada occur mainly in the western coast, the Grand Banks and regions of Quebec along the St. Lawrence river. In Canada, nuclear power plants are located in Ontario, Quebec and New Brunswick where the seismicity levels are low to moderate. Consequently, seismic isolator bearings have not been used in the existing nuclear power plants in Canada. The present paper examines the effect of using seismic isolator bearings in the design for the new CANDU3 which would be suitable for regions having high seismicity. The CANDU3 Nuclear Power Plant is rated at 450 MW of net output power and is a smaller version of its predecessor CANDU6 successfully operating in Canada and abroad. The design of CANDU3 is being developed by AECL CANDU. Advanced technologies for design, construction and plant operation have been utilized. During the conceptual development of the CANDU3 design, various design options including the use of isolator bearings were considered. The present paper presents an overview of seismic isolation technology and summarizes the analytical work for predicting the seismic behavior of the CANDU3 reactor building. A lumped-parameter dynamic model for the reactor building is used for the analysis. The characteristics of the bearings are utilized in the analysis work. The time-history modal analysis has been used to compute the seismic responses. Seismic responses of the reactor building with and without isolator bearings are compared. The isolator bearings are found to reduce the accelerations of the reactor building. As a result, a lower level of seismic qualification for components and systems would be required. The use of these bearings however increases rigid body seismic displacements of the structure requiring special considerations in the layout and interfaces for

  10. Are atomic power plants saver than nuclear power plants

    International Nuclear Information System (INIS)

    Roeglin, H.C.

    1977-01-01

    It is rather impossible to establish nuclear power plants against the resistance of the population. To prevail over this resistance, a clarification of the citizens-initiatives motives which led to it will be necessary. This is to say: It is quite impossible for our population to understand what really heappens in nuclear power plants. They cannot identify themselves with nuclear power plants and thus feel very uncomfortable. As the total population feels the same way it is prepared for solidarity with the citizens-initiatives even if they believe in the necessity of nuclear power plants. Only an information-policy making transparent the social-psychological reasons of the population for being against nuclear power plants could be able to prevail over the resistance. More information about the technical procedures is not sufficient at all. (orig.) [de

  11. Employing modern power plant simulators in nuclear power plants

    International Nuclear Information System (INIS)

    Niedorf, V.; Storm, J.

    2005-01-01

    At the present state of the art, modern power plant simulators are characterized by new qualitative features, thus enabling operators to use them far beyond the traditional field of training. In its first part, this contribution presents an overview of the requirements to be met by simulators for multivalent uses. In part two, a survey of the uses and perspectives of simulation technology in power plants is presented on the basis of experience accumulated by Rheinmetall Defence Electronics (RDE).Modern simulators are shown to have applications by far exceeding traditional training areas. Modular client - sever systems on standard computers allow inexpensive uses to be designed at several levels, thus minimizing maintenance cost. Complex development and running time environments, like the SEMS developed by RDE, have made power plant simulators the workhorses of power plant engineers in all power plant areas. (orig.)

  12. 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.)

  13. The year 2000 power plant

    International Nuclear Information System (INIS)

    Roman, H.T.

    1989-01-01

    Every utility seeks extended service life from its existing power plants before building new ones. It is not easy to justify a new power plant. The licensing and cost of new plants have become uncertain. In response to these conditions, electric utilities are undertaking plant life-extension studies and, in some cases, reconditioning/upgrading old power plants to significantly increase useful service life. Other technologies like robotics and artificial intelligence/expert systems are also being developed to reduce operating and maintenance (O and M) expenses, to remove workers from potentially hazardous environments, and to reduce plant downtime. Together, these steps represent an interim solution, perhaps providing some relief for the next few decades. However, there are serious physical and economic limits to retrofitting new technology into existing power plants. Some old plants will simply be beyond their useful life and require retirement. In nuclear plants, for instance, retrofit may raise important and time-consuming licensing/safety issues. Based on their robotics and artificial intelligence experience, the authors of this article speculate bout the design of the year 2000 power plant - a power plant they feel will naturally incorporate liberal amounts of robotic and artificial intelligence technologies

  14. Images of nuclear power plants

    International Nuclear Information System (INIS)

    Hashiguchi, Katsuhisa; Misumi, Jyuji; Yamada, Akira; Sakurai, Yukihiro; Seki, Fumiyasu; Shinohara, Hirofumi; Misumi, Emiko; Kinjou, Akira; Kubo, Tomonori.

    1995-01-01

    This study was conducted to check and see, using Hayashi's quantification method III, whether or not the respondents differed in their images of a nuclear power plant, depending on their demographic variables particularly occupations. In our simple tabulation, we compared subject groups of nuclear power plant employees with general citizens, nurses and students in terms of their images of a nuclear power plant. The results were that while the nuclear power plant employees were high in their evaluations of facts about a nuclear power plant and in their positive images of a nuclear power plant, general citizens, nurses and students were overwhelmingly high in their negative images of a nuclear power plant. In our analysis on category score by means of the quantification method III, the first correlation axis was the dimension of 'safety'-'danger' and the second correlation axis was the dimension of 'subjectivity'-'objectivity', and that the first quadrant was the area of 'safety-subjectivity', the second quadrant was the area of 'danger-subjectivity', the third quadrant as the area of 'danger-objectivity', and the forth quadrant was the area of 'safety-objectivity'. In our analysis of sample score, 16 occupation groups was compared. As a result, it was found that the 16 occupation groups' images of a nuclear power plant were, in the order of favorableness, (1) section chiefs in charge, maintenance subsection chiefs, maintenance foremen, (2) field leaders from subcontractors, (3) maintenance section members, operation section members, (4) employees of those subcontractors, (5) general citizens, nurses and students. On the 'safety-danger' dimension, nuclear power plant workers on the one hand and general citizens, nurses and students on the other were clearly divided in terms of their images of a nuclear power plant. Nuclear power plant workers were concentrated in the area of 'safety' and general citizens, nurses and students in the area of 'danger'. (J.P.N.)

  15. Applicability of a track-based multiprocess portable robot to some maintenance tasks in CANDU nuclear plants

    International Nuclear Information System (INIS)

    Hazel, B.; Fihey, J.-L.; Laroche, Y.

    2000-01-01

    Hydro-Quebec has developed a six-axis, track-based, multiprocess robot. This lightweight (30 kg) compact robot travels on a bent track with a radius of curvature ranging from 1 m to infinity (straight track). Standard and tandem wires GMAW, FCAW and Narrow gap TIG welding as well as plasma gouging and cutting, electrical and pneumatic rough and precision grinding, and profile measurement functionalities have been incorporated. A description of this technology an its newly developed functionalities is given in this paper. Since 1995, a number of industrial and R and D projects have been performed using this technology now called the Scompi technology. The main field of application is the in situ repair of hydraulic turbine runners. However some applications have been developed in the nuclear field. One particular development was funded by the International Thermonuclear Experimental Reactor (ITER) project. Scompi was selected by the ITER US Home Team for a demonstration of remote techniques for welding, cutting and rewelding the 30 m diameter, 17 m high, vacuum vessel. The demonstration involved all position robotic plasma cutting and NG-TIG welding of a 316L, 40 mm thick, double wall. In 1998, two Scompi robots working in tandem performed in York, Pa, the joint welding and cutting of a full scale portion of the vacuum vessel. In 1995, the applicability of the Scompi technology to the repair of the divider plates in the four steam generators at Gentilly-2 was evaluated based on a joint proposal by Ontario Hydro Technologies (now Ontario Power Technologies-OPT) and Hydro-Quebec. A MIG welding procedure was proposed for the horizontal and vertical divider plates welds. A complete simulation of the robot and primary head demonstrated the feasibility of the concept. However, based on cost and scheduling, it was decided to proceed with a manual repair. Nevertheless it is anticipated that this technology will find its niche in the maintenance of Candu reactors. (author)

  16. Applicability of a track-based multiprocess portable robot to some maintenance tasks in CANDU nuclear plants

    Energy Technology Data Exchange (ETDEWEB)

    Hazel, B.; Fihey, J.-L.; Laroche, Y. [Hydro-Quebec, Varennes, Quebec (Canada)

    2000-07-01

    Hydro-Quebec has developed a six-axis, track-based, multiprocess robot. This lightweight (30 kg) compact robot travels on a bent track with a radius of curvature ranging from 1 m to infinity (straight track). Standard and tandem wires GMAW, FCAW and Narrow gap TIG welding as well as plasma gouging and cutting, electrical and pneumatic rough and precision grinding, and profile measurement functionalities have been incorporated. A description of this technology an its newly developed functionalities is given in this paper. Since 1995, a number of industrial and R and D projects have been performed using this technology now called the Scompi technology. The main field of application is the in situ repair of hydraulic turbine runners. However some applications have been developed in the nuclear field. One particular development was funded by the International Thermonuclear Experimental Reactor (ITER) project. Scompi was selected by the ITER US Home Team for a demonstration of remote techniques for welding, cutting and rewelding the 30 m diameter, 17 m high, vacuum vessel. The demonstration involved all position robotic plasma cutting and NG-TIG welding of a 316L, 40 mm thick, double wall. In 1998, two Scompi robots working in tandem performed in York, Pa, the joint welding and cutting of a full scale portion of the vacuum vessel. In 1995, the applicability of the Scompi technology to the repair of the divider plates in the four steam generators at Gentilly-2 was evaluated based on a joint proposal by Ontario Hydro Technologies (now Ontario Power Technologies-OPT) and Hydro-Quebec. A MIG welding procedure was proposed for the horizontal and vertical divider plates welds. A complete simulation of the robot and primary head demonstrated the feasibility of the concept. However, based on cost and scheduling, it was decided to proceed with a manual repair. Nevertheless it is anticipated that this technology will find its niche in the maintenance of Candu reactors. (author)

  17. Nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Urata, Hidehiro; Oya, Takashi

    1996-11-05

    The present invention provides a highly safe light water-cooled type nuclear power plant capable of reducing radiation dose by suppressing deposition of activated corrosion products by a simple constitution. Namely, equipments and pipelines for fluid such as pumps at least in one of fluid systems such as a condensate cleanup system are constituted by a material containing metal species such as Zn having an effect of suppressing deposition of radioactivity. Alternatively, the surface of these equipments and pipelines for fluids on which water passes is formed by a coating layer comprising a material containing a metal having a radiation deposition suppressing effect. As a result, radioactivity deposited on the equipments and pipelines for fluids is reduced. In addition, since the method described above may be applied only at least to a portion of the members constituting at least one of the systems for fluids, it is economical. Accordingly, radiation dose upon inspection of equipments and pipelines for fluids can be reduced simply and reliably. (I.S.)

  18. Nuclear power plant

    International Nuclear Information System (INIS)

    Urata, Hidehiro; Oya, Takashi.

    1996-01-01

    The present invention provides a highly safe light water-cooled type nuclear power plant capable of reducing radiation dose by suppressing deposition of activated corrosion products by a simple constitution. Namely, equipments and pipelines for fluid such as pumps at least in one of fluid systems such as a condensate cleanup system are constituted by a material containing metal species such as Zn having an effect of suppressing deposition of radioactivity. Alternatively, the surface of these equipments and pipelines for fluids on which water passes is formed by a coating layer comprising a material containing a metal having a radiation deposition suppressing effect. As a result, radioactivity deposited on the equipments and pipelines for fluids is reduced. In addition, since the method described above may be applied only at least to a portion of the members constituting at least one of the systems for fluids, it is economical. Accordingly, radiation dose upon inspection of equipments and pipelines for fluids can be reduced simply and reliably. (I.S.)

  19. Pipelines in power plants

    International Nuclear Information System (INIS)

    Oude-Hengel, H.H.

    1978-01-01

    Since the end of the Sixties, steam-transporting pipelines are given great attention, as pipeline components often fail, partially even long before their designed operation time is over. Thus, experts must increasingly deal with questions of pipelines and their components. Design and calculation, production and operation of pipelines are included in the discussion. Within the frame of this discussion, planners, producers, operators, and technical surveillance personnel must be able to offer a homogenous 'plan for assuring the quality of pipelines' in fossil and nuclear power plants. This book tries to make a contribution to this topic. 'Quality assuring' means efforts made for meeting the demands of quality (reliability). The book does not intend to complete with well-known manuals, as for as a complete covering of the topic is concerned. A substantial part of its sections serves to show how quality assurance of pipelines can be at least partially obtained by surveillance measures beginning with the planning, covering the production, and finally accompanying the operation. There is hardly need to mention that the sort of planning, production, and operation has an important influence on the quality. This is why another part of the sections contain process aspects from the view of the planners, producers, and operators. (orig.) [de

  20. Nuclear power plants

    International Nuclear Information System (INIS)

    Kiyokawa, Teruyuki; Soman, Yoshindo.

    1985-01-01

    Purpose: To constitute a heat exchanger as one unit by integrating primary and secondary coolant circuits with secondary coolant circuit and steam circuit into a single primary circuit and steam circuit. Constitution: A nuclear power plant comprises a nuclear reactor vessel, primary coolant pipeways and a leakage detection system, in which a dual-pipe type heat exchanger is connected to the primary circuit pipeway. The heat conduction tube of the heat exchanger has a dual pipe structure, in which the inside of the inner tube is connected to the primary circuit pipeway, the outside of the outer tube is connected to steam circuit pipeway and a fluid channel is disposed between the inner and outer tubes and the fluid channel is connected to the inside of an expansion tank for intermediate heat medium. The leak detection system is disposed to the intermediate heat medium expansion tank. Sodium as the intermediate heat medium is introduced from the intermediate portion (between the inner and outer tubes) by way of inermediate heat medium pipeways to the intermediate heat medium expansion tank and, further, to the intermediate portion for recycling. (Kawakami, Y.)

  1. Underground nuclear power plant

    International Nuclear Information System (INIS)

    Takahashi, Hideo.

    1997-01-01

    In an underground-type nuclear power plant, groups of containing cavities comprising a plurality of containing cavities connected in series laterally by way of partition walls are disposed in parallel underground. Controlled communication tunnels for communicating the containing cavities belonging to a control region to each other, and non-controlled communication tunnels for communicating containing cavities belonging to a non-controlled area to each other are disposed underground. A controlled corridor tunnel and a non-controlled corridor tunnel extended so as to surround the containing cavity groups are disposed underground, and the containing cavities belonging to the controlled area are connected to the controlled corridor tunnel respectively, and the containing cavities belonging to the non-controlled area are connected to the non-controlled corridor tunnel respectively. The excavating amount of earth and sand upon construction can be reduced by disposing the containing cavity groups comprising a plurality of containing cavities connected in series laterally. The time and the cost for the construction can be reduced, and various excellent effects can be provided. (N.H.)

  2. CANDU flexible and economical fuel technology in China

    Energy Technology Data Exchange (ETDEWEB)

    Mingjun, C. [CNNC Nuclear Power Operation Management Co., Zhejiang (China); Zhenhua, Z.; Zhiliang, M. [CNNC Third Qinshan Nuclear Power Co., Zhejiang (China); Cottrell, C.M.; Kuran, S. [Candu Energy Inc., Mississauga, ON (Canada)

    2014-07-01

    Use in CANDU reactor is one good option of recycled uranium (RU) and thorium (Th) resource. It is also good economy to CANDU fuel. Since 2008 Qinshan CANDU Plant and our partners (Candu Energy and CNNC and NPIC) have made great efforts to develop the engineering technologies of Flexible and Economical Fuel (RU and Th) in CANDU type reactor and finding the CANDU's position in Chinese closed fuel cycle (CFC) system. This paper presents a proposal of developing strategy and implementation plan. Qinshan CANDU reactors will be converted to use recycled and depleted uranium based fuels, a first-of-its-kind. The fuel is composed of both recycled and depleted uranium and simulating natural uranium behavior. This paper discusses its development, design, manufacture and verification tested with success and the full core implementation plan by the end of 2014. (author)

  3. Development of a Web-based CANDU Core Management Procedure Automation System

    International Nuclear Information System (INIS)

    Lee, Sanghoon; Kim, Eunggon; Park, Daeyou; Yeom, Choongsub; Suh, Hyungbum; Kim, Sungmin

    2006-01-01

    CANDU reactor core needs efficient core management to increase safety, stability, high performance as well as to decrease operational cost. The most characteristic feature of CANDU is so called 'on-power refueling' i.e., there is no shutdown during refueling in opposition to that of PWR. Although this on-power refueling increases the efficiency of the plant, it requires heavy operational task and difficulties in real time operation such as regulating power distribution, burnup distribution, LZC statistics, the position of control devices and so on. To enhance the CANDU core management, there are several approaches to help operator and reduce difficulties, one of them is the COMOS (CANDU Core On-line Monitoring System). It has developed as an online core surveillance system based on the standard incre instrumentation and the numerical analysis codes such as RFSP (Reactor Fueling Simulation Program). As the procedure is getting more complex and the number of programs is increased, it is required that integrated and cooperative system. So, KHNP and IAE have been developing a new web-based system which can support effective and accurate reactor operational environment called COMPAS that means CANDU cOre Management Procedure Automation System. To ensure development of successful system, several steps of identifying requirements have been performed and Software Requirement Specification (SRS) document was developed. In this paper we emphasis on the how to keep consistency between the requirements and system products by applying requirement traceability methodology

  4. Extrapolating power-ramp performance criteria for current and advanced CANDU fuels

    Energy Technology Data Exchange (ETDEWEB)

    Tayal, M.; Chassie, G.G

    2000-06-01

    To improve the precision and accuracy of power-ramp performance criteria for high-burnup fuel, we have examined in-reactor fuel performance data as well as out-reactor test data. The data are consistent with some of the concepts used in the current formulations for defining fuel failure thresholds, such as size of power-ramp and extent of burnup. Our review indicates that there is a need to modify some other aspects of the current formulations; therefore, a modified formulation is presented in this paper. The improvements mainly concern corrodent concentration and its relationships with threshold stress for failure. The new formulation is consistent with known and expected trends such as strength of Zircaloy in corrosive environment, timing of the release of fission products to the pellet-to-sheath gap, CANLUB coating, and fuel burnup. Because of the increased precision and accuracy, the new formulation is better able to identify operational regimes that are at risk of power-ramp failures; this predictive ability provides enhanced protection to fuel against power-ramp defects. At die same time, by removing unnecessary conservatisms in other areas, the new formulation permits a greater range of defect-free operational envelope as well as larger operating margins in regions that are, in fact, not prone to power-ramp failures. (author)

  5. Power plants and safety 1982

    International Nuclear Information System (INIS)

    1982-01-01

    The papers of this volume deal with the whole range of safety issues from planning and construction to the operation of power plants, and discuss also issues like availability and safety of power plants, protective clothes and their incommodating effect, alternatives for rendering hot-water generators safe and the safety philosophy in steam turbine engineering. (HAG) [de

  6. Thermodynamic optimization of power plants

    NARCIS (Netherlands)

    Haseli, Y.

    2011-01-01

    Thermodynamic Optimization of Power Plants aims to establish and illustrate comparative multi-criteria optimization of various models and configurations of power plants. It intends to show what optimization objectives one may define on the basis of the thermodynamic laws, and how they can be applied

  7. Owners of nuclear power plants

    International Nuclear Information System (INIS)

    Wood, R.S.

    1991-07-01

    This report indicates percentage ownership of commercial nuclear power plants by utility companies. The report includes all plants operating, under construction, docketed for NRC safety and environmental reviews, or under NRC antitrust review, but does not include those plants announced but not yet under review or those plants formally cancelled. Part 1 of the report lists plants alphabetically with their associated applicants or licensees and percentage ownership. Part 2 lists applicants or licensees alphabetically with their associated plants and percentage ownership. Part 1 also indicates which plants have received operating licenses (OLS)

  8. Power exports - a U.S. market for CANDU (question mark)

    International Nuclear Information System (INIS)

    McIntyre, H.C.

    1982-01-01

    The advantages to both countries of power exports from Canada to the U.S.A. are explained. The policy of the Canadian federal government now tends to favour such exports. While Hydro-Quebec leads in exports, Ontario Hydro plans to sell 1.2 GW of coal-generated power to New Jersey. New Brunswick now has applications before the National Energy Board for export of 0.35 GW from Point Lepreau. A second 0.6 GW unit at Point Lepreau, dedicated to the export of electricity, has been suggested

  9. Nuclear power plant diagnostic system

    International Nuclear Information System (INIS)

    Prokop, K.; Volavy, J.

    1982-01-01

    Basic information is presented on diagnostic systems used at nuclear power plants with PWR reactors. They include systems used at the Novovoronezh nuclear power plant in the USSR, at the Nord power plant in the GDR, the system developed at the Hungarian VEIKI institute, the system used at the V-1 nuclear power plant at Jaslovske Bohunice in Czechoslovakia and systems of the Rockwell International company used in US nuclear power plants. These diagnostic systems are basically founded on monitoring vibrations and noise, loose parts, pressure pulsations, neutron noise, coolant leaks and acoustic emissions. The Rockwell International system represents a complex unit whose advantage is the on-line evaluation of signals which gives certain instructions for the given situation directly to the operator. The other described systems process signals using similar methods. Digitized signals only serve off-line computer analyses. (Z.M.)

  10. Korean experience in CANDU-PHWR operation

    International Nuclear Information System (INIS)

    Sang-kee Park

    1987-01-01

    Among KEPCO's 9 nuclear power units, Korea Nuclear Unit No. 3, the Wolsung Nuclear Power Plant is the only CANDU-PHWR Unit, while the rest of 8 others are PWR units. The unit was designed by Atomic Energy of Canada, Ltd(AECL) of Canada, who also perfomed overall project management for the plant construction under the provisions and arrangement of the relevant contracts. The gross electrical output of the plant is 678.7 MWe and thermal output of the reactor is 2061 MWth. While these figures lead to lower plant eficiency than LWR counterparts, unit energy cost for fuel is more favorable than LWRs because natural uranium is utilized for the fuel bundles, some of which are already being fabricated domestically. Annual capacity factors for 1983 and 1984 could have been improved, if two major planned outages for the modification works on steam generator internals and one major forced outage form the heavy water spill incident could be eliminated. The heavy water spill incident in November, 1984 brought plant staffs many lessons to learn and many things to contemplate. Unique design concepts and features such as on-power refuelling, poison prevent mode, versatile plant control system built around digital computers and power step back/set back logics may be credited for these relatively good plant performances. Human related factors such as staff's technical capabilities and strong will toward good performance were other elements which could not be overlooked

  11. Korean experience in CANDU-PHWR operation

    International Nuclear Information System (INIS)

    Park, S.K.

    1988-01-01

    Among KEPCO's 9 nuclear power units, Korea Nuclear Unit No. 3, the Wolsung Nuclear Power Plant is the only CANDU-PHWR Unit, while the rest of 8 others are PWR units. The unit was designed by Atomic Energy of Canada, Ltd. of Canada, who also performed overall project management for the plant construction under the provisions and arrangement of the relevant contracts. The gross electrical output of the plant is 678.7 MWe and thermal output of the reactor is 2061 MWth. While these figures lead to lower plant efficiency than LWR counterparts, unit energy cost for fuel is more favorable than LWRs because natural uranium is utilized for the fuel bundles, some of which are already being fabricated domestically. Annual capacity factors for 1983 and 1984 could have been improved, if two major planned outages for the modification works on steam generator internals and one major forced outage from the heavy water spill incident could be eliminated. The heavy water spill incident in November, 1984 brought plant staffs many lessons to learn and many things to contemplate. Unique design concepts and features such as on-power refuelling, poison prevent mode, versatile plant control system built around digital computers and power step back/set back logics may be credited for these relatively good plant performances. Human related factors such as staff's technical capabilities and strong will toward good performance were other elements which could not be overlooked

  12. Pressure drop variation as a function of axial and radial power distribution in CANDU fuel channel with standard and CANFLEX 43 bundles

    International Nuclear Information System (INIS)

    Catana, Alexandru; Department of Energy Danila, Nicolae; Prisecaru, Ilie; Dupleac, Daniel

    2007-01-01

    CANDU 600 nuclear reactors are usually fuelled with STANDARD (STD), 37 rods fuel bundles. Natural uranium (NU) dioxide (UO 2 ), is used as fuel composition. A new fuel bundle geometry called CANFLEX (CFX) with 43 rods is proposed and some new fuel composition are considered. Flexibility is the key word for the attempt to use some different fuel geometries and compositions for CANDU 600 nuclear reactors as well as for innovative ACR-700/1000 nuclear reactors. The fuel bundle considered in this paper is CFX-RU-0.90 that encodes the CANFLEX geometry, recycled dioxide uranium (RU) with 0.90% enrichment. The goal of this proposal is ambitious: a higher average discharge burn-up up to 14000 MWd/tU and, for the same amount of generated electric power, reduction in nuclear fuel fabrication, reduction of spent nuclear fuel radioactive waste and reduction of refueling operational work by using fewer bundles. An improved sub-channel approach for thermal-hydraulic analysis is used in this paper to compute some flow parameters, mainly the pressure drop along the CANDU 600 fuel channel when STD or CFX-RU-0.90 fuel bundles. Also an intermediate CFX-NU fuel bundle are used, for gradual comparison. For CFX-RU- 0.90 four fuel bundle shift refueling scheme is used instead of eight, that will determine different axial power distributions. At the same time radial power distribution is affected by the geometry and by the fuel composition of fuel bundle type used. Some other thermal-hydraulic flow parameters will be influenced, too. One of the most important parameter is pressure drop (PD) along the fuel channel because of its importance in drag force evaluation. We start with an axial power distribution, which is characteristic for a refueling scheme of eight or four fuel bundles on a shift. Comparative results are presented between STD37, CFX-NU CFX-RU-0.90 fuel bundles in a CANDU nuclear reactor operating conditions. Neutron flux distribution analysis shows that four bundle shift

  13. Nuclear energy in Canada: the CANDU system

    International Nuclear Information System (INIS)

    Robertson, J.A.L.

    1979-10-01

    Nuclear electricity in Canada is generated by CANDU nuclear power stations. The CANDU reactor - a unique Canadian design - is fuelled by natural uranium and moderated by heavy water. The system has consistently outperformed other comparable nuclear power systems in the western world, and has an outstanding record of reliability, safety and economy. As a source of energy it provides the opportunity for decreasing our dependence on dwindling supplies of conventional fossil fuels. (auth)

  14. VGB Congress 'Power Plants 2006'

    International Nuclear Information System (INIS)

    Anon.

    2006-01-01

    The VGB Congress 'Power Plants' took place in Dresden, 27 th to 29 th September 2006 under the auspices of the Federal Minister for Economics and Technology, Michael Glos. The motto of this year's Congress was 'Future becomes Reality - Investments in New Power Plants'. More than 1,200 participants from Germany and abroad attended the plenary and technical lectures on the topics 'Market and Competition' as well as 'Technology, Operation and Environment' for information and discussion. Special papers were dealing with further issues like 'Generation Market in Europe', 'Clean Power Technology Platform', French policy for new power plants as well as potentials and technology of renewables. (orig.)

  15. Nuclear power plant V-1

    International Nuclear Information System (INIS)

    1998-01-01

    The nuclear power plant Bohunice V -1 is briefly described. This NPP consists from two reactor units. Their main time characteristics are (Reactor Unit 1, Reactor Unit 2): beginning of construction - 24 April 1972; first controlled reactor power - 27 November 1978, 15 March 1980; connection to the grid - 17 December 1978, 26 March 1980; commercial operation - 1 April 1980, 7 January 1981. This leaflet contains: NPP V-1 construction; Major technological equipment (Primary circuit: Nuclear reactor [WWER 440 V230 type reactor];Steam generator; Reactor Coolant Pumps; Primary Circuit Auxiliary Systems. Secondary circuit: Turbine generators, Nuclear power plant electrical equipment; power plant control) and technical data

  16. Some environmental effects of emissions from CANDU nuclear generating stations and heavy water plants

    International Nuclear Information System (INIS)

    Effer, W.R.

    Non-radioactive releases during normal operation of Ontario Hydro's nuclear generating stations and heavy water plants are summarized and related to existing regulations and guidelines. Low-grade heat in the circulating cooling water discharge is the most important of the non-radioactive effluents. Some of the hydrological, biological and water quality aspects of thermal discharges are discussed in relation to the operation of Ontario Hydro's thermal generating stations on the Great Lakes. Chemical releases to air or water include chlorine, hydrogen sulphide, water treatment plant effluents, oily waste water and sewage lagoon effluents. The significance of the first two of these releases to the environment is reviewed, particularly in relation to Great Lakes water quality and biological concerns. (author)

  17. Emergency power systems at nuclear power plants

    International Nuclear Information System (INIS)

    1982-01-01

    This Guide applies to nuclear power plants for which the total power supply comprises normal power supply (which is electric) and emergency power supply (which may be electric or a combination of electric and non-electric). In its present form the Guide provides general guidance for all types of emergency power systems (EPS) - electric and non-electric, and specific guidance (see Appendix A) on the design principles and the features of the emergency electric power system (EEPS). Future editions will include a second appendix giving specific guidance on non-electric power systems. Section 3 of this Safety Guide covers information on considerations that should be taken into account relative to the electric grid, the transmission lines, the on-site electrical supply system, and other alternative power sources, in order to provide high overall reliability of the power supply to the EPS. Since the nuclear power plant operator does not usually control off-site facilities, the discussion of methods of improving off-site reliability does not include requirements for facilities not under the operator's control. Sections 4 to 11 of this Guide provide information, recommendations and requirements that would apply to any emergency power system, be it electric or non-electric

  18. CANDU operating experience

    International Nuclear Information System (INIS)

    McConnell, L.G.; Woodhead, L.W.; Fanjoy, G.R.

    1982-03-01

    The CANDU Pressurized Heavy Water (CANDU-PHW) type of nuclear electric generating station has been developed jointly by Atomic Energy of Canada Limited and Ontario Hydro. This paper highlights Ontario Hydro's operating experience using the CANDU-PHW system, with a focus on the operating performance and costs, reliability of system components and nuclear safety considerations both to the workers and the public

  19. AND THERMAL POWER PLANTS

    Directory of Open Access Journals (Sweden)

    Alduhov Oleg Aleksandrovich

    2012-10-01

    Full Text Available Investigation of the atmospheric dispersion as part of the process of selection of sites to accommodate nuclear and thermal power plants is performed to identify concentration fields of emissions and to assess the anthropogenic impact produced on the landscape components and human beings. Scattering properties of the atmospheric boundary layer are mainly determined by the turbulence intensity and the wind field. In its turn, the turbulence intensity is associated with the thermal stratification of the boundary layer. Therefore, research of the atmospheric dispersion is reduced to the study of temperature and wind patterns of the boundary layer. Statistical processing and analysis of the upper-air data involves the input of the data collected by upper-air stations. Until recently, the upper-air data covering the standard period between 1961 and 1970 were applied for these purposes, although these data cannot assure sufficient reliability of assessments in terms of the properties of the atmospheric dispersion. However, recent scientific and technological developments make it possible to substantially increase the data coverage by adding the upper-air data collected within the period between 1964 and 2010. The article has a brief overview of BL_PROGS, a specialized software package designated for the processing of the above data. The software package analyzes the principal properties of the atmospheric dispersion. The use of the proposed software package requires preliminary development of a database that has the information collected by an upper-air station. The software package is noteworthy for the absence of any substantial limitations imposed onto the amount of the input data that may go up in proportion to the amount of the upper-air data collected by upper-air stations.

  20. Qinshan CANDU commissioning - a successful partnership

    International Nuclear Information System (INIS)

    Alikhan, S.; Thomson, J.; Jun, G.; Guoyuan, J.

    2004-01-01

    The Qinshan CANDU Nuclear Power Plant consists of 2 x 728 MWe CANDU 6 units, built in Zhejiang Province, China, by the Third Qinshan Nuclear Power Company (TQNPC) as the owner and Atomic Energy of Canada Limited (AECL) as the main contractor. The Contract between China National Nuclear Corporation (CNNC) and AECL was signed in November 1996 and became effective on February 12, 1997 with scheduled completion dates of February 12, 2003 for Unit 1 and November 12, 2003 for Unit 2. Unit 1 was declared in-service on December 31, 2002, 43 days ahead of schedule and Unit 2 was declared in service on July 20, 2003, 115 days ahead of schedule. The successful partnership between AECL, Bechtel, Hitachi and TQNPC working as a team is the key to this success. Total commissioning period from first energization of the system service transformer to in-service for both units was 20.7 months, which is significantly better than the experience at other comparable CANDU 6 units. It has clearly demonstrated the benefits of building two units together, about 6 months apart, to achieve optimum utilization of resources already mobilized for the first unit; the second unit is commissioned with less than 40% of the effort required for the first unit. Since in-service to the end of March 2004, Unit 1 has operated at a gross capacity factor of 93% and Unit 2 at 82.5%, including loss of production for one month in August 2003 to repair the failure of turbine LP blades tie-wire. (author)

  1. The next generation of CANDU: reactor design to meet future energy markets

    International Nuclear Information System (INIS)

    Hopwood, J.M.; Love, J.W.; Wren, D.J.

    2001-01-01

    Nuclear power plant designs for the future must respond to increasingly demanding market requirements. This means that value can be gained from substantial product development directed at these requirements. For the CANDU system, AECL has adopted the revolutionary approach, accommodating significant changes to design while retaining traditional CANDU strengths. The focus of the new design is to achieve a 40% reduction in capital cost, quicken construction time and higher efficiency. Key aspects of the new design include: light water coolant, smaller core, slightly enriched fuel, higher temperature and pressure coolant. Work is well advanced on the preliminary design

  2. Radiochemistry in nuclear power plants

    International Nuclear Information System (INIS)

    Schwarz, W.

    2007-01-01

    Radiochemistry is employed in nuclear power plants not as an end in itself but, among other things, as a main prerequisite of optimum radiation protection. Radiochemical monitoring of various loops provides important information about sources of radioactivity, activity distribution in the plant and its changes. In the light of these analytical findings, plant crews are able to take measures having a positive effect on radiation levels in the plant. The example of a BWR plant is used to show, among other things, how radiochemical analyses helped to reduce radiation levels in a plant and, as a consequence, to decrease clearly radiation exposure of the personnel despite higher workloads. (orig.)

  3. Peak power and heavy water production from electrolytic H2 and O2 using CANDU reactors

    International Nuclear Information System (INIS)

    Hammerli, M.; Stevens, W.H.; Bradley, W.J.; Butler, J.P.

    1976-04-01

    A combined energy storage - heavy water production system is presented. Off-peak nuclear energy is stored in the form of electrolytic H 2 (and O 2 ) from which a large fraction of the deuterium has been transferred to water in an H 2 /H 2 O deuterium exchange catalytic column. The main features and advantages of the combined electrolysis -catalytic exchange D 2 O process are discussed. Significant quantities of D 2 O could be produced economically at reasonable peak to base power cost ratios. Thirty to forty percent of the primary electric energy should be available for peak energy via either gas-steam turbines or fuel cells. (author)

  4. Analysis of Multiple Spurious Operation Scenarios for Decay Heat Removal Function of CANDU Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Youngseung; Bae, Yeon-kyoung; Kim, Myungsu [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    The worst fire broke out in the Browns Ferry Nuclear Power Plant on March 22, 1975. A fire occurrence in a nuclear power plant has recognized a latently serious incident. Nuclear power plants should achieve and maintain the safe shutdown conditions during and after the occurrence of a fire. Functions of the safe shutdown are five such as the shutdown function, the decay heat removal function, the containment function, monitoring and control function, and the supporting function for CANDU type reactors. The purpose of this paper is to analyze that the decay heat removal function of the safe shutdown functions for CANDU type reactors is achieved under the fire induced multiple spurious operation. The scenarios of the fire induced multiple spurious operations (MSO) for the systems used for the decay heat cooling were analyzed. Additionally, Integrated Severe Accident Analysis code for CANDU plants (ISAAC) for determining success criteria of thermal hydraulic analysis was used. Decay heat cooling systems of CANDU reactors are the auxiliary feedwater system, the emergency water supply system, and the shutdown cooling system. A big fire can threat the safety of nuclear power plants, and safe shutdown conditions. The regulatory body in Korea requires the fire hazard analysis including fire induced MSOs. The safe shutdown functions for CANDU reactors are the shutdown function, the decay heat removal function, the containment function, the monitoring and control function, and the supporting service function. The number of spurious operations for the auxiliary feedwater system is more than six and that for the emergency water supply system is one. Additionally, misoperations for the shutdown cooling system are more than two. Accordingly, if total nine components could be spuriously operated, the decay heat removal function would be lost entirely.

  5. Analysis of Multiple Spurious Operation Scenarios for Decay Heat Removal Function of CANDU Reactors

    International Nuclear Information System (INIS)

    Lee, Youngseung; Bae, Yeon-kyoung; Kim, Myungsu

    2016-01-01

    The worst fire broke out in the Browns Ferry Nuclear Power Plant on March 22, 1975. A fire occurrence in a nuclear power plant has recognized a latently serious incident. Nuclear power plants should achieve and maintain the safe shutdown conditions during and after the occurrence of a fire. Functions of the safe shutdown are five such as the shutdown function, the decay heat removal function, the containment function, monitoring and control function, and the supporting function for CANDU type reactors. The purpose of this paper is to analyze that the decay heat removal function of the safe shutdown functions for CANDU type reactors is achieved under the fire induced multiple spurious operation. The scenarios of the fire induced multiple spurious operations (MSO) for the systems used for the decay heat cooling were analyzed. Additionally, Integrated Severe Accident Analysis code for CANDU plants (ISAAC) for determining success criteria of thermal hydraulic analysis was used. Decay heat cooling systems of CANDU reactors are the auxiliary feedwater system, the emergency water supply system, and the shutdown cooling system. A big fire can threat the safety of nuclear power plants, and safe shutdown conditions. The regulatory body in Korea requires the fire hazard analysis including fire induced MSOs. The safe shutdown functions for CANDU reactors are the shutdown function, the decay heat removal function, the containment function, the monitoring and control function, and the supporting service function. The number of spurious operations for the auxiliary feedwater system is more than six and that for the emergency water supply system is one. Additionally, misoperations for the shutdown cooling system are more than two. Accordingly, if total nine components could be spuriously operated, the decay heat removal function would be lost entirely

  6. Next Generation Geothermal Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Brugman, John; Hattar, Mai; Nichols, Kenneth; Esaki, Yuri

    1995-09-01

    A number of current and prospective power plant concepts were investigated to evaluate their potential to serve as the basis of the next generation geothermal power plant (NGGPP). The NGGPP has been envisaged as a power plant that would be more cost competitive (than current geothermal power plants) with fossil fuel power plants, would efficiently use resources and mitigate the risk of reservoir under-performance, and minimize or eliminate emission of pollutants and consumption of surface and ground water. Power plant concepts were analyzed using resource characteristics at ten different geothermal sites located in the western United States. Concepts were developed into viable power plant processes, capital costs were estimated and levelized busbar costs determined. Thus, the study results should be considered as useful indicators of the commercial viability of the various power plants concepts that were investigated. Broadly, the different power plant concepts that were analyzed in this study fall into the following categories: commercial binary and flash plants, advanced binary plants, advanced flash plants, flash/binary hybrid plants, and fossil/geothed hybrid plants. Commercial binary plants were evaluated using commercial isobutane as a working fluid; both air-cooling and water-cooling were considered. Advanced binary concepts included cycles using synchronous turbine-generators, cycles with metastable expansion, and cycles utilizing mixtures as working fluids. Dual flash steam plants were used as the model for the commercial flash cycle. The following advanced flash concepts were examined: dual flash with rotary separator turbine, dual flash with steam reheater, dual flash with hot water turbine, and subatmospheric flash. Both dual flash and binary cycles were combined with other cycles to develop a number of hybrid cycles: dual flash binary bottoming cycle, dual flash backpressure turbine binary cycle, dual flash gas turbine cycle, and binary gas turbine

  7. Internal dose from tritium at Wolsung nuclear power plant

    International Nuclear Information System (INIS)

    Hee Geun Kim; Jeong Yull Dho; Myung Jae Song

    1995-01-01

    Tritium is produced in large quantities at heavy water nuclear power reactors via the neutron activation reaction 2 H(n,γ) 3 H. At Wolsung nuclear power plant which has a CANDU reactor, the tritium concentrations in coolant and in moderator systems are 1.5 Ci/Kg-D 2 O and 35 Ci/kg-D 2 O, respectively, after 12 years of operation. The airborne tritium concentration in main access area is normally less than 5 MPCa except short-term peaks. The average tritium concentrations in main access controlled areas are normally less than 100 MPCa. Tritium is mainly present in the air of workplace of CANDU reactors as a tritiated water vapour. Airborne tritiated water vapour enters the workers body via inhalation and absorption through skin and can result in a significant dose. The occupational doses from tritium at Wolsung NPP have been maintained below 1 man-Sv per year so far. The tritium contribution to the total plant man-Sv changes between 30 percent and 50 percent. For the mitigation of tritium inhalation, various protective equipment are being used at Wolsung NPP. The respirator system was devised at Wolsung NPP in order to remove tritiated water vapours from the inhaled air. A respirator is connected to a small plastic bottle filled with ice cubes. The system devised shows a good tritium removal efficiency. The air pressure drop through the ice cubes is minimal. The operation cost of the system is also very cheap. Further mitigation of tritium inhalation is heavily dependant on the source term reduction. One of the ultimate solutions is to introduce a tritium removal facility. (author). 7 figs., 3 tabs

  8. Automation technology in power plants

    International Nuclear Information System (INIS)

    Essen, E.R.

    1995-01-01

    In this article a summery of the current architecture of modern process control systems in power plants and future trends have been explained. The further development of process control systems for power plants is influenced both by the developments in component and software technologies as well as the increased requirements of the power plants. The convenient and low cost configuration facilities of new process control systems have now reached a significance which makes it easy for customers to decide to purchase. (A.B.)

  9. Management of a 600 MW CANDU project to facilitate electricity export

    International Nuclear Information System (INIS)

    Gunter, G.E.

    1983-06-01

    The export of electricity from 600-MW CANDU nuclear power plants built in Canada remains feasible providing certain requirements continue to be met. The principal objective in developing nuclear power resources for export is that they must produce economically attractive electricity. A review of the experience of construction and operation of Point Lepreau Unit 1 suggests an inherent ability to reduce construction costs and shorten construction schedules so as to make electrical power output from these stations even more attractive to export customers

  10. Man and nuclear power plants

    International Nuclear Information System (INIS)

    Anon.

    1978-01-01

    According to the Inst. fuer Unfallforschung/TUeV Rheinland, Koeln, the interpretation of empirical data gained from the operation of nuclear power plants at home and abroad during the period 1967-1975 has shown that about 38% of all reactor accidents were caused by human failures. These occured either during the design and construction, the commissioning, the reconditioning or the operation of the plants. This very fact stresses human responsibility for the safety of nuclear power plants, in spite of those plants being automated to a high degree and devices. (orig.) [de

  11. Owners of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Hudson, C.R.; White, V.S.

    1996-11-01

    Commercial nuclear power plants in this country can be owned by a number of separate entities, each with varying ownership proportions. Each of these owners may, in turn, have a parent/subsidiary relationship to other companies. In addition, the operator of the plant may be a different entity as well. This report provides a compilation on the owners/operators for all commercial power reactors in the United States. While the utility industry is currently experiencing changes in organizational structure which may affect nuclear plant ownership, the data in this report is current as of July 1996. The report is divided into sections representing different aspects of nuclear plant ownership.

  12. Owners of nuclear power plants

    International Nuclear Information System (INIS)

    Wood, R.S.

    1979-12-01

    The following list indicates percentage ownership of commercial nuclear power plants by utility companies as of December 1, 1979. The list includes all plants licensed to operate, under construction, docketed for NRC safety and envionmental reviews, or under NRC antitrust review. It does not include those plants announced but not yet under review or those plants formally cancelled. In many cases, ownership may be in the process of changing as a result of antitrust license conditions and hearings, altered financial conditions, changed power needs, and other reasons. However, this list reflects only those ownership percentages of which the NRC has been formally notified

  13. Competitive breeder power plants

    International Nuclear Information System (INIS)

    Winkleblack, R.K.

    1984-01-01

    To utilize the fissile material that is accumulating in the utilities' spent fuel pools, breeder plants must be less expensive than current LWR costs (or utilities will not buy nuclear plants in the near future) and also be highly reliable. The fundamental differences between LWRs and LMFBRs are discussed and recommendations are made for making the most of these differences to design a superior breeder plant that can sell in the future, opening the way to U.S. utilities becoming self-sufficient for fuel supply for centuries

  14. Improved operation in CANDU plants with CAN8 PHT pump seals

    International Nuclear Information System (INIS)

    Graham, T.; McInnes, D.; Rhodes, D.

    1997-01-01

    The CAN8 PHT pump seal is currently operating in twenty-one pumps, twelve at Bruce A, seven at Bruce B and in both pumps at Grand Gulf Nuclear Station (GGNS). The CAN8 seal has markedly improved performance over the CAN2 seal previously used at the Bruce stations and the SU seals previously used at GGNS. Details of the performance improvements are discussed. Prior to installation in Bruce B, the CAN8 seal was slightly modified and then demonstrated to be resistant to reverse pressurization failures, since this was a known failure mechanism with the CAN2 seal. Subsequent experience showed that Bruce A was also susceptible to reverse pressure incidents. A review of plant operating procedures at Bruce A showed reverse pressure was likely the initiating factor for several previously unexplained seal disturbances. The reverse pressure failure mechanism is described, as are the improved system operating procedures designed to prevent it. Preventative procedures have now been implemented across Ontario Hydro Nuclear. The ability to track down seal failure mechanisms such as this is greatly enhanced by the improved system monitoring and data retrieval now in place at Bruce A and Bruce B. (author)

  15. CANDU fuel cycle options in Korea

    International Nuclear Information System (INIS)

    Boczar, P. G.; Fehrenbach, P. J.; Meneley, D. A.

    1996-01-01

    FBR reactors. If the objective of a national fuel-cycle program is the minimization of actinide waste or destruction of long-lived fission products, then studies have shown the superiority of CANDU reactors in meeting this objective. Long-term energy security can be assured either through the thorium cycle or through a CANDU/FBR system, in which the FBR would be operated as a 'fuel factory,'providing the fissile material to power a number of lower-cost, high-efficiency CANDU reactors. In summary, the CANDU reactor's simple fuel design, high neutron economy, and on-line fuelling provide flexibility to respond to changing fuel-cycle requirements in the short term and in the indefinite future

  16. Space nuclear reactor power plants

    International Nuclear Information System (INIS)

    Buden, D.; Ranken, W.A.; Koenig, D.R.

    1980-01-01

    Requirements for electrical and propulsion power for space are expected to increase dramatically in the 1980s. Nuclear power is probably the only source for some deep space missions and a major competitor for many orbital missions, especially those at geosynchronous orbit. Because of the potential requirements, a technology program on space nuclear power plant components has been initiated by the Department of Energy. The missions that are foreseen, the current power plant concept, the technology program plan, and early key results are described

  17. Reactor physics aspects of CANDU reactors

    International Nuclear Information System (INIS)

    Critoph, E.

    1980-01-01

    These four lectures are being given at the Winter Course on Nuclear Physics at Trieste during 1978 February. They constitute part of the third week's lectures in Part II: Reactor Theory and Power Reactors. A physical description of CANDU reactors is given, followed by an overview of CANDU characteristics and some of the design options. Basic lattice physics is discussed in terms of zero energy lattice experiments, irradiation effects and analytical methods. Start-up and commissioning experiments in CANDU reactors are reviewed, and some of the more interesting aspects of operation discussed - fuel management, flux mapping and control of the power distribution. Finally, some of the characteristics of advanced fuel cycles that have been proposed for CANDU reactors are summarized. (author)

  18. Nuclear power plant operator licensing

    International Nuclear Information System (INIS)

    1997-01-01

    The guide applies to the nuclear power plant operator licensing procedure referred to the section 128 of the Finnish Nuclear Energy Degree. The licensing procedure applies to shift supervisors and those operators of the shift teams of nuclear power plant units who manipulate the controls of nuclear power plants systems in the main control room. The qualification requirements presented in the guide also apply to nuclear safety engineers who work in the main control room and provide support to the shift supervisors, operation engineers who are the immediate superiors of shift supervisors, heads of the operational planning units and simulator instructors. The operator licensing procedure for other nuclear facilities are decided case by case. The requirements for the basic education, work experience and the initial, refresher and complementary training of nuclear power plant operating personnel are presented in the YVL guide 1.7. (2 refs.)

  19. Nuclear Power Plant Simulation Game.

    Science.gov (United States)

    Weiss, Fran

    1979-01-01

    Presents a nuclear power plant simulation game which is designed to involve a class of 30 junior or senior high school students. Scientific, ecological, and social issues covered in the game are also presented. (HM)

  20. Calculation of the radial and axial flux and power distribution for a CANDU 6 reactor with both the MCNP6 and Serpent codes

    International Nuclear Information System (INIS)

    Hussein, M.S.; Bonin, H.W.; Lewis, B.J.

    2014-01-01

    The most recent versions of the Monte Carlo-based probabilistic transport code MCNP6 and the continuous energy reactor physics burnup calculation code Serpent allow for a 3-D geometry calculation accounting for the detailed geometry without unit-cell homogenization. These two codes are used to calculate the axial and radial flux and power distributions for a CANDU6 GENTILLY-2 nuclear reactor core with 37-element fuel bundles. The multiplication factor, actual flux distribution and power density distribution were calculated by using a tally combination for MCNP6 and detector analysis for Serpent. Excellent agreement was found in the calculated flux and power distribution. The Serpent code is most efficient in terms of the computational time. (author)

  1. Calculation of the radial and axial flux and power distribution for a CANDU 6 reactor with both the MCNP6 and Serpent codes

    Energy Technology Data Exchange (ETDEWEB)

    Hussein, M.S.; Bonin, H.W., E-mail: mohamed.hussein@rmc.ca, E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, ON (Canada); Lewis, B.J., E-mail: Brent.Lewis@uoit.ca [Univ. of Ontario Inst. of Tech., Faculty of Energy Systems and Nuclear Science, Oshawa, ON (Canada)

    2014-07-01

    The most recent versions of the Monte Carlo-based probabilistic transport code MCNP6 and the continuous energy reactor physics burnup calculation code Serpent allow for a 3-D geometry calculation accounting for the detailed geometry without unit-cell homogenization. These two codes are used to calculate the axial and radial flux and power distributions for a CANDU6 GENTILLY-2 nuclear reactor core with 37-element fuel bundles. The multiplication factor, actual flux distribution and power density distribution were calculated by using a tally combination for MCNP6 and detector analysis for Serpent. Excellent agreement was found in the calculated flux and power distribution. The Serpent code is most efficient in terms of the computational time. (author)

  2. Elecnuc. Nuclear power plants worldwide

    International Nuclear Information System (INIS)

    1998-01-01

    This small folder presents a digest of some useful information concerning the nuclear power plants worldwide and the situation of nuclear industry at the end of 1997: power production of nuclear origin, distribution of reactor types, number of installed units, evolution and prediction of reactor orders, connections to the grid and decommissioning, worldwide development of nuclear power, evolution of power production of nuclear origin, the installed power per reactor type, market shares and exports of the main nuclear engineering companies, power plants constructions and orders situation, evolution of reactors performances during the last 10 years, know-how and development of nuclear safety, the remarkable facts of 1997, the future of nuclear power and the energy policy trends. (J.S.)

  3. Robotics for nuclear power plants

    International Nuclear Information System (INIS)

    Shiraiwa, Takanori; Watanabe, Atsuo; Miyasawa, Tatsuo

    1984-01-01

    Demand for robots in nuclear power plants is increasing of late in order to reduce workers' exposure to radiations. Especially, owing to the progress of microelectronics and robotics, earnest desire is growing for the advent of intellecturized robots that perform indeterminate and complicated security work. Herein represented are the robots recently developed for nuclear power plants and the review of the present status of robotics. (author)

  4. Robotics for nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Shiraiwa, Takanori; Watanabe, Atsuo; Miyasawa, Tatsuo

    1984-10-01

    Demand for robots in nuclear power plants is increasing of late in order to reduce workers' exposure to radiations. Especially, owing to the progress of microelectronics and robotics, earnest desire is growing for the advent of intellecturized robots that perform indeterminate and complicated security work. Herein represented are the robots recently developed for nuclear power plants and the review of the present status of robotics.

  5. Decommissioning of nuclear power plants

    International Nuclear Information System (INIS)

    Vollradt, J.

    1977-01-01

    A survey of the main questions of decommissioning of nuclear power plants will be given in the sight of German utilities (VDEW-Working group 'Stillegung'). The main topics are: 1) Definitions of decommissioning, entombment, removal and combinations of such alternatives; 2) Radioactive inventory (build up and decay); 3) Experience up to now; 4) Possibilities to dismantle are given by possibility to repair nuclear power plants; 5) Estimated costs, waste, occupational radiation dose; 6) German concept of decommissioning. (orig./HK) [de

  6. CANDU Digital Control Computer upgrade options

    International Nuclear Information System (INIS)

    De Jong, M.S.; De Grosbois, J.; Qian, T.

    1997-01-01

    This paper reviews the evolution of Digital Control Computers (DCC) in CANDU power plants to the present day. Much of this evolution has been to meeting changing control or display requirements as well as the replacement of obsolete, or old and less reliable technology with better equipment that is now available. The current work at AECL and Canadian utilities to investigate DCC upgrade options, alternatives, and strategies are examined. The dependence of a particular upgrade strategy on the overall plant refurbishment plans are also discussed. Presently, the upgrade options range from replacement of individual obsolete system components, to replacement of the entire DCC hardware without changing the software, to complete replacement of the DCCs with a functionally equivalent system using new control computer equipment and software. Key issues, constraints and objectives associated with these DCC upgrade options are highlighted. (author)

  7. PV power plants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-07-01

    Within the international seminar of the Ostbayerisches Technologie-Transfer-Institut e.V. (OTTI) at 11th June, 2012 in Munich (Federal Republic of Germany), the following lectures were held: (1) Technical due diligence (Dietmar Obst); (2) Certification / rating system for large PV plants (Robert Pfatischer); (3) O and M requirements (Lars Rulf); (4) IR photography for large scale systems (Bernhard Weinreich); (5) New market models for PV systems - direct marketing and sales of PV electricity (Martin Schneider); (6) Needs and benefits for plant certification for grid connection and operation (Christoph Luetke-Lengerich); (7) Lare volume module testing / Screening in the field and workshop (Semir Merzoug); (8) Dismantling costs of large scale PV plants (Siegfried Schimpf).

  8. Inertial fusion commercial power plants

    International Nuclear Information System (INIS)

    Logan, B.G.

    1994-01-01

    This presentation discusses the motivation for inertial fusion energy, a brief synopsis of five recently-completed inertial fusion power plant designs, some general conclusions drawn from these studies, and an example of an IFE hydrogen synfuel plant to suggest that future fusion studies consider broadening fusion use to low-emission fuels production as well as electricity

  9. Emergency power systems at nuclear power plants

    International Nuclear Information System (INIS)

    1991-01-01

    This Safety Guide was prepared as part of the Nuclear Safety Standards programme for establishing Codes and Safety Guides relating to nuclear power plants (NPPs). The first edition of the present Safety Guide was developed in the early 1980s. The text has now been brought up-to-date, refined in several details and amended to include non-electrical diverse and independent power sources. This Guide applies to NPP for which the total power supply comprises a normal power supply and an emergency power supply (EPS), which may be electrical or a combination of electrical and non-electrical. The Guide provides general guidance for all types of EPS and specific guidance on the design safety requirements and the features of the electrical and non-electrical portions of the EPS. 9 figs, 2 tabs

  10. Organization patterns of PWR power plants

    International Nuclear Information System (INIS)

    Leicman, J.

    1980-01-01

    Organization patterns are shown for the St. Lucia 1, North Anna, Sequoyah, and Beaver Valley nuclear power plants, for a typical PWR power plant in the USA, for the Biblis/RWE-KWU nuclear power plants and for a four-unit nuclear power plant operated by Electricite de France as well as for the Loviisa power plant. Organization patterns are also shown for relatively independent and non-independent nuclear power plants according to IAEA recommendations. (J.P.)

  11. Low-power wind plants

    International Nuclear Information System (INIS)

    Kovalenko, V.I.; Shevchenko, Yu.V.; Shikhajlov, N.A.; Kokhanevich, V.P.; Tanan, G.L.

    1993-01-01

    Design peculiarities, as well as the prospects of development and introduction of the low-power (from 0.5 up to 4 kW) wind power plants (WPP) are considered. The variants of WPP with vertical and horizontal rotation axis are described. The data characterizing cost and structure of expenditures on WPP manufacture and operation are given

  12. Energy sources and power plants

    International Nuclear Information System (INIS)

    Schulz, Detlef; Schulz, Karen

    2013-01-01

    Energy is obtained from various energy sources (coal, petroleum, natural gas, nuclear fuels, wind energy, solar energy, hydro power, biomass, geothermal energy). These differ in each case with respect to their availability, methods of their production and the required power plant technologies. As technologies of the future fuel cells and nuclear fusion are traded. [de

  13. Space power plants

    Science.gov (United States)

    Khudyakov, S. A.

    1985-05-01

    Power generators in space are examined. A semiconducting photoelectric converter (FEP) which converts the energy of solar radiation directly into electrical energy is discussed. The operating principle of an FEP is based on the interaction of solar light with a crystal semiconductor, in the process of which the photons produce free electrons, carriers of an electrical charge, in the crystal. Areas with a strong electrical field created specially under the effect of the p-n junction trap the freed electrons and divide them in such a fashion that a current and corresponding electrical power appear in the load circuit. The absorption of light in metals and pure semiconductors is outlined.

  14. Qinshan CANDU project open top construction method

    International Nuclear Information System (INIS)

    Petrunik, K.J.; Wittann, K.; Khan, A.; Ricciuti, R.; Ivanov, A.; Chen, S.

    2003-01-01

    The significant schedule reductions achieved on the Qinshan CANDU Project were due in large part to the incorporation of advanced construction technologies in project design and delivery. For the Qinshan Project, a number of key advantages were realized through the use of the 'Open Top' construction method. This paper discusses the Qinshan Phase III CANDU Project Open Top implementation method. The Open Top method allowed major equipment to be installed simply, via the use of a Very Heavy Lift (VHL) crane and permitted the use of large-scale modularization. The advantages of Open Top construction, such as simplified access, more flexible project scheduling, improved construction safety and quality, and reduced labours are presented. The large-scale modularization of the Reactor Building Dousing System and the Open Top installation method and advantages relative to traditional CANDU 6 construction practices are also presented. Finally, major improvements for future CANDU plant construction using the Open Top method are discussed. (author)

  15. Alternative Concept to Enhance the Disposal Efficiency for CANDU Spent Fuel Disposal System

    International Nuclear Information System (INIS)

    Lee, Jong Youl; Cho, Dong Geun; Kook, Dong Hak; Lee, Min Soo; Choi, Heui Joo

    2011-01-01

    There are two types of nuclear reactors in Korea and they are PWR type and CANDU type. The safe management of the spent fuels from these reactors is very important factor to maintain the sustainable energy supply with nuclear power plant. In Korea, a reference disposal system for the spent fuels has been developed through a study on the direct disposal of the PWR and CANDU spent fuel. Recently, the research on the demonstration and the efficiency analyses of the disposal system has been performed to make the disposal system safer and more economic. PWR spent fuels which include a lot of reusable material can be considered being recycled and a study on the disposal of HLW from this recycling process is being performed. CANDU spent fuels are considered being disposed of directly in deep geological formation, since they have little reusable material. In this study, based on the Korean Reference spent fuel disposal System (KRS) which was to dispose of both PWR type and CANDU type, the more effective CANDU spent fuel disposal systems were developed. To do this, the disposal canister for CANDU spent fuels was modified to hold the storage basket for 60 bundles which is used in nuclear power plant. With these modified disposal canister concepts, the disposal concepts to meet the thermal requirement that the temperature of the buffer materials should not be over 100 .deg. C were developed. These disposal concepts were reviewed and analyzed in terms of disposal effective factors which were thermal effectiveness, U-density, disposal area, excavation volume, material volume etc. and the most effective concept was proposed. The results of this study will be used in the development of various wastes disposal system together with the HLW wastes from the PWR spent fuel recycling process.

  16. Power plant simulation

    Energy Technology Data Exchange (ETDEWEB)

    Hacking, D [Marconi Simulation (United Kingdom)

    1992-09-01

    Over many years in the field of simulation Marconi has developed and adopted a number of procedures and methodologies for the management, design and development of an extensive range of training equipment. This equipment encompasses desktop computer-based training systems, generic training devices. The procurement of a training simulator is clearly dictated by the perceived training requirement or problem. Also, it should preferably involve or follow a detailed training needs analysis. Although the cost benefits of training are often difficult to quantify, a simulator is frequently easier to justify if plant familiarisation and training can be provided in advance of on-the-job experience. This is particularly true if the target operators have little hands-on experience of similar plant either in terms of processes or the operator interface. (author).

  17. Organizing nuclear power plant operation

    International Nuclear Information System (INIS)

    Adams, H.W.; Rekittke, K.

    1987-01-01

    With the preliminary culmination in the convoy plants of the high standard of engineered safeguards in German nuclear power plants developed over the past twenty years, the interest of operators has now increasingly turned to problems which had not been in the focus of attention before. One of these problems is the organization of nuclear power plant operation. In order to enlarge the basis of knowledge, which is documented also in the rules published by the Kerntechnischer Ausschuss (Nuclear Technology Committee), the German Federal Minister of the Interior has commissioned a study of the organizational structures of nuclear power plants. The findings of that study are covered in the article. Two representative nuclear power plants in the Federal Republic of Germany were selected for the study, one of them a single-unit plant run by an independent operating company in the form of a private company under German law (GmbH), the other a dual-unit plant operated as a dependent unit of a utility. The two enterprises have different structures of organization. (orig.) [de

  18. TVA's nuclear power plant experience

    International Nuclear Information System (INIS)

    Willis, W.F.

    1979-01-01

    This paper reviews TVA's nuclear power plant design and construction experience in terms of schedule and capital costs. The completed plant in commercial operation at Browns Ferry and six additional plants currently under construction represent the nation's largest single commitment to nuclear power and an ultimate investment of $12 billion by 1986. The presentation is made in three separate phases. Phase one will recapitulate the status of the nuclear power industry in 1966 and set forth the assumptions used for estimating capital costs and projecting project schedules for the first TVA units. Phase two describes what happened to the program in the hectic early 1979's in terms of expansion of scope (particularly for safety features), the dramatic increase in regulatory requirements, vendor problems, stretchout of project schedules, and unprecedented inflation. Phase three addresses the assumptions used today in estimating schedules and plant costs for the next ten-year period

  19. Prospects for power plant technology

    International Nuclear Information System (INIS)

    Schilling, H.D.

    1993-01-01

    Careful conservation of resources in the enlarged context of the rational utilization of energy, the environment and capital will determine future power plant technology. The mainstays will be the further development of power plant concepts based on fossil (predominantly coal) and nuclear fuels; world-wide, also regenerative and CO 2 -free hydro-electric power will play a role. Rapid conversion of the available potential requires clear, long-term stable and reliable political framework conditions for the release of the necessary entrepreneurial forces. (orig.) [de

  20. Partner of nuclear power plants

    International Nuclear Information System (INIS)

    Gribi, M.; Lauer, F.; Pauli, W.; Ruzek, W.

    1992-01-01

    Sulzer, the Swiss technology group, is a supplier of components and systems for nuclear power plants. Important parts of Swiss nuclear power stations, such as containments, reactor pressure vessels, primary pipings, are made in Winterthur. Sulzer Thermtec AG and some divisions of Sulzer Innotec focus their activities on servicing and backfitting nuclear power plants. The European market enjoys priority. New types of valves or systems are developed as economic solutions meeting more stringent criteria imposed by public authorities or arising from operating conditions. (orig.) [de

  1. Nuclear power plant V-2

    International Nuclear Information System (INIS)

    1998-01-01

    The nuclear power plant Bohunice V -2 is briefly described. This NPP consists from two reactor units. Their main time characteristics are (Reactor Unit 1, Reactor Unit 2): beginning of construction - December 1976; first controlled reactor power - 7 August 1984, 2 August 1985; connection to the grid - 20 August 1984, 9 August 1985; commercial operation - 14 February 1985, 18 December 1985. This leaflet contains: NPP V-2 construction; Major technological equipment [WWER 440 V230 type reactor; Nuclear Power plant operation safety (Safety barriers; Safety systems [Active safety systems, Passive safety systems]); Centralized heat supply system; Scheme of Bohunice V-2 NPP and technical data

  2. Preliminary analysis for u tube degradation in CANDU steam generator using CATHENA

    Energy Technology Data Exchange (ETDEWEB)

    Shin, So Eun; Lee, Jeong Hun; Park, Tong Kyu; Hwang, Su Hyun [FNC Technology Co., Seoul (Korea, Republic of); Jung, Jong Yeo [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    The interest in plant safety and integrity has been increasing due to long term operation of nuclear power plants (NPPs) and lots of efforts have been devoted to developing the degradation evaluation model for all the Structure, System, and Components (SSCs) of NPPs in these days. The efforts, however, were mainly concentrated on pressurized light water reactors (PWRs) in domestic. In contrast, the study for the aging degradation of counterparts of CANDU (CANada Deuterium Uranium) reactors has been rarely performed, even though Wolsong unit 1 (WS1), that is a CANDU 6 NPP in Korea, has been operating for almost 30 years. Therefore, the assessment of the aging degradation is required and the proper and exact evaluation model for the aging degradation of SCCs of CANDU, especially WS1, is urgently needed. In this study, the aging degradation of steam generators (SGs) in WS1 was mainly discussed. Based on cases of the aging degradation of SGs in overseas CANDU reactors, the major potential aging mechanisms of SGs were estimated since there has been no case of accident due to degradation in CANDU NPPs in Korea . Some core parameters which are indicators of the degree of degradation were calculated by CATHENA (Canadian algorithm for thermal hydraulic network analysis). In the result of comparing two calculation cases; core parameters for only aged SGs in fresh plant and those for all the aged component, it can be concluded that aging of SGs is a main component in the degradation assessment of CANDU NPPs, and keeping the integrity of steam generator (SG) tubes is important to guarantee the safety of the NPPs.

  3. Dispatchable Solar Power Plant Project

    Energy Technology Data Exchange (ETDEWEB)

    Price, Henry [Solar Dynamics LLC, Broomfield, CO (United States)

    2018-01-31

    As penetration of intermittent renewable power increases, grid operators must manage greater variability in the supply and demand on the grid. One result is that utilities are planning to build many new natural gas peaking power plants that provide added flexibility needed for grid management. This report discusses the development of a dispatchable solar power (DSP) plant that can be used in place of natural gas peakers. Specifically, a new molten-salt tower (MST) plant has been developed that is designed to allow much more flexible operation than typically considered in concentrating solar power plants. As a result, this plant can provide most of the capacity and ancillary benefits of a conventional natural gas peaker plant but without the carbon emissions. The DSP system presented was designed to meet the specific needs of the Arizona Public Service (APS) utility 2017 peaking capacity request for proposals (RFP). The goal of the effort was to design a MST peaker plant that had the operational capabilities required to meet the peaking requirements of the utility and be cost competitive with the natural gas alternative. The effort also addresses many perceived barriers facing the commercial deployment of MST technology in the US today. These include MST project development issues such as permitting, avian impacts, visual impacts of tower CSP projects, project schedule, and water consumption. The DSP plant design is based on considerable analyses using sophisticated solar system design tools and in-depth preliminary engineering design. The resulting DSP plant design uses a 250 MW steam power cycle, with solar field designed to fit on a square mile plot of land that has a design point thermal rating of 400 MWt. The DSP plant has an annual capacity factor of about 16% tailored to deliver greater than 90% capacity during the critical Arizona summer afternoon peak. The table below compares the All-In energy cost and capacity payment of conventional combustion turbines

  4. Simulation technology for power plants

    International Nuclear Information System (INIS)

    Kuwabara, Kazuo; Yanai, Katsuya.

    1988-01-01

    In the simulation of nuclear power stations, there are the simulation for the training of plant operation, the plant simulation for analyzing the operation of an electric power system, the simulation for controlling a core, the simulation for the safety analysis of reactors, the simulation for the design analysis of plants and so on as the typical ones. The outline and the technical features of these simulations are described. With the increase of capacity and complexity of thermal power plants, recently the automation of operation has advanced rapidly. The chance of starting up and stopping plants by operators themselves is few, and the chance of actually experiencing troubles also is few as the reliability of plants improved. In order to maintain the ability of coping with plant abnormality, an operation supporting system is strongly demanded. Operation training simulators and used widely now, and there are the simulators for analysis, those of replica type, those of versatile compact type and so on. The system configuration, modeling techniques, training function and others of the replica type are explained. In hydroelectric plants, the behavior of water in penstocks, the characteristics of water turbines, the speed control system for water turbines and the characteristics of generators become the main subjects of simulation. These are described. (Kako, I.)

  5. ALARA at nuclear power plants

    International Nuclear Information System (INIS)

    Baum, J.W.

    1991-01-01

    Implementation of the ALARA principle at nuclear power plants presents a continuing challenge for health physicists at utility corporate and plant levels, for plant designers, and for regulatory agencies. The relatively large collective doses at some plants are being addressed through a variety of dose reduction techniques. Initiatives by the ICRP, NCRP, NRC, INPO, EPRI, and BNL ALARA Center have all contributed to a heightened interest and emphasis on dose reduction. The NCRP has formed Scientific Committee 46-9 which is developing a report on ALARA at Nuclear Power Plants. It is planned that this report will include material on historical aspects, management, valuation of dose reduction ($/person-Sv), quantitative and qualitative aspects of optimization, design, operational considerations, and training. The status of this work is summarized in this report

  6. Safety benefits from CANDU reactor replacement - a case study

    International Nuclear Information System (INIS)

    Mottram, R.; Millard, J.W.F.; Purdy, P.

    2011-01-01

    Both total core replacement and core retubing have been used in the CANDU industry. For future plant refurbishments, based on experience both in new construction and in recent refurbishments, the concept of total core replacement has been revisited. This builds on practices for replacement of other large plant equipment like boilers. The Bruce CANDU reactors, with their local shield tanks built around the Calandria and containment closely located around that Calandria Shield Tank Assembly (CSTA), are believed to be good candidates for core replacement. A structured process was used to design a replacement CSTA suitable for Bruce A use. The work started with a study of opportunities for safety enhancements in the core. This progressed into design studies and related design assist safety analysis on the reactor. A key element of the work involved consideration of how verified features from later CANDU designs, and from our new reactor design work, could be tailored to fit this replacement core. The replacement reactor core brings in structural improvements in both calandria and end shield, and safety improvements like the natural circulation enhancing moderator cooling layout and further optimized reactivity layouts to improve shutdown system performance. Bruce Power are currently studying the business implications of this and retube techniques as part of preparation for future refurbishments. The work explained in this paper is in the context of the safety related changes and the work to choose and quantify them. (author)

  7. Safety benefits from CANDU reactor replacement. A case study

    International Nuclear Information System (INIS)

    Mottram, R.; Millard, J.W.F.; Purdy, P.

    2011-01-01

    Both total core replacement and core retubing have been used in the CANDU industry. For future plant refurbishments, based on experience both in new construction and in recent refurbishments, the concept of total core replacement has been revisited. This builds on practices for replacement of other large plant equipment like boilers. The Bruce CANDU reactors, with their local shield tanks built around the Calandria and containment closely located around that Calandria Shield Tank Assembly (CSTA), are believed to be good candidates for core replacement. A structured process was used to design a replacement CSTA suitable for Bruce A use. The work started with a study of opportunities for safety enhancements in the core. This progressed into design studies and related design assist safety analysis on the reactor. A key element of the work involved consideration of how verified features from later CANDU designs, and from our new reactor design work, could be tailored to fit this replacement core. The replacement reactor core brings in structural improvements in both calandria and end shield, and safety improvements like the natural circulation enhancing moderator cooling layout and further optimized reactivity layouts to improve shutdown system performance. Bruce Power are currently studying the business implications of this and retube techniques as part of preparation for future refurbishments. The work explained in this paper is in the context of the safety related changes and the work to choose and quantify them. (author)

  8. LBB in Candu plants

    Energy Technology Data Exchange (ETDEWEB)

    Kozluk, M.J.; Vijay, D.K. [Ontario Hydro Nuclear, Toronto, Ontario (Canada)

    1997-04-01

    Postulated catastrophic rupture of high-energy piping systems is the fundamental criterion used for the safety design basis of both light and heavy water nuclear generating stations. Historically, the criterion has been applied by assuming a nonmechanistic instantaneous double-ended guillotine rupture of the largest diameter pipes inside of containment. Nonmechanistic, meaning that the assumption of an instantaneous guillotine rupture has not been based on stresses in the pipe, failure mechanisms, toughness of the piping material, nor the dynamics of the ruptured pipe ends as they separate. This postulated instantaneous double-ended guillotine rupture of a pipe was a convenient simplifying assumption that resulted in a conservative accident scenario. This conservative accident scenario has now become entrenched as the design basis accident for: containment design, shutdown system design, emergency fuel cooling systems design, and to establish environmental qualification temperature and pressure conditions. The requirement to address dynamic effects associated with the postulated pipe rupture subsequently evolved. The dynamic effects include: potential missiles, pipe whipping, blowdown jets, and thermal-hydraulic transients. Recent advances in fracture mechanics research have demonstrated that certain pipes under specific conditions cannot crack in ways that result in an instantaneous guillotine rupture. Canadian utilities are now using mechanistic fracture mechanics and leak-before-break assessments on a case-by-case basis, in limited applications, to support licensing cases which seek exemption from the need to consider the various dynamic effects associated with postulated instantaneous catastrophic rupture of high-energy piping systems inside and outside of containment.

  9. Economics of CANDU-PHW

    International Nuclear Information System (INIS)

    McConnell, L.G.; Woodhead, L.W.; Fanjoy, G.R.

    1982-03-01

    The CANDU-Pressurized Heavy Water (CANDU-PHW) type of nuclear-electric generating station has been developed jointly by Atomic Energy of Canada Limited and Ontario Hydro. This paper discusses the cost of producing electricity from CANDU, presents actual cost experience of CANDU and coal in Ontario, presents projected CANDU and coal costs in Ontario and compares CANDU and Light Water Reactor cost estimates in Ontario

  10. Economics of CANDU-PHW

    International Nuclear Information System (INIS)

    Jackson, H.A.; Woodhead, L.W.; Fanjoy, G.R.

    1984-03-01

    The CANDU-Pressurized Heavy Water (CANDU-PHW) type of nuclear-electric generating station has been developed jointly by Atomic Energy of Canada Limited and Ontario Hydro. This paper discusses the cost of producing electricity from CANDU, presents actual cost experience of CANDU and coal in Ontario, presents projected CANDU and coal costs in Ontario and compares CANDU and Light Water Reactor cost estimates in Ontario

  11. Economics of CANDU-PHW

    International Nuclear Information System (INIS)

    Jackson, H.A.; Horton, E.P.; Woodhead, L.W.; Fanjoy, G.R.

    1985-03-01

    The CANDU-Pressurized Heavy Water (CANDU-PHW) type of nuclear-electric generating station has been developed jointly by Atomic Energy of Canada Limited and Ontario Hydro. This paper discusses the cost of producing electricity from CANDU, presents actual cost experience of CANDU and coal in Ontario, presents projected CANDU and coal costs in Ontario and compares CANDU and Light Water Reactor cost estimates in Ontario

  12. Ocean power plants

    Energy Technology Data Exchange (ETDEWEB)

    Mazurkiedicz, B.; Sliwa, B.

    1982-01-01

    Plans are examined for OTES of close and open cycles. Examples of design of TO are presented. Main design elements of the OTES are indicated and their arrangement. The OTES can be realized even now with comparatively small capital investments. Searches are made for solutions which would make it possible to construct the OTES and not in tropical regions, i.e., with very small temperature differences. The studies indicated that with a difference of temperatures 4.5/sup 0/C and temperature of the thermal water 5.5/sup 0/C, it is possible to build OTES with power 100 MW. With difference in temperature 5/sup 0/C, the power will reach 130 MW.

  13. Submarine nuclear power plant

    International Nuclear Information System (INIS)

    Enohara, Masami; Araragi, Fujio.

    1980-01-01

    Purpose: To provide a ballast tank, and nuclear power facilities within the containment shell of a pressure resistance structure and a maintenance operator's entrance and a transmission cable cut-off device at the outer part of the containment shell, whereby after the construction, the shell is towed, and installed by self-submerging, and it can be refloated for repairs by its own strength. Constitution: Within a containment shell having a ballast tank and a pressure resisting structure, there are provided nuclear power facilities including a nuclear power generating chamber, a maintenance operator's living room and the like. Furthermore, a maintenance operator's entrance and exit device and a transmission cable cut-off device are provided within the shell, whereby when it is towed to a predetermined a area after the construction, it submerges by its own strength and when any repair inspection is necessary, it can float up by its own strength, and can be towed to a repair dock or the like. (Yoshihara, H.)

  14. Future fuel cycle development for CANDU reactors

    International Nuclear Information System (INIS)

    Hatcher, S.R.; McDonnell, F.N.; Griffiths, J.; Boczar, P.G.

    1987-01-01

    The CANDU reactor has proven to be safe and economical and has demonstrated outstanding performance with natural uranium fuel. The use of on-power fuelling, coupled with excellent neutron economy, leads to a very flexible reactor system with can utilize a wide variety of fuels. The spectrum of fuel cycles ranges from natural uranium, through slightly enriched uranium, to plutonium and ultimately thorium fuels which offer many of the advantages of the fast breeder reactor system. CANDU can also burn the recycled uranium and/or the plutonium from fuel discharged from light water reactors. This synergistic relationship could obviate the need to re-enrich the reprocessed uranium and allow a simpler reprocessing scheme. Fule management strategies that will permit future fuel cycles to be used in existing CANDU reactors have been identified. Evolutionary design changes will lead to an even greater flexibility, which will guarantee the continued success of the CANDU system. (author)

  15. Mathematical modeling of CANDU-PHWR

    Energy Technology Data Exchange (ETDEWEB)

    Gaber, F.A.; Aly, R.A.; El-Shal, A.O. [Atomic Energy Authority, Cairo (Egypt)

    2003-07-01

    The paper deals with the transient studies of CANDU 600 pressurized Heavy Water Reactor (PHWR). This study involved mathematical modeling of CANDU-PHWR to study its thermodynamic performances. Modeling of CANDU-PHWR was based on lumped parameter technique. The reactor model includes the neutronic, reactivity, and fuel channel heat transfer. The nuclear reactor power was modelled using the point kinetics equations with six groups of delayed neutrons and the reactivity feed back due to the changes in the fuel temperature and coolant temperature. The CANDU-PHWR model was coded in FORTRAN language and solved by using a standard numerical technique. The adequacy of the model was tested by assessing the physical plausibility of the obtained results. (author)

  16. Assessment of DUPIC fuel compatibility with CANDU-6

    Energy Technology Data Exchange (ETDEWEB)

    Choi, H B; Roh, G H; Jeong, C J; Rhee, B W; Choi, J W [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-07-01

    The compatibility of DUPIC fuel with the existing CANDU reactor was assessed. The technical issues of DUPIC fuel compatibility were chosen based on the CANDU physics design requirements and inherent characteristics of DUPIC fuel. The compatibility was assessed for the reference DUPIC fuel composition which was determined to reduce the composition heterogeneity and improve the spent PWR fuel utilization. Preliminary studies on a CANDU core loaded with DUPIC fuel have shown that the nominal power distribution is flatter than that of a natural uranium core when a 2-bundle shift refueling scheme is used, which reduces the reactivity worths of devices in the core and, therefore, the performance of reactivity devices was assessed. The safety of the core was assessed by a LOCA simulation and it was found that the power pulse upon LOCA can be maintained below that in the natural uranium core when a poison material is used in the DUPIC fuel. For the feasibility of handling DUPIC fuel in the plant, it will be necessary to introduce new equipment to load the DUPIC fuel in the refueling magazine. The radiation effect of DUPIC fuel on both the reactor hardware and the environment will require a quantitative analysis later. (author).

  17. Preliminary assessment on compatibility of DUPIC fuel with CANDU-6

    International Nuclear Information System (INIS)

    Choi, Hang-Bok; Roh, G.H.; Jeong, C.J.; Rhee, B.W.; Choi, J.W.; Boss, C.R.

    1997-01-01

    The compatibility of DUPIC fuel with the existing CANDU-6 reactor was assessed. The technical issues of DUPIC fuel compatibility were chosen based on the CANDU physics design requirements and inherent characteristics of DUPIC fuel. The compatibility was assessed for the reference DUPIC fuel composition which was determined to reduce the composition heterogeneity and improve the spent PWR fuel utilization. Preliminary studies on a CANDU core loaded with DUPIC fuel have shown that the nominal power distribution is flatter than that of a natural uranium core when a 2-bundle shift refueling scheme is used, which reduces the reactivity worths of devices in the core and, therefore, the performance of reactivity devices was assessed. The safety of the core was assessed by a LOCA simulation and it was found that the power pulse upon LOCA can be maintained below that in the natural uranium core when a poison material is used in the DUPIC fuel. For the feasibility of handling DUPIC fuel in the plant, it will be necessary to introduce new equipment to load the DUPIC fuel in the refueling magazine. The radiation effect of DUPIC fuel on both the reactor hardware and the environment will be qualitatively analyzed later. (author)

  18. Chemistry in power plants 2011

    International Nuclear Information System (INIS)

    2011-01-01

    Within the VGB Powertech conference from 25th to 27th October, 2011, in Munich (Federal Republic of Germany), the following lectures and poster contributions were presented: (1) The revised VGB standard for water-steam-cycle Chemistry; (2) Switchover from neutral operation to oxygen treatment at the power station Stuttgart-Muenster of EnBW Kraftwerke AG; (3) Steam contamination with degradation products of organic matters present in the feedwater of the Lanxess-Rubber cogeneration plant; (4) Laboratory scale on-line noble metal deposition experiments simulating BWR plant conditions; (5) Building a new demin installation for the power plant EPZ in Borssele; (6) Replacement of the cooling tower installations in the nuclear power plant Goesgen-Daenien AG; (7) Aging of IEX resins in demin plants - Cost optimisation by adaptation of regenerants; (8) The largest DOW trademark EDI System at a combined cycled plant in Europe; (9) Upgrading river Main water to boiler feed water - Experiences with ultrafiltration; (10) Experiences with treatment of the water-steam-cycle in the RDF power plant Nehlsen Stavenhagen with film-forming amines; (11) Comparative modelling of the bubbles thermal collapse and cavitations for estimation of bubbles collapse influence; (12) Overcoming the steam quality - issues from an HRSG for the production of process steam; (13) Legionella - new requirements for power plant operation; (14) How the right chemistry in the FGD helps to improve the removal in the waste water treatment plant; (15) High efficiency filtration in dry/semi-dry FGD plants; (16) Expanding the variety of renewable fuels in the biomass power plant Timelkam using the chemical input control; (17) Corrosion, operating experiences and process improvements to increase the availability and operating time of the biomass power plant Timelkam; (18) The influence of temperature on the measurement of the conductivity of highly diluted solutions; (19) A multiparameter instrumentation approach

  19. CANDU steam generator life management

    International Nuclear Information System (INIS)

    Tapping, R.L.; Nickerson, J.; Spekkens, P.; Maruska, C.

    1998-01-01

    Steam generators are a critical component of a nuclear power reactor, and can contribute significantly to station unavailability, as has been amply demonstrated in Pressurized Water Reactors (PWRs). CANDU steam generators are not immune to steam generator degradation, and the variety of CANDU steam generator designs and tube materials has led to some unexpected challenges. However, aggressive remedial actions, and careful proactive maintenance activities, have led to a decrease in steam generator-related station unavailability of Canadian CANDUs. AECL and the CANDU utilities have defined programs that will enable existing or new steam generators to operate effectively for 40 years. Research and development work covers corrosion and mechanical degradation of tube bundles and internals, chemistry, thermalhydraulics, fouling, inspection and cleaning, as well as provision for specially tool development for specific problem solving. A major driving force is development of CANDU-specific fitness-for-service guidelines, including appropriate inspection and monitoring technology to measure steam generator condition. Longer-range work focuses on development of intelligent on-line monitoring for the feedwater system and steam generator. New designs have reduced risk of corrosion and fouling, are more easily inspected and cleaned, and are less susceptible to mechanical damage. The Canadian CANDU utilities have developed programs for remedial actions to combat degradation of performance (Gentilly-2, Point Lepreau, Bruce A/B, Pickering A/B), and have developed strategic plans to ensure that good future operation is ensured. The research and development program, as well as operating experience, has identified where improvements in operating practices and/or designs can be made in order to ensure steam generator design life at an acceptable capacity factory. (author)

  20. Financing Solar Thermal Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Kistner, Rainer [National Renewable Energy Lab. (NREL), Golden, CO (United States); Price, Henry W. [National Renewable Energy Lab. (NREL), Golden, CO (United States)

    1999-04-14

    The commercialization of concentrating solar power technology took a major step forward in the mid 1980s and early 1990s with the development of the SEGS plants in California. Over the years they have proven that parabolic trough power technologies are the most cost-effective approach for commercial scale solar power generation in the sunbelt countries of the world. However, the question must be asked why no additional solar power plants have been build following the bankruptcy of the developer of the SEGS projects, LUZ International Limited. Although many believe the SEGS projects were a success as a result of parabolic trough technology they employ, in truth, the SEGS projects were developed simply because they represented an attractive opportunity for investors. Simply stated, no additional projects have been developed because no one has been able to put together a similarly attractive financial package to potential investors. More than $1.2 billion in private capital was raised in debt and equity financing for the nine SEGS plants. Investors and bankers who make these investments are the real clients for solar power technologies. They are not interested in annual solar to electric efficiencies, but in risk, return on investments, and coverage ratios. This paper will take a look at solar power projects from the financier’s perspective. The challenge in moving forward is to attract private investors, commercial lenders, and international development agencies and to find innovative solutions to the difficult issues that investment in the global power market poses for solar power technologies.

  1. Financing Solar Thermal Power Plants

    International Nuclear Information System (INIS)

    Price, Henry W.; Kistner, Rainer

    1999-01-01

    The commercialization of concentrating solar power technology took a major step forward in the mid 1980s and early 1990s with the development of the SEGS plants in California. Over the years they have proven that parabolic trough power technologies are the most cost-effective approach for commercial scale solar power generation in the sunbelt countries of the world. However, the question must be asked why no additional solar power plants have been build following the bankruptcy of the developer of the SEGS projects, LUZ International Limited. Although many believe the SEGS projects were a success as a result of parabolic trough technology they employ, in truth, the SEGS projects were developed simply because they represented an attractive opportunity for investors. Simply stated, no additional projects have been developed because no one has been able to put together a similarly attractive financial package to potential investors. More than $1.2 billion in private capital was raised in debt and equity financing for the nine SEGS plants. Investors and bankers who make these investments are the real clients for solar power technologies. They are not interested in annual solar to electric efficiencies, but in risk, return on investments, and coverage ratios. This paper will take a look at solar power projects from the financier's perspective. The challenge in moving forward is to attract private investors, commercial lenders, and international development agencies and to find innovative solutions to the difficult issues that investment in the global power market poses for solar power technologies

  2. Financing solar thermal power plants

    International Nuclear Information System (INIS)

    Kistner, R.; Price, H.

    1999-01-01

    The commercialization of concentrating solar power technology took a major step forward in the mid 1980s and early 1990s with the development of the SEGS plants in California. Over the years they have proven that parabolic trough power technologies are the most cost-effective approach for commercial scale solar power generation in the sunbelt countries of the world. However, the question must be asked why no additional solar power plants have been built following the bankruptcy of the developer of the SEGS projects, LUZ International Limited. Although many believe the SEGS projects were a success as a result of parabolic trough technology they employ, in truth, the SEGS projects were developed simply because they represented an attractive opportunity for investors. Simply states, no additional projects have been developed because no one has been able to put together a similarly attractive financial package to potential investors. More than $1.2 billion in private capital was raised in debt and equity financing for the nine SEGS plants. Investors and bankers who make these investments are the real clients for solar power technologies. They are not interested in annual solar to electric efficiencies, but in risk, return on investments, and coverage ratios. This paper will take a look at solar power projects form the financier's perspective. The challenge in moving forward is to attract private investors, commercial lenders, and international development agencies and to find innovative solutions to the difficult issues that investment in the global power market poses for solar power technologies

  3. Thermal Power Plant Performance Analysis

    CERN Document Server

    2012-01-01

    The analysis of the reliability and availability of power plants is frequently based on simple indexes that do not take into account the criticality of some failures used for availability analysis. This criticality should be evaluated based on concepts of reliability which consider the effect of a component failure on the performance of the entire plant. System reliability analysis tools provide a root-cause analysis leading to the improvement of the plant maintenance plan.   Taking in view that the power plant performance can be evaluated not only based on  thermodynamic related indexes, such as heat-rate, Thermal Power Plant Performance Analysis focuses on the presentation of reliability-based tools used to define performance of complex systems and introduces the basic concepts of reliability, maintainability and risk analysis aiming at their application as tools for power plant performance improvement, including: ·         selection of critical equipment and components, ·         defini...

  4. Types and properties of elastomer materials used in CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    You, Ho Sik; Jeong, Jin Kon [Korea Atomic Energy Research Institute, Daeduk (Korea, Republic of)

    1996-05-01

    Properties and kinds of elastomer materials used in a CANDU power plant have been described. The elastomer materials have been used as a sealing material in the components f nuclear power plant since they have many excellent properties that can not be seen in other materials. It is very important to select proper elastomer materials used in the nuclear power plant are required to have resistance to temperature as well as radiation. According to the experimental results performed at some laboratories including the Chalk River Laboratory of AECL, elastomer materials with high resistance to temperature and radiation are Nitrile, Ethylene, Propylene and Butyl. These materials have been used in a lot of components of Wolsong unit 1 and Wolsong 2, 3 and 4 which are under elastomer material. Therefore, the studies on the standardization are currently under way to limit about 10 different kinds of elastomer materials to be used in the plant. 16 tabs., 1 fig., 12 refs. (Author) .new.

  5. Nuclear power plant diagnostics

    International Nuclear Information System (INIS)

    Hollo, E.; Siklossy, P.

    1982-01-01

    The cooling circuit vibration diagnostic system of the Block 1 of the Paks nuclear power station is described. The automatic online vibration monitoring system consisting presently of 42 acceleration sensors and 9 pressure fluctuation sensors, which could be extended, performs both global and local inspection of the primary cooling circuit and its components. The offline data processing system evaluates the data for failure mode analysis. The software under development will be appropriate for partial preliminary identification of failure reasons during their initial phases. The installation experiences and the preliminary results during the hot operational testing of Block 1 are presented. (Sz.J.)

  6. Nuclear power plants

    International Nuclear Information System (INIS)

    Usui, Eizo.

    1981-01-01

    Purpose: To prevent boiling of saturated water in the drain tank of a humidity separator by charging cooling water in the drain tank upon power decrease of a turbine. Constitution: Saturated water is separated from high pressure turbine exhausts in a humidity separator and stored in a drain tank. The saturated water in the drain tank is controlled to a constant level and the excess water is sent to a condensator and a feedwater heater. A cooling water feed pipe is branched as a cooling water feed pipe from the exhaust side of a reactor feedwater pump and connected by way of a closing valve to a spray nozzle provided in the drain tank. While the closing valve is usually closed to keep the water level constant in the drain tank, the closing valve is opened upon sudden decrease in the turbine power to charge the condensates by way of the cooling water feed pipe to the drain tank. Thus, the saturated water is mixed with the dondensates and the temperature is lowered to prevent boiling of the saturated water. (Kawakami, Y.)

  7. Successful completion of the Qinshan phase III nuclear power plant-a successful model for Chinese-Canadian cooperation

    International Nuclear Information System (INIS)

    Peng Xiaoxing

    2004-01-01

    This report documents Qinshan CANDU project construction and commissioning experience as well as management strategies and approaches that contributed to the successful completion of the project. The Qinshan phase III (CANDU) nuclear power plant was built in record times: Unit 1 achieved commercial operation on December 31, 2002 and Unit 2 on July 24, 2003, 43 days and 112 days ahead of schedule respectively. The reference plant design is the Wolsong 3 and 4 CANDU-6 units in the Republic of Korea. Improvements in design and construction methods allowed Unit 1 to be constructed in 51.5 Months from First Concrete to Criticality-a record in China for nuclear power plants. The key factors are project management and project management tools, quality assurance, construction methods (including open top construction, heavy lifts and modularization), electronic documentation with configuration control that provides up-to-date on-line information, CADDS design linked with material management, specialized material control including bar coding, and planning. The introduction of new design and construction techniques was achieved by combining conventional AECL practices with working experiences in China. The most advanced tools and techniques for achieving optimum construction quality, schedule and cost were used. Successful application of advanced project management methods and tools will benefit TQNPC in operation of the station, and the Chinese contractors in advancing their capabilities in future nuclear projects in China and enhancing their opportunities internationally. TQNPC's participation in Quality surveillance (QS) activities of nuclear steam plant (NSP) and Balance of Plant (BOP) offshore equipment benefited TQNPC in acquiring knowledge of specific equipment manufacturing processes, which can be applied to similar activities in China. China has established the capability of manufacturing CANDU fuel and becoming self-reliant in fuel supply. Excellent co-operation and

  8. Subchannel analysis code development for CANDU fuel channel

    International Nuclear Information System (INIS)

    Park, J. H.; Suk, H. C.; Jun, J. S.; Oh, D. J.; Hwang, D. H.; Yoo, Y. J.

    1998-07-01

    Since there are several subchannel codes such as COBRA and TORC codes for a PWR fuel channel but not for a CANDU fuel channel in our country, the subchannel analysis code for a CANDU fuel channel was developed for the prediction of flow conditions on the subchannels, for the accurate assessment of the thermal margin, the effect of appendages, and radial/axial power profile of fuel bundles on flow conditions and CHF and so on. In order to develop the subchannel analysis code for a CANDU fuel channel, subchannel analysis methodology and its applicability/pertinence for a fuel channel were reviewed from the CANDU fuel channel point of view. Several thermalhydraulic and numerical models for the subchannel analysis on a CANDU fuel channel were developed. The experimental data of the CANDU fuel channel were collected, analyzed and used for validation of a subchannel analysis code developed in this work. (author). 11 refs., 3 tabs., 50 figs

  9. Ocean power plants

    Energy Technology Data Exchange (ETDEWEB)

    Dembicki, E.

    1982-01-01

    In the fall of 1980 on the shores of the Hawaiian Islands, a floating laboratory of the United States was successfully introduced for testing a heat exchanger and pipes for collecting cold water of the OTES with power of 1 MW. The first American OTES N=10-40 MW should start operation in 1985. By the year 2000, ..sigma..N of the U.S. OTES should reach 10 GW. The Japanese OTES N=10-25 MW should start up in 1989. The experimental OTES N=100 KW has been in operation since October 1981 on the Nauru Island. An OTES of 2 MW is under construction. The concern Empain-Schneider is involved in planning the OTES of closed cycle in France, and the concern CGE is planning the OTES of open cycle.

  10. Maintenance of nuclear power plants

    International Nuclear Information System (INIS)

    Lashgari, Farbod.

    1995-01-01

    This paper is about maintenance of nuclear power plants. In part one, the outage management of nuclear power plants has described. Meaning of the outage and objectives of outage management is given in introduction. The necessity of a long-term outage strategy is shown in chapter one. The main parts of an outage are as follows: Planning; Preparation; Execution, Each of them and also post-outage review have been explained in the followed chapters. Part two deals with technical details of main primary components of nuclear power plant type WWER. After an introduction about WWER reactors, in each chapter first the general and detailed description of main primary components has given and then their maintenance schedules and procedures. Chapter about reactor and steam generator is related to both types of WWER-440 and WWER-1000, but chapter about reactor coolant pump has specified to WWER-1000 to be more in details.(author)

  11. Toxic releases from power plants

    International Nuclear Information System (INIS)

    Rubin, E.S.

    1999-01-01

    Beginning in 1998, electric power plants burning coal or oil must estimate and report their annual releases of toxic chemicals listed in the Toxics Release Inventory (TRI) published by the US Environmental Protection Agency (EPA). This paper identifies the toxic chemicals of greatest significance for the electric utility sector and develops quantitative estimates of the toxic releases reportable to the TRI for a representative coal-fired power plant. Key factors affecting the magnitude and types of toxic releases for individual power plants also are discussed. A national projection suggests that the magnitude of electric utility industry releases will surpass those of the manufacturing industries which current report to the TRI. Risk communication activities at the community level will be essential to interpret and provide context for the new TRI results

  12. Thermal power plants and environment

    International Nuclear Information System (INIS)

    1997-01-01

    Recent versions of the air quality models which are reviewed and approved from the Environmental Protection Agency (EPA) are analysed in favour of their application in simple and complex terrain, different meteorological conditions and modifications in the sources of pollutant emissions. Improvement of the standard methods for analysis of the risks affecting the environment from different energy sources has been carried out. The application of the newly introduced model enabled (lead to performing) risk analysis of the coal power plants compared to other types of energy sources. Detailed investigation of the risk assessment and perception from coal power plants, has been performed and applied to the Macedonian coal power plants. Introducing the concept of 'psychological pollution', a modification of the standard models and programs for risk assessment from various energy sources has been suggested (proposed). The model has been applied to REK Bitola, where statistically relevant differences in relation to the control groups have been obtained. (Original)

  13. Reliability Characteristics of Power Plants

    Directory of Open Access Journals (Sweden)

    Zbynek Martinek

    2017-01-01

    Full Text Available This paper describes the phenomenon of reliability of power plants. It gives an explanation of the terms connected with this topic as their proper understanding is important for understanding the relations and equations which model the possible real situations. The reliability phenomenon is analysed using both the exponential distribution and the Weibull distribution. The results of our analysis are specific equations giving information about the characteristics of the power plants, the mean time of operations and the probability of failure-free operation. Equations solved for the Weibull distribution respect the failures as well as the actual operating hours. Thanks to our results, we are able to create a model of dynamic reliability for prediction of future states. It can be useful for improving the current situation of the unit as well as for creating the optimal plan of maintenance and thus have an impact on the overall economics of the operation of these power plants.

  14. Robotics for nuclear power plants

    International Nuclear Information System (INIS)

    Nakayama, Ryoichi; Kimura, Motohiko; Abe, Akira

    1993-01-01

    A continuing need exists for automatic or remote-controlled machines or robots which can perform inspection and maintenance tasks in nuclear power plants. Toshiba has developed several types of monofunctional and multi- functional robots for such purposes over the past 20 years, some of which have already been used in actual plants. This paper describes new multifunctional robots for inspection and maintenance. An inspection robot has been applied in an actual plant for two years for performance testing. Maintenance robots for grinding tasks have also been developed, which can be easily teleoperated by the operator using automatic control. These new robots are expected to be applied to actual inspection and maintenance work in nuclear power plants. (author)

  15. The security management of spent filter cartridge in Qinshan phase 3 (heavy water reactor) nuclear power plant

    International Nuclear Information System (INIS)

    Xue Dahai

    2005-01-01

    Qinshan phase 3 nuclear power plant is the first CANDU plant that China fetched in from Canada, and both two units operate under well condition up to now. The radioactive wastes produced during the unit operation mainly include technical waste, spent resin, and spent filter cartridge. The spent filter cartridge is one important part both in the volume and radioactivity of the radioactive waste, and it is the important content of radioactive waste management. Different from PWR, part of high radioactive spent filter in CANDU unit comes from heavy water system such as moderator system. It has to be dried through blowing before replaced from the system. But this working procedure result the filtrate dreg become flexible, and it can bring on the risk of internal or external exposure. It is very important to pay high attention to control the contamination spread during spent filter inside transfer. (authors)

  16. Nuclear power plant

    International Nuclear Information System (INIS)

    Inami, Ichiro; Kobayashi, Minoru.

    1995-01-01

    In a condensate cleanup system and a reactor water cleanup system of a BWR-type reactor, in which primary coolants flow, there is disposed a filtering and desalting device using hollow thread membrane filter and ion exchange resin for a condensate cleanup system, and using a high temperature filter made of a metal, a metal oxide or ceramics as a filtering material and a precoat filter made of a powdery ion exchange resin as a filtering material for a reactor water cleanup system. This can completely remove cruds generated in the condensate system. Since the reactor water cleanup system comprises the powdery resin precoat-type filtering and desalting device and the high temperature filter using ceramics, ionic impurities such as radioactive materials can be removed. Accordingly, cruds are not carried into the inside of the reactor, and since the radioactive concentration in the reactor water is reduced, radiation exposure upon periodical inspection can be minimized almost to zero, to attain a clean plant. (T.M.)

  17. R and M considerations in the selection of class 3 emergency power system for Candu-type nuclear power station

    Energy Technology Data Exchange (ETDEWEB)

    Hashmi, M F; Eng, P [Canatom LTD, Montreal, P.Q. (Canada)

    1975-07-01

    Reliability of emergency power system is an important consideration in the design of Electrical Power Supplies for any power station. Some other factors enter in the picture depending on the requirements for the safe shutdown of the nuclear reactor. A Reliability and Maintainability (R and M) comparison is drawn between diesel engines and gas turbines to make a choice of the prime mover for the generator. The type and amount of redundancy of the generator sets is then investigated to establish high reliability. Effects of automatic interconnections between class III buses, generator groupings and synchronization is taken into account. Next, failure modes of the systems are considered and methods are sought to reduce the significant failure modes. Recommendations are made to improve the system at design and specification stage. Economics and maintainability are given due consideration throughout the unit selection, system analysis and improvement. Two-100 pc generators are considered against four-50 pc generators; in both cases, the generators form two independent groups of 100 pc capability.

  18. CANDU severe accident analysis

    International Nuclear Information System (INIS)

    Negut, Gheorghe; Catana, Alexandru; Prisecaru, Ilie; Dupleac, Daniel

    2007-01-01

    Romania is a EU member since January first 2007. This country faces now new challenges which imply also the nuclear power reactors now in operation. Romania operates since 1996 a CANDU nuclear power reactor and soon will start up a second unit. In EU PWR reactors are mostly operated, so that the Romania's reactors have to meet EU standards. Safety analysis guidelines require to model severe accidents for reactors of this type. Starting from previous studies a thermal-hydraulic model for a degraded CANDU core was developed. The initiating event is assumed to be a LOCA with simultaneous loss of moderator and coolant and the failure of emergency core cooling system (ECCS). This type of accident is likely to modify the reactor geometry and will lead to a severe accident development. When the coolant temperatures inside a pressure tube reaches 1000 deg. C, a contact between pressure tube and calandria tube occurs and the decay heat is transferred to the moderator. Due to the lack of cooling, the moderator eventually begins to boil and is expelled, through the calandria vessel relief ducts, into the containment. Therefore the calandria tubes (fuel channels) uncover, then disintegrate and fall down to the calandria vessel bottom. All the quantity of calandria moderator is vaporized and expelled, the debris will heat up and eventually boil. The heat accumulated in the molten debris will be transferred through the calandria vessel wall to the shield water tank surrounding the calandria vessel. The thermal hydraulics phenomena described above are modeled, analyzed and compared with the existing data. (authors)

  19. VGB congress 'Power Plants 2009'

    International Nuclear Information System (INIS)

    Anon.

    2009-01-01

    The VGB Congress 'Power Plants 2009' took place in Lyon/France from 23rd to 25th September 2009 and was themed 'Addressing Climate Change - Winning Public Acceptance through Advanced Technologies'. Nearly 1,300 participants attended the plenary and technical lectures and had the opportunity to discus the current topics of electricity and heat generation. The study carried out by VGB according to which EU-27 requires about 475.000 MW of new power plant capacity was also presented. Specific papers were addressing further topics. The Congress was rounded off by a side-programme and technical visits. (orig.)

  20. CANDU - Canadian experience and expectations with the heavy-water reactor

    International Nuclear Information System (INIS)

    Foster, J.S.; Russell, S.H.

    1977-05-01

    The paper describes the evolution of the CANDU nuclear-power plants with particular reference to the objectives of safety, reliability and economy; the development of industrial capacity for the supply of fuel, components and heavy water; and the prospective development of advanced fuel cycles and the projected results. It provides data on radiation, releases, and exposures, internal and external to the power plants; plant availability, capacity factors and other performance data; heavy water production data with reference to safety, reliability, and economics; projections of the performance of CANDU reactors operating on a thorium-U-233 cycle and the development required to establish this cycle; and intent with respct to spent-fuel management and radioactive-waste storage. (author)

  1. Loviisa nuclear power plant analyzer

    International Nuclear Information System (INIS)

    Porkholm, K.; Nurmilaukas, P.; Tiihonen, O.; Haenninen, M.; Puska, E.

    1992-12-01

    The APROS Simulation Environment has been developed since 1986 by Imatran Voima Oy (IVO) and the Technical Research Centre of Finland (VTT). It provides tools, solution algorithms and process components for use in different simulation systems for design, analysis and training purposes. One of its main nuclear applications is the Loviisa Nuclear Power Plant Analyzer (LPA). The Loviisa Plant Analyzer includes all the important plant components both in the primary and in the secondary circuits. In addition, all the main control systems, the protection system and the high voltage electrical systems are included. (orig.)

  2. Operation of nuclear power plants

    International Nuclear Information System (INIS)

    Severa, P.

    1988-04-01

    The textbook for training nuclear power plant personnel is centred on the most important aspects of operating modes of WWER-440 reactors. Attention is devoted to the steady state operation of the unit, shutdown, overhaul with refuelling, physical and power start-up. Also given are the regulations of shift operation and the duties of individual categories of personnel during the shift and during the change of shifts. (Z.M.). 3 figs., 1 tab

  3. Construction of new nuclear power plant in Bulgaria - incentives and problems

    International Nuclear Information System (INIS)

    Popov, M.; Bataklieva, L.; Hinovski, I.

    2002-01-01

    Advantages and disadvantages of constriction of new nuclear power plant in Bulgaria are discussed. Several options for further development are considered: Completion of the Belene NPP; Evolutionary design installations PWR technology, 600 MW (AP-600 Westinghouse design; B-407 Russian design); Standard design of Framatome, 900 MW; Standard design - heavy water technology CANDU-6 (700 MW) and other.The results of the preliminary economic analyses show that the first two options are approximately equal, but the all options are acceptable from investment point of view, considering an average macro economic environment. The need of assessment of all aspects (technical, economical, financial, social, environmental, national infrastructure) of nuclear sector development is outlined

  4. CANDU fuel cycle economic efficiency assessments using the IAEA-MESSAGE-V code

    International Nuclear Information System (INIS)

    Prodea, Iosif; Margeanu, Cristina Alice; Aioanei, Corina; Prisecaru, Ilie; Danila, Nicolae

    2007-01-01

    The main goal of the paper is to evaluate different electricity generation costs in a CANDU Nuclear Power Plant (NPP) using different nuclear fuel cycles. The IAEA-MESSAGE code (Model for Energy Supply Strategy Alternatives and their General Environmental Impacts) will be used to accomplish these assessments. This complex tool was supplied by International Atomic Energy Agency (IAEA) in 2002 at 'IAEA-Regional Training Course on Development and Evaluation of Alternative Energy Strategies in Support of Sustainable Development' held in Institute for Nuclear Research Pitesti. It is worthy to remind that the sustainable development requires satisfying the energy demand of present generations without compromising the possibility of future generations to meet their own needs. Based on the latest public information in the next 10-15 years four CANDU-6 based NPP could be in operation in Romania. Two of them will have some enhancements not clearly specified, yet. Therefore we consider being necessary to investigate possibility to enhance the economic efficiency of existing in-service CANDU-6 power reactors. The MESSAGE program can satisfy these requirements if appropriate input models will be built. As it is mentioned in the dedicated issues, a major inherent feature of CANDU is its fuel cycle flexibility. Keeping this in mind, some proposed CANDU fuel cycles will be analyzed in the paper: Natural Uranium (NU), Slightly Enriched Uranium (SEU), Recovered Uranium (RU) with and without reprocessing. Finally, based on optimization of the MESSAGE objective function an economic hierarchy of CANDU fuel cycles will be proposed. The authors used mainly public information on different costs required by analysis. (authors)

  5. Building of nuclear power plant

    International Nuclear Information System (INIS)

    Saito, Takashi.

    1997-01-01

    A first nuclear plant and a second nuclear power plant are disposed in adjacent with each other in a building for a nuclear reactor. A reactor container is disposed in each of the plants, and each reactor container is surrounded by a second containing facility. A repairing chamber capable of communicating with the secondary containing facilities for both of the secondary containing facilities is disposed being in contact with the second containing facility of each plant for repairing control rod driving mechanisms or reactor incorporated-type recycling pumps. Namely, the repairing chamber is in adjacent with the reactor containers of both plants, and situated between both of the plants as a repairing chamber to be used in common for both plants. Air tight inlet/exit doors are formed to the inlets/exits of both plants of the repairing chamber. Space for the repairing chamber can be reduced to about one half compared with a case where the repairing chamber is formed independently on each plant. (I.N.)

  6. CANDU safety under severe accidents

    International Nuclear Information System (INIS)

    Snell, V.G.; Howieson, J.Q.; Alikhan, S.; Frescura, G.M.; King, F.; Rogers, J.T.; Tamm, H.

    1996-01-01

    The characteristics of the CANDU reactor relevant to severe accidents are set first by the inherent properties of the design, and second by the Canadian safety/licensing approach. The pressure-tube concept allows the separate, low-pressure, heavy-water moderator to act as a backup heat sink even if there is no water in the fuel channels. Should this also fail, the calandria shell itself can contain the debris, with heat being transferred to the water-filled shield tank around the core. Should the severe core damage sequence progress further, the shield tank and the concrete reactor vault significantly delay the challenge to containment. Furthermore, should core melt lead to containment overpressure, the containment behaviour is such that leaks through the concrete containment wall reduce the possibility of catastrophic structural failure. The Canadian licensing philosophy requires that each accident, together with failure of each safety system in turn, be assessed (and specified dose limits met) as part of the design and licensing basis. In response, designers have provided CANDUs with two independent dedicated shutdown systems, and the likelihood of Anticipated Transients Without Scram is negligible. Probabilistic safety assessment studies have been performed on operating CANDU plants, and on the 4 x 880 MW(e) Darlington station now under construction; furthermore a scoping risk assessment has been done for a CANDU 600 plant. They indicate that the summed severe core damage frequency is of the order of 5 x 10 -6 /year. 95 refs, 3 tabs

  7. CANDU safety under severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Snell, V G; Howieson, J Q [Atomic Energy of Canada Ltd. (Canada); Alikhan, S [New Brunswick Electric Power Commission (Canada); Frescura, G M; King, F [Ontario Hydro (Canada); Rogers, J T [Carleton Univ., Ottawa, ON (Canada); Tamm, H [Atomic Energy of Canada Ltd. (Canada). Whiteshell Research Lab.

    1996-12-01

    The characteristics of the CANDU reactor relevant to severe accidents are set first by the inherent properties of the design, and second by the Canadian safety/licensing approach. The pressure-tube concept allows the separate, low-pressure, heavy-water moderator to act as a backup heat sink even if there is no water in the fuel channels. Should this also fail, the calandria shell itself can contain the debris, with heat being transferred to the water-filled shield tank around the core. Should the severe core damage sequence progress further, the shield tank and the concrete reactor vault significantly delay the challenge to containment. Furthermore, should core melt lead to containment overpressure, the containment behaviour is such that leaks through the concrete containment wall reduce the possibility of catastrophic structural failure. The Canadian licensing philosophy requires that each accident, together with failure of each safety system in turn, be assessed (and specified dose limits met) as part of the design and licensing basis. In response, designers have provided CANDUs with two independent dedicated shutdown systems, and the likelihood of Anticipated Transients Without Scram is negligible. Probabilistic safety assessment studies have been performed on operating CANDU plants, and on the 4 x 880 MW(e) Darlington station now under construction; furthermore a scoping risk assessment has been done for a CANDU 600 plant. They indicate that the summed severe core damage frequency is of the order of 5 x 10{sup -6}/year. 95 refs, 3 tabs.

  8. Off-shore nuclear power plant

    International Nuclear Information System (INIS)

    Nakanishi, T.

    1980-01-01

    In order to avoid losses of energy and seawater pollution an off-shore nuclear power plant is coupled with a power plant which utilizes the temperature difference between seawater and hot reactor cooling water. According to the invention the power plant has a working media loop which is separated from the nuclear power plant. The apparative equipment and the operational characteristics of the power plant are the subject of the patent. (UWI) [de

  9. Managing ageing of Karachi nuclear power plant (KANUPP)

    International Nuclear Information System (INIS)

    Hoda, M. Qamrul; Jamal-ur-Rahman

    2002-01-01

    Full text: Karachi Nuclear power Plant (KANUPP), a 137 MWe CANDU plant, was built on a turn-key basis by the Canadian General Electric (CGE) in the late sixties. The plant with a design life of thirty years went into commercial operation in October, 1972. After nearly three decades of operation, KANUPP like all other plants has faced the problem of equipment ageing and obsolescence. KANUPP has been striving hard to combat these problems with the assistance from IAEA, COG and WANO. During early eighties IAEA expert missions were called at KANUPP on various safety issues and on their recommendations various projects such as Safe Operation of KANUPP, Technical Upgradation Project, Balancing, Modernization and Rehabilitation project were initiated to combat equipment ageing and obsolescence. KANUPP has made substantial progress in the implementation of the tasks under these projects and the operating life of the plant is expected to be extended by 15 years. Three IAEA expert missions were invited at KANUPP during 1999-2000 to carry out assessment of the ongoing activities related to plant ageing management. On the basis of their recommendations an ageing Management Programme has been formally established at KANUPP to manage age-related degradation of plant systems, structures and components important to safety and to ensure that the required safety functions are available for the extended life of the plant. Being close to completion of its nominal design life in October 2002, KANUPP has already applied to Pakistan Nuclear Regulatory Authority for renewal of operating license. This paper briefly describes the activities related to ageing management of KANUPP to fulfil the requirement for its continued operation beyond its design life. (author)

  10. CANDU fuel cycle options in Korea

    International Nuclear Information System (INIS)

    Boczar, P.G.; Fehrenbach, P.J.; Meneley, D.A.

    1996-04-01

    The easiest first step in CANDU fuel-cycle evolution may be the use of slightly enriched uranium (SEU), including recovered uranium from reprocessed LWR spent fuel. Relatively low enrichment (up to 1.2%) will result in a twoto three-fold reduction in the quantity of spent fuel per unit energy production, reductions in fuel-cycle costs, and greater flexibility in the design of new reactors. The CANFLEX (CANDU FLEXible) fuel bundle would be the optimal fuel carrier. A country that has both CANDU and PWR reactors can exploit the natural synergism between these two reactor types to minimize overall waste production, and maximize energy derived from the fuel. This synergism can be exploited through several different fuel cycles. A high burnup CANDU MOX fuel design could be used to utilize plutonium from conventional reprocessing or more advanced reprocessing options (such as co-processing). DUPIC (Direct Use of Spent PWR Fuel In CANDU) represents a recycle option that has a higher degree of proliferation resistance than does conventional reprocessing, since it uses only dry processes for converting spent PWR fuel into CANDU fuel, without separating the plutonium. Good progress is being made in the current KAERI, AECL, and U.S. Department of State program in demonstrating the technical feasibility of DUPIC. In the longer term, CANDU reactors offer even more dramatic synergistic fuel cycles with PWR or FBR reactors. If the objective of a national fuel-cycle program is the minimization of actinide waste or destruction of long-lived fission products, then studies have shown the superiority of CANDU reactors in meeting this objective. Long-term energy security can be assured either through the thorium cycle or through a CANDU 1 FBR system, in which the FBR would be operated as a 'fuel factory,' providing the fissile material to power a number of lower-cost, high efficiency CANDU reactors. In summary, the CANDU reactor's simple fuel design, high neutron economy, and on

  11. Standard compliance - NDE performance demonstration/inspection in the CANDU industry

    International Nuclear Information System (INIS)

    Choi, E.

    2011-01-01

    CANDU nuclear power plants are operated in 3 provinces in Canada for electric power generation. A table in the paper will show the built and operating plants in Ontario, Quebec, New Brunswick and overseas. The regulator for nuclear power in Canada is the Canadian Nuclear Safety Commission (CNSC). The CNSC holds the plant licensees accountable for compliance to CSA N285.4 for periodic inspections. The Standard basically specifies the 'what, when, where, how, how much and how frequently' NDE is to be done on pressure retaining systems and components in CANDU nuclear power plants. In inspection methods, the Standard specifies they must be non-destructive. The NDE methods were grouped into visual, dimensional, surface, volumetric and integrative. The Standard also specifies that the licensees are responsible for the performance demonstration (PD) of the adequacy of the procedures and the proficiency of the personnel. This paper describes the Standard's requirement in NDE qualification and presents a joint project participated by Canadian and overseas CANDU owners. The sub-project for NDE included providing evidence and technical justification on the adequacy of the procedures and the proficiency of the personnel. The paper describes the qualification methodology followed by the participants. This will be followed by how the participants produced Inspection Specification, tools and procedures, personnel training and qualification programs, test and qualification samples, independent peer reviews and Technical Justification. (author)

  12. Maintenance, rehabilitation, long life-the CANDU potential

    International Nuclear Information System (INIS)

    Torgerson, D.F.; Charlebois, P.; Hopkins, J.

    1998-01-01

    Plant life extension beyond the original design life is becoming an attractive economic consideration in the nuclear industry. Plant Life Management and life extension considerations have been built into the complete life cycle of the CANDU plant. The plant life management studies demonstrate that life extension for operating plants beyond 30 years is economically viable. The new CANDU designs benefit from this experience feedback and as a result, the plant design basis is now 40 years or better with potential for economical life extension. AECL is therefore confident that the new CANDU designs will exceed the performance record of the first generation CANDU 6 units and is committed to providing continued support and services during the operating life of the plant

  13. Docommissioning of nuclear power plants

    International Nuclear Information System (INIS)

    Essmann, J.

    1981-01-01

    The German utilities operating nuclear power plants have long concerned themselves with aspects of decommissioning and for this purpose an engineering company was given a contract to study the entire spectrum of decommissioning. The results of this study have been available in autumn 1980 and it is possible to discuss all the aspects of decommissioning on a new basis. Following these results no change in the design concept of LWR nuclear power plants in operation or under construction is necessary because the techniques, necessary for decommissioning, are fully available today. The technical feasibility of decommissioning for power plants of Biblis A and KRB type has been shown in detail. The calculations of the quantity of waste produced during removal of a nuclear power plant could be confirmed and it could be determined with high procedure. The radiation dose to the decommissioning personnel is in the range of the radiation protection regulations and is in the same range as the radiation dose to the personnel within a yearly inservice inspection. (AF)

  14. Fire protection in power plants

    International Nuclear Information System (INIS)

    Penot, J.

    1986-01-01

    Graphex-CK 23 is a unique sodium fire extinction product. Minimum amounts of powder are required for very fast action. The sodium can be put to use again, when the fire has been extinguished. It can be applied in other industrial branches and with other metals, e.g. sodium/potassium circuits or lithium coolant in power plants. [de

  15. Noise from wind power plants

    International Nuclear Information System (INIS)

    Ljunggren, S.

    2001-12-01

    First, the generation of noise at wind power plants and the character of the sound is described. The propagation of the sound and its dependence on the structure of the ground and on wind and temperature is treated next. Models for calculation of the noise emission are reviewed and examples of applications are given. Different means for reducing the disturbances are described

  16. Underwater nuclear power plant structure

    International Nuclear Information System (INIS)

    Severs, S.; Toll, H.V.

    1982-01-01

    A structure for an underwater nuclear power generating plant comprising a triangular platform formed of tubular leg and truss members upon which are attached one or more large spherical pressure vessels and one or more small cylindrical auxiliary pressure vessels. (author)

  17. World nuclear power plant capacity

    International Nuclear Information System (INIS)

    1991-01-01

    This report provides the background information for statistics and analysis developed by NUKEM in its monthly Market Report on the Nuclear Fuel Cycle. The assessments in this Special Report are based on the continuous review of individual nuclear power plant projects. This Special Report begins with tables summarizing a variety of nuclear power generating capacity statistics for 1990. It continues with a brief review of the year's major events regarding each country's nuclear power program. The standard NUKEM Market Report tables on nuclear plant capacity are given on pages 24 and 25. Owing to space limitations, the first year shown is 1988. Please refer to previous Special Reports for data covering earlier years. Detailed tables for each country list all existing plants as well as those expected by NUKEM to be in commercial operation by the end of 2005. An Appendix containing a list of abbreviations can be found starting on page 56. Only nuclear power plants intended for civilian use are included in this Special Report. Reactor lifetimes are assumed to be 35 years for all light water reactors and 30 years for all other reactor types, unless other data or definite decommissioning dates have been published by the operators. (orig./UA) [de

  18. Nuclear power plant emergency preparedness

    International Nuclear Information System (INIS)

    2005-01-01

    The guide sets forth detailed requirements on how the licensee of a nuclear power plant shall plan, implement and maintain emergency response arrangements. The guide is also applied to nuclear material and nuclear waste transport in situations referred to in guide YVL 6.5. Requirements on physical protection are presented in a separate guide of Finnish Radiation and Nuclear Safety Authority (STUK)

  19. Westinghouse ICF power plant study

    International Nuclear Information System (INIS)

    Sucov, E.W.

    1980-10-01

    In this study, two different electric power plants for the production of about 1000 MWe which were based on a CO 2 laser driver and on a heavy ion driver have been developed and analyzed. The purposes of this study were: (1) to examine in a self consistent way the technological and institutional problems that need to be confronted and solved in order to produce commercially competitive electricity in the 2020 time frame from an inertial fusion reactor, and (2) to compare, on a common basis, the consequences of using two different drivers to initiate the DT fuel pellet explosions. Analytic descriptions of size/performance/cost relationships for each of the subsystems comprising the power plant have been combined into an overall computer code which models the entire plant. This overall model has been used to conduct trade studies which examine the consequences of varying critical design values around the reference point

  20. Valve maintainability in CANDU-PHW nuclear generating stations

    International Nuclear Information System (INIS)

    Pothier, N.E.; Crago, W.A.

    1977-09-01

    Design, application, layout and administrative factors which affect valve maintainability in CANDU-PHW power reactors are identified and discussed. Some of these are illustrated by examples based on prototype reactor operation experience. Valve maintainability improvements resulting from laboratory development and maintainability analysis, have been incorporated in commercial CANDU-PHW nuclear generating stations. These, also, are discussed and illustrated. (author)

  1. Power conditioning devices in nuclear power plants

    International Nuclear Information System (INIS)

    Shida, Toichi.

    1979-01-01

    Purpose: To automatically prevent the liquid level from lowering in a reactor even if a feedwater adjusting valve is locked in a bwr type nuclear power plant. Constitution: Where a feedwater adjusting valve should be locked, and if the mismatching degree between the main steam flow rate and the feedwater flow rate exceeds a predetermined value and the mismatched state continues for a certain period, the value set to a main control for setting the recycling flow rate is changed corresponding to the mismatching degree to compensate the same thereby preventing the liquid level from lowering in the reactor. (Ikeda, J.)

  2. Fuzzy logic control for improved pressurizer systems in nuclear power plants

    International Nuclear Information System (INIS)

    Brown, Chris; Gabbar, Hossam A.

    2014-01-01

    Highlights: • Improved performance of the pressurizer system in a CANDU nuclear power plant (NPP). • Inventory control for the pressurizer system in NPP. • Compare fuzzy logic with PID in pressurizer system in NPP. • Develop a fuzzy controller to regulate the pressurizer inventory control. • Compare control performance with current proportional controller used at NPP. - Abstract: The pressurizer system in a CANDU nuclear power plant is responsible for maintaining the pressure of the primary heat transport system to ensure the plant is operated within its safe operating envelope. The inventory control for the pressurizer system use a combination of level sensors, feed valves and bleed valves to ensure that there is adequate room in the pressurizer to accommodate any swell or shrinkage in the PHT system. The Darlington Nuclear Generating Station (DNGS) in Ontario, Canada currently uses a proportional controller for the bleed and feed valves to regulate the pressurizer inventory control which can result in large coolant level overshoot along with excessive settling times. The purpose of this paper is to develop a fuzzy controller to regulate the pressurizer inventory control and compare its performance to the current proportional controller used at DNGS. The simulation of the pressurizer inventory control system shows the fuzzy controller performs better than the proportional controller in terms of settling time and overshoot

  3. Nuclear power plant life management

    International Nuclear Information System (INIS)

    Rorive, P.; Berthe, J.; Lafaille, J.P.; Eussen, G.

    1998-01-01

    Several definitions can be given to the design life of a nuclear power plant just as they can be attributed to the design life of an industrial installation: the book-keeping life which is the duration of the provision for depreciation of the plant, the licensed life which corresponds to the duration for which the plant license has been granted and beyond which a new license should be granted by the safety authorities, the design life which corresponds to the duration specified for ageing and fatigue calculations in the design of some selected components during the plant design phase, the technical life which is the duration of effective technical operation and finally the economic life corresponding to the duration of profitable operation of the plant compared with other means of electricity production. Plant life management refers to the measures taken to cope with the combination of licensed, design, technical and economical life. They can include repairs and replacements of components which have arrived to the end of their life due to known degradation processes such as fatigue, embrittlement, corrosion, wear, erosion, thermal ageing. In all cases however, it is of great importance to plan the intervention so as to minimise the economic impact. Predictive maintenance is used together with in-service inspection programs to fulfil this goal. The paper will go over the methodologies adopted in Belgium in all aspects of electrical, mechanical and civil equipment for managing plant life. (author)

  4. Ultimate pressure capacity of CANDU 6 containment structures

    International Nuclear Information System (INIS)

    Radulescu, J.P.; Pradolin, L.; Mamet, J.C.

    1997-01-01

    This paper summarizes the analytical work carried out and the results obtained when determining the ultimate pressure capacity (UPC) of the containment structures of CANDU 6 nuclear power plants. The purpose of the analysis work was to demonstrate that such containment structures are capable of meeting design requirements under the most severe accident conditions. For this concrete vessel subjected to internal pressure, the UPC was defined as the pressure causing through cracking in the concrete. The present paper deals with the overall behaviour of the containment. The presence of openings, penetrations and the ultimate pressure of the airlocks were considered separately. (author)

  5. Role of water lubricated bearings in Candu reactors

    International Nuclear Information System (INIS)

    Kumar, Ashok N.

    1999-01-01

    During the twentieth century a great emphasis was placed in understanding and defining the operating regime of oil and grease lubricated components. Major advances have been made through elastohydrodynamic lubrication theory in the quantifying the design life of heavily loaded components such as rolling element bearings and gears. Detailed guidelines for the design of oil and grease lubricated components are widely available and are being applied to the successful design of these components. However similar guidelines for water lubricated components are either not available or not well documented. It is often forgotten that the water was used as a lubricant in several components as far back as 1884 B.C. During the twentieth century the water lubricated components continued to play a major role in some high technology industries such as in the power generation plants. In CANDU nuclear reactors water lubrication of several critical components always occupied a pride place and in most cases the only practical mode of lubrication of several critical components always occupied a pride place and in most cases the only practical mode of lubrication. This paper presents some examples of the major water lubricated components in a CANDU reactors. Major part of the paper is focused on presenting an example of successful operating history of water lubricated bearings used in the HT pumps are presented. Both types of bearings have been qualified by tests for operation under normal as well as under more severe postulated condition of loss-of-coolant-accident (LOCA). These bearings have been designed to operate for the 30 years in the existing CANDU 6 (600 MW) reactors. However for the next generation of CANDU 6 reactors which go into service in the year 2003, the HT pump bearing life has been extended to 40 years. (author)

  6. Operation and maintenance of nuclear power plants

    International Nuclear Information System (INIS)

    Ackermann, G.

    1987-01-01

    This textbook gives a systematic introduction into the operational and maintenance activities in nuclear power plants with pressurized water reactors. Subjects: (1) Setup and operational behaviour of power reactors, (2) setup of nuclear power plants, (3) radiation protection and nuclear safety, (4) nuclear fuel, (5) constructional layout of nuclear power plants, (6) management, and (7) maintenance. 158 figs., 56 tabs

  7. Report of the Federal Ministry for the Environment, Nature Conservation, Buildings and Nuclear Safety (BMUB) on the topical peer review aging management in nuclear power plants and research reactors

    International Nuclear Information System (INIS)

    2017-01-01

    The report of the Federal Environmental Ministry (BMUB) on the topical peer review aging management in nuclear power plants and research reactors covers the following issues: comprehensive requirements for aging management and its implementation, electric cables, non accessible pipes, reactor pressure vessel, calandria/pressure tubes (CANDU), concrete containment, pre-stressed concrete reactor pressure vessel (AGR).

  8. Plant diagnostics in power stations

    International Nuclear Information System (INIS)

    Sturm, A.; Doering, D.

    1985-01-01

    The method of noise diagnostics is dealt with as a part of plant diagnostics in nuclear power stations. The following special applications are presented: (1) The modular noise diagnostics system is used for monitoring primary coolant circuits and detecting abnormal processes due to mechanical vibrations, loose parts or leaks. (2) The diagnostics of machines and plants with antifriction bearings is based on bearing vibration measurements. (3) The measurement of the friction moment by means of acoustic emission analysis is used for evaluating the operational state of slide bearings

  9. Maintenance of Nuclear Power Plants

    International Nuclear Information System (INIS)

    Quintana, J. M.; Sanchez, J. T.

    2002-01-01

    With this article about the Maintenance in nuclear power plants we will try to give to see the importance of this kind of installations but the problems found by the clients and contractors to face it, and some possible solutions to improve it. It is necessary to understand this problem like something inner to the installation and must be considerate like a benefit for the same. Of course, there must be adequate Sevices Companies in direct relation with the installation that take the responsibility of assuming and understanding the correct fulfillment of the fixed milestones to get the optimal working of the whole plant systems. (Author)

  10. CANDU advanced fuel R and D programs for 1997 - 2006 in Korea

    International Nuclear Information System (INIS)

    Suk, H.C.; Yang, M.S.; Sim, K-S.; Yoo, K.J.

    1997-01-01

    KAERI has a comprehensive product development program of CANFLEX and DUPIC fuels to introduce them into CANDU reactors in Korea and a clear vision of how the product will evolve over the next 10 years. CANDU reactors are not the majority of nuclear power plants in Korea, but they produce significant electricity to contribute Korea's economic growth as well as to satisfy the need for energy. The key targets of the development program are safety enhancement, reduction of spent fuel volume, and economic improvements, using the inherent characteristics and advantages of CANDU technology The CANFLEX and DUPIC R and D programs are conducted currently under the second stage of Korea's Nuclear Energy R and D Project as a national mid- and long-term program over the next 10 years from 1997 to 2006. The specific activities of the programs have taken account of the domestic and international environment concerning on non-proliferation in the Peninsula of Korea. As the first of the development products in the short-term, the CANFLEX-NU fuel will be completely developed jointly by KAERI/AECL and will be useful for the older CANDU-6 Wolsong unit 1. As the second product, the CANFLEX-0.9 % equivalent SEU fuel is expected to be completely developed within the next decade. It will be used in CANDU-6 reactors in Korea immediately after the development, if the existing RU in the world is price competitive with natural uranium. The DUPIC R and D program, as a long term program, is expected to demonstrate the possibility of use of used PWR fuel in CANDU reactors in Korea during the next 10 years. The pilot scale fabrication facility would be completed around 2010. (author)

  11. CANDU advanced fuel R and D programs for 1997 - 2006 in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Suk, H.C.; Yang, M.S.; Sim, K-S.; Yoo, K.J. [Korea Atomic Energy Research Inst., Yusong, Taejon (Korea, Republic of)

    1997-07-01

    KAERI has a comprehensive product development program of CANFLEX and DUPIC fuels to introduce them into CANDU reactors in Korea and a clear vision of how the product will evolve over the next 10 years. CANDU reactors are not the majority of nuclear power plants in Korea, but they produce significant electricity to contribute Korea's economic growth as well as to satisfy the need for energy. The key targets of the development program are safety enhancement, reduction of spent fuel volume, and economic improvements, using the inherent characteristics and advantages of CANDU technology The CANFLEX and DUPIC R and D programs are conducted currently under the second stage of Korea's Nuclear Energy R and D Project as a national mid- and long-term program over the next 10 years from 1997 to 2006. The specific activities of the programs have taken account of the domestic and international environment concerning on non-proliferation in the Peninsula of Korea. As the first of the development products in the short-term, the CANFLEX-NU fuel will be completely developed jointly by KAERI/AECL and will be useful for the older CANDU-6 Wolsong unit 1. As the second product, the CANFLEX-0.9 % equivalent SEU fuel is expected to be completely developed within the next decade. It will be used in CANDU-6 reactors in Korea immediately after the development, if the existing RU in the world is price competitive with natural uranium. The DUPIC R and D program, as a long term program, is expected to demonstrate the possibility of use of used PWR fuel in CANDU reactors in Korea during the next 10 years. The pilot scale fabrication facility would be completed around 2010. (author)

  12. Distributed computer control systems in future nuclear power plants

    International Nuclear Information System (INIS)

    Yan, G.; L'Archeveque, J.V.R.; Watkins, L.M.

    1978-09-01

    Good operating experience with computer control in CANDU reactors over the last decade justifies a broadening of the role of digital electronic and computer related technologies in future plants. Functions of electronic systems in the total plant context are reappraised to help evolve an appropriate match between technology and future applications. The systems research, development and demonstration program at CRNL is described, focusing on the projects pertinent to the real-time data acquisition and process control requirements. (author)

  13. Decommissioning of nuclear power plants

    International Nuclear Information System (INIS)

    Friske, A.; Thiele, D.

    1988-01-01

    The IAEA classification of decommissioning stages is outlined. The international development hitherto observed in decommissioning of nuclear reactors and nuclear power stations is presented. The dismantling, cutting and decontamination methods used in the decommissioning process are mentioned. The radioactive wastes from decommissioning are characterized, the state of the art of their treatment and disposal is given. The radiation burdens and the decommissioning cost in a decommissioning process are estimated. Finally, some evaluation of the trends in the decommissioning process of nuclear power plants is given. 54 refs. (author)

  14. Advancing CANDU Technology Through R and D

    International Nuclear Information System (INIS)

    Torgerson, David F.

    1993-01-01

    CANDU reactors are evolving to meet future requirements using incremental changes as opposed to revolutionary design changes. The main elements for advancing the technology reducing capital and operating, increasing capacity factors, increasing passive safety, and enhancing fuel/fuel cycle flexibility. These elements are being addressed by carrying out research and development in the areas of safety, plant systems and components, heavy water production, information technology, fuel channels, and fuel/fuel cycle technology. In safety, the focus is on using the inherent features of CANDU to enhance passive or natural safety concepts, such as the use of the moderator as an effective heat sink, and the development of advanced fuels to improve critical heat flux and to reduce source terms. Plant systems and components work includes improvements to plant systems such as steam generators, heat exchangers, pump seals, and advanced control room technology. Heavy water processes are being developed that can be used with existing hydrogen production plants, or that can be used in a stand-alone mode. Information technology is being developed to cover all aspects of CANDU design, construction, and operation. Fuel channel improvements include elucidation and application of basic materials science for life extension, and the development of advanced non-destructive examination methods. Fuel and fuel cycle work is focusing on LWR/CANDU synergy, such as the use of recovered uranium and the direct use of spent PWR fuel in CANDU reactor, advanced fuels to improve burnup and economics (e. g., the joint AECB/KAERI Conflux program), and low void reactivity fuel to enhance passive safety. This paper gives an overview of some of the R and D supporting these activities, with particular emphasis on Alice's vision for advancing CANDU technology over the next 10 years

  15. CANDU-PHW fuel management

    International Nuclear Information System (INIS)

    Frescura, G.M.; Wight, A.L.

    1982-01-01

    This report covers the material presented in a series of six lectures at the Winter College on Nuclear Physics and Reactors held at the International Centre for Theoretical Physics in Trieste, Italy, Jan 22 - March 28, 1980. The report deals with fuel management in natural uranium fuelled CANDU-PHW reactors. Assuming that the reader has a basic knowledge of CANDU core physics, some of the reactor systems which are more closely related to fuelling are described. This is followed by a discussion of the methods used to calculate the power distribution and perform fuel management analyses for the equilibrium core. A brief description of some computer codes used in fuel management is given, together with an overview of the calculations required to provide parameters for core design and support the accident analysis. Fuel scheduling during approach to equilibrium and equilibrium is discussed. Fuel management during actual reactor operation is discussed with a review of the operating experience for some of the Ontario Hydro CANDU reactors. (author)

  16. Safety in nuclear power plants

    International Nuclear Information System (INIS)

    Koeberlein, K.

    1987-01-01

    In nuclear power plants large amounts of radioactive fission products ensue from the fission of uranium. In order to protect the environment, the radioactive material is confined in multiple 'activity barriers' (crystal matrix of the fuel, fuel cladding, coolant boundary, safety containment, reactor building). These barriers are protected by applying a defense-in-depth concept (high quality requirements, protection systems which recognize and terminate operational incidents, safety systems to cope with accidents). In spite of a favorable safety record of German nuclear power plants it is obvious - and became most evident by the Chernobyl accident - that absolute safety is not achievable. At Chernobyl, however, design disadvantages of that reactor type (like positive reactivity feedback of coolant voiding, missing safety containment) played an important role in accident initiation and progression. Such features of the Russian 'graphite-moderated pressure tube boiling water reactor' are different from those of light water reactors operating in western countries. The essential steps of the waste management of the nuclear fuel cycle ('Entsorgung') are the interim storage, the shipment, and the reprocessing of the spent fuel and the final repository of radioactive waste. Reprocessing means the separation of fossil material (uranium, plutonium) from radioactive waste. Legal requirements for radiological protection of the environment, which are identical for nuclear power plants and reprocessing plant, are complied with by means of comprehensive filter systems. Safety problems of a reprocessing plant are eased considerably by the fact that system pressures, process temperatures and energy densities are low. In order to confine the radioactive waste from the biosphere for a very long period of time, it is to be discarded after appropriate treatment into the deep geological underground of salt domes. (orig./HP) [de

  17. Nuclear power plants and environment

    International Nuclear Information System (INIS)

    Agudo, E.G.; Penteado Filho, A.C.

    1980-01-01

    The question of nuclear power plants is analysed in details. The fundamental principles of reactors are described as well as the problems of safety involved with the reactor operation and the quantity and type of radioactive released to the environment. It shows that the amount of radioactive is very long. The reactor accidents has occurred, as three mile island, are also analysed. (M.I.A.)

  18. Candu 6: versatile and practical fuel technology

    International Nuclear Information System (INIS)

    Hopwood, J. M.; Saroudis, J.

    2013-01-01

    CANDU reactor technology was originally developed in Canada as part of the original introduction of peaceful nuclear power in the 1960s and has been continuously evolving and improving ever since. The CANDU reactor system was defined with a requirement to be able to efficiently use natural uranium (NU) without the need for enrichment. This led to the adaptation of the pressure tube approach with heavy water coolant and moderator together with on-power fuelling, all of which contribute to excellent neutron efficiency. Since the beginning, CANDU reactors have used [NU] fuel as the fundamental basis of the design. The standard [NU] fuel bundle for CANDU is a very simple design and the simplicity of the fuel design adds to the cost effectiveness of CANDU fuelling because the fuel is relatively straightforward to manufacture and use. These characteristics -- excellent neutron efficiency and simple, readily-manufactured fuel -- together lead to the unique adaptability of CANDU to alternate fuel types, and advancements in fuel cycles. Europe has been an early pioneer in nuclear power; and over the years has accumulated various waste products from reactor fuelling and fuel reprocessing, all being stored safely but which with passing time and ever increasing stockpiles will become issues for both governments and utilities. Several European countries have also pioneered in fuel reprocessing and recycling (UK, France, Russia) in what can be viewed as a good neighbor policy to make most efficient use of fuel. The fact remains that CANDU is the most fuel efficient thermal reactor available today [NU] more efficient in MW per ton of U compared to LWR's and these same features of CANDU (on-power fuelling, D 2 O, etc) also enable flexibility to adapt to other fuel cycles, particularly recycling. Many years of research (including at ICN Pitesti) have shown CANDU capability: best at Thorium utilization; can use RU without re-enrichment; can readily use MOX. Our premise is that

  19. Availability of thermal power plants

    International Nuclear Information System (INIS)

    Nitsch, D.; Schmitz, H.

    1981-01-01

    Availability data based on unique uniform, and clearly defined concepts and methods of acquisition have been compiled by the VGB since 1970. The data are published in anual reports. These reports contain availability data of fossil-fuelled units, combined gas/steam units, nuclear power plants, and gas turbine plants in Germany and abroad, listed by unit size fuel type, time of operation, and application. For the purpose of comparison, the data for the years since 1970 are presented as well as data averaged for the whole period under report. The main results for the year 1980 are presented now that the greater part of the plants has been evaluated. The complete evaluation will be published towards the end of 1981. (orig.) [de

  20. Elecnuc. Nuclear power plants in the world

    International Nuclear Information System (INIS)

    2003-01-01

    This 2003 version of Elecnuc contents information, data and charts on the nuclear power plants in the world and general information on the national perspectives concerning the electric power industry. The following topics are presented: 2002 highlights; characteristics of main reactor types and on order; map of the French nuclear power plants; the worldwide status of nuclear power plants on 2002/12/3; units distributed by countries; nuclear power plants connected to the Grid by reactor type groups; nuclear power plants under construction; capacity of the nuclear power plants on the grid; first electric generations supplied by a nuclear unit; electrical generation from nuclear plants by country at the end 2002; performance indicator of french PWR units; trends of the generation indicator worldwide from 1960 to 2002; 2002 cumulative Load Factor by owners; nuclear power plants connected to the grid by countries; status of license renewal applications in Usa; nuclear power plants under construction; Shutdown nuclear power plants; exported nuclear power plants by type; exported nuclear power plants by countries; nuclear power plants under construction or order; steam generator replacements; recycling of Plutonium in LWR; projects of MOX fuel use in reactors; electricity needs of Germany, Belgium, Spain, Finland, United Kingdom; electricity indicators of the five countries. (A.L.B.)